ML20317A001
| ML20317A001 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 12/28/2020 |
| From: | Joel Wiebe Plant Licensing Branch III |
| To: | Bryan Hanson Exelon Generation Co, Exelon Nuclear |
| Wiebe J | |
| References | |
| EPID L-2020-LLA-0038 | |
| Download: ML20317A001 (35) | |
Text
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION December 28, 2020 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO)
Exelon 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. 219, 219, 223, AND 223 REGARDING REVISION OF TECHNICAL SPECIFICATIONS 5.6.5, CORE OPERATING LIMITS REPORT (COLR) (EPID L-2020-LLA-0038)
Dear Mr. Hanson:
The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 219 to Renewed Facility Operating License No. NPF-72 and Amendment No. 219 to Renewed Facility Operating License No. NPF-77 for the Braidwood Station, Units 1 and 2 (Braidwood), and Amendment No. 223 to Renewed Facility Operating License No. NPF-37 and Amendment No. 223 to Renewed Facility Operating License No. NPF-66 for the Byron Station, Unit Nos. 1 and 2 (Byron). The amendments are in response to your application dated February 28, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20063L400).
The amendments revise Braidwood and Byron Technical Specification (TS) 5.6.5, Core Operating Limits Report (COLR). The proposed changes revise TS 5.6.5.b to replace the current NRC-approved loss-of-coolant accident (LOCA) methodologies with a single, newer NRC-approved LOCA methodology, the FULL SPECTRUM'1 LOCA Evaluation Model (FSLOCATM EM), that is contained in WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology).
The NRC staff has determined that the related safety evaluation contains proprietary information pursuant to Title 10 of the Code of Federal Regulations Section 2.390, Public inspections, exemptions, requests for withholding. The proprietary information is indicated by text enclosed with double brackets. The proprietary version of the safety evaluation is provided as 1 FULL SPECTRUM and FSLOCA are trademarks or registered trademarks of Westinghouse Electric Company LLC, its subsidiaries, and/or affiliates in the United States of America and may be registered in other countries through the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners. to this letter contains proprietary information. When separated from Enclosure 5, this document is DECONTROLLED.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION. The NRC staff has also prepared a non-proprietary version of the safety evaluation, which is provided in Enclosure 6.
A Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Joel S. Wiebe, 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. 219 to NPF-72
- 2. Amendment No. 219 to NPF-77
- 3. Amendment No. 223 to NPF-37
- 4. Amendment No. 223 to NPF-66
- 5. Safety Evaluation (proprietary)
- 6. Safety Evaluation (non-proprietary) cc without Enclosure 5: Listserv
EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-456 BRAIDWOOD STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 219 Renewed License No. NPF-72
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Exelon Generation Company, LLC (the licensee) dated February 28, 2020, 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 operating 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. 219 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 during refueling outage A1R22 currently scheduled for April 2021.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: December 28, 2020
EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-457 BRAIDWOOD STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 219 Renewed License No. NPF-77
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Exelon Generation Company, LLC (the licensee) dated February 28, 2020, 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 operating 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. 219 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 during refueling outage A2R22 currently scheduled for April 2022.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: December 28, 2020
ATTACHMENT TO LICENSE AMENDMENT NOS. 219 AND 219 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 (TSs), with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE INSERT License No. NPF-72 License No. NPF-72 Page 3 Page 3 License No. NPF-77 License No. NPF-77 Page 3 Page 3 TSs TSs Page 5.6 - 4 Page 5.6 - 4
(2)
Exelon Generation Company, 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)
Exelon Generation Company, 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)
Exelon Generation Company, 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)
Exelon Generation Company, 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. 219 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. 219
(2)
Exelon Generation Company, 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)
Exelon Generation Company, 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)
Exelon Generation Company, 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)
Exelon Generation Company, 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. 219 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. 219
EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-454 BYRON STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 223 Renewed License No. NPF-37
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Exelon Generation Company, LLC (the licensee) dated February 28, 2020, 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. 223 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 during refueling outage B1R24 currently scheduled for September 2021.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: December 28, 2020
EXELON GENERATION COMPANY, LLC DOCKET NO. STN 50-455 BYRON STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 223 Renewed License No. NPF-66
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Exelon Generation Company, LLC (the licensee) dated February 28, 2020, 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-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. 223 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 during refueling outage B2R23 currently scheduled for April 2022.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: December 28, 2020
ATTACHMENT TO LICENSE AMENDMENT NOS. 223 AND 223 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 License and Appendix A, Technical Specifications (TSs), with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
REMOVE INSERT License No. NPF-37 License No. NPF-37 Page 3 Page 3 License No. NPF-66 License No. NPF-66 Page 3 Page 3 TSs TSs Page 5.6 - 4 Page 5.6 - 4
(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. 223 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. 223
(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. 223 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.
Renewed License No. NPF-66 Amendment No. 223
Reporting Requirements 5.6 BYRON UNITS 1 & 2 5.6 4 Amendment 223 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 6.
WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," November 2016.
7.
Not Used.
8.
Not Used.
9.
WCAP-10216-P-A, Revision 1, "Relaxation of Constant Axial Offset Control - FQ Surveillance Technical Specification," February 1994.
10.
WCAP-8745-P-A, "Design Bases for the Thermal Overpower T and Thermal Overtemperature T Trip Functions,"
September 1986.
11.
WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis," October 1999.
12.
WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995, (Westinghouse Proprietary).
13.
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLOTM," July 2006, (Westinghouse Proprietary).
c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met; and d.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
ENCLOSURE 6 (NON-PROPRIETARY)
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 219 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-72, AMENDMENT NO. 219 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-77, AMENDMENT NO. 223 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-37, AND AMENDMENT NO. 223 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-66 EXELON GENERATION COMPANY, 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 Proprietary information has been redacted from this document pursuant to Section 2.390 of Title 10 of the Code of Federal Regulations.
Redacted information is identified by blank text enclosed within ((double brackets)).
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 219 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-72, AMENDMENT NO. 219 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-77, AMENDMENT NO. 223 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-37, AND AMENDMENT NO. 223 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-66 EXELON GENERATION COMPANY, 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
1.0 INTRODUCTION
By letter dated February 28, 2020 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML20063L400), Exelon Generation Company, LLC (the licensee) submitted a license amendment request (LAR) to Renewed Facility Operating License Nos. NPF-72 and NPF-77 for the Braidwood Station (Braidwood), Units 1 and 2, respectively, and Renewed Facility Operating License Nos. NPF-37 and NPF-66 for Byron Station (Byron),
Unit Nos. 1 and 2, respectively. Portions of the LAR contain sensitive unclassified non-safeguards information and accordingly, have been withheld from public disclosure pursuant to Section 2.390, Public inspection, exemptions, requests for withholding, of Title 10 of the Code of Federal Regulations (10 CFR).
The amendments would revise Technical Specification (TS) 5.6.5, Core Operating Limits Report (COLR), to replace the existing loss-of-coolant accident (LOCA) methodologies with the U.S. Nuclear Regulatory Commission (NRC, the Commission)-approved LOCA methodology contained in Westinghouse Electric Company, LLCs topical report WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology) dated November 2016 (ADAMS Package Accession No. ML17277A130), which was used for LOCA reanalysis for Braidwood, Units 1 and 2, and Byron, Units 1 and 2.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION
2.0 REGULATORY EVALUATION
2.1 Description of the Licensees Proposed Changes The licensee proposed changes to TS 5.6.5b for Braidwood, Units 1 and 2 and Byron, Units 1 and 2, to reflect the use of the NRC-approved LOCA methodology.
The current TS 5.6.5.b includes the following three NRC-approved LOCA methodologies (numbered as 5.6.5.b.6 through 5.6.5.b.8):
- 6.
WCAP-16009-P-A, Revision 0, Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM), January 2005.
- 7.
WCAP-10079-P-A, NOTRUMP, A Nodal Transient Small Break and General Network Code, August 1985.
- 8.
WCAP-10054-P-A, Westinghouse Small Break ECCS [Emergency Core Cooling System] Evaluation Model using NOTRUMP Code, August 1985.
The proposed revision to TS 5.6.5.b would replace these three LOCA methodologies with WCAP-16996-P-A, Revision 1, as shown below.
- 6.
WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), November 2016.
- 7.
Not Used.
- 8.
Not Used.
The NRC staffs evaluation in Section 3.0 of this SE determines that the licensee appropriately applied the Full Spectrum Loss of Coolant Accident (FSLOCA) Evaluation Methodology (EM) to Braidwood, Units 1 and 2 and Byron, Units 1 and 2. In Section 3.4 of this SE the NRC staff concludes that the existing three methodologies in TS 5.6.5.b.6 through 5.6.5.b.8 are no longer needed to address the full spectrum of LOCA break sizes. The FSLOCA EM approved method replaces the existing methodology listed in TS 5.6.5.b.6 and the existing methodologies in TS 5.6.5.b.7 and 5.6.5.b.8 are deleted.
2.2 Regulatory Review The NRC staff considered the following regulations and guidance during its review of the proposed changes.
Regulations Under 10 CFR 50.92(a), determinations on whether to issue an LAR are guided by the considerations that govern the issuance of initial licenses or construction permits to the extent applicable and appropriate. Both the common standards for licenses and construction permits in 10 CFR 50.40(a), and those specifically for issuance of operating licenses in
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public.
The regulation under 10 CFR 50.36 requires that TSs 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; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports.
The regulation under 10 CFR 50.90 requires NRC approval for any modification to, addition to, or deletion from the TSs.
The regulation under 10 CFR 50.36(c)(5) states that administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.
The following paragraphs of 10 CFR 50.46(b) require, in part, that:
(1)
Peak cladding temperature. The calculated maximum fuel element cladding temperature shall not exceed 2200 °F [degrees Fahrenheit].
(2)
Maximum cladding oxidation. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.
(3)
Maximum hydrogen generation. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
(4)
Coolable geometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling.
Guidance Documents The following documents provide guidance for the review of LOCA analysis:
NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, Section 15.6.5, Revision 3, Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary, dated March 2007 (ADAMS Accession No. ML070550016).
NRC Regulatory Guide (RG) 1.157, Best-Estimate Calculations of Emergency Core Cooling System Performance, dated May 1989 (ADAMS Accession No. ML003739584).
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION NRC RG 1.203, Transient and Accident Analysis Methods, dated December 2005 (ADAMS Accession No. ML053500170).
NRC Generic Letter 88-16, Removal of Cycle Specific Parameter Limits from Technical Specifications, October 4, 1988 (ADAMS Accession No. ML031200485).
3.0 TECHNICAL EVALUATION
The NRC staff evaluated the licensee's application to determine whether the proposed changes are consistent with the regulations and guidance discussed in Section 2.0 of this safety evaluation. The staff reviewed the proposed changes to verify that the new LOCA methodology is an approved NRC code and that all limitations and conditions are met, that the licensee appropriately applied the LOCA EM to Braidwood, Units 1 and 2 and Byron, Units 1 and 2, and that the results meet the acceptance criteria of 10 CFR 50.46(b)(1) through (4).
Separate analyses were performed by the licensee for Braidwood, Unit 1 and Byron, Unit 1, and Braidwood, Unit 2, and Byron, Unit 2, due to the different steam generator designs in the Unit 1 plants versus Unit 2 plants. The Braidwood, Unit 2, and Byron, Unit 2, plants have the original Westinghouse Model D5 steam generators installed; whereas, replacement steam generators supplied by Babcock and Wilcox International were installed in Braidwood, Unit 1, and Byron, Unit 1. Because different steam generator designs would be expected to affect transient behavior and analysis results, separate analyses were performed and the results provided in the LAR. The units are otherwise functionally similar with respect to the major systems and components.
3.1 Description of FULL SPECTRUM LOCA Methodology As described in WCAP-16996-P-A, Revision 1, the purpose of the Full Spectrum LOCA (FSLOCA) Methodology EM is to build on the ASTRUM EM, by extending the applicability of the WCOBRA/TRAC Code to include the treatment of small break LOCA (SBLOCA) and intermediate break LOCA (IBLOCA) scenarios. The term Full Spectrum specifies that the new EM is intended to resolve the full spectrum of LOCA scenarios that result from a postulated break in the cold leg of a pressurized-water reactor (PWR). The break sizes considered in the Westinghouse FSLOCA methodology include any break size in which break flow is beyond the capacity of the normal charging pumps, up to and including a double ended guillotine rupture with a break flow area equal to two times the pipe area.
The break size spectrum is divided into two regions. Region I includes breaks that are typically defined as SBLOCAs. Region II includes break sizes that are typically defined as large break LOCAs (LBLOCAs).
The FSLOCA EM explicitly considers the effects of fuel pellet thermal conductivity degradation (TCD) and other burnup-related effects by calibrating to fuel rod performance data input generated by the PAD5 code, which explicitly models TCD and is benchmarked to high burnup data. The fuel pellet thermal conductivity model in the WCOBRA/TRAC-TF2 code used in the FSLOCA EM explicitly accounts for pellet TCD.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.2 Limitations and Conditions The safety evaluation for WCAP-16996-P-A, Revision 1 (ADAMS Package Accession No. ML17207A124) contains 15 limitations and conditions that must be met in order to implement the NRC-approved FSLOCA EM.
A summary of each limitation and condition and how it has been met as stated by the licensee in its application dated February 28, 2020, and the associated NRC staff findings are provided below.
Limitation and Condition 1 - Applicability with Regard to LOCA Transient Phases The FSLOCA EM is not approved to demonstrate compliance with 10 CFR 50.46 acceptance criterion (b)(5) related to the long-term cooling.
The analyses for Braidwood, Units 1 and 2 and Byron, Units 1 and 2 with the FSLOCA EM are only being used to demonstrate compliance with 10 CFR 50.46 (b)(1) through (b)(4).
Given that the licensee is not using the FSLOCA EM to demonstrate compliance with 10 CFR 50.46(b)(5), the NRC staff finds that the licensee has met the requirements of Limitation and Condition 1.
Limitation and Condition 2 - Applicability with Regard to Type of PWR Plants The FSLOCA EM is approved for the analysis of Westinghouse-designed 3-loop and 4-loop PWRs with cold-side injection. Analyses should be executed consistent with the approved method, or any deviations from the approved method should be described and justified.
Braidwood, Units 1 and 2, and Byron, Units 1 and 2, are Westinghouse-designed 4-loop PWRs with cold-side injection, so they are within the scope of the NRC-approved methodology. The analyses for Braidwood, Units 1 and 2, and Byron, Units 1 and 2, utilize the NRC-approved FSLOCA methodology, except for the changes which were previously transmitted to the NRC pursuant to 10 CFR 50.46 in Westinghouse Letter, LTR-NRC-18-30, dated July 18, 2018 (ADAMS Accession No. ML19288A174). The NRC staff notes that in the attachment to the letter, Westinghouse categorized these changes or errors into two separate groups, non-discretionary and discretionary changes. Four of those changes affected FSLOCA EM.
The estimated effect of three of the changes caused no impact on the calculated peak cladding temperature (PCT). The fourth change is an improvement in the pump component momentum equation at low pump speed which will either have no impact or a small penalizing impact on Region I analyses.
After completion of the analyses for Braidwood, Units 1 and 2, and Byron, Units 1 and 2, two errors were discovered in the WCOBRA/TRAC-TF2 code that can occur under certain conditions. These errors were reported to the NRC in Westinghouse Letter, LTR-NRC-19-6, dated February 7, 2019 (ADAMS Accession No. ML19042A380) pursuant to the 10 CFR 50.46 reporting requirement.
Review of the attachment to the letter by the NRC staff reveals that one of the errors has no impact on analyses performed using the FSLOCA EM, and that engineering judgment supported by sensitivity calculations showed that correcting the other error had minimal impact on LOCA transient calculations, leading to an estimated PCT impact of 0 °F.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The licensee stated in its LAR that the power increase in the hot rod and hot assembly due to energy redistribution in the application of the FSLOCA EM to Braidwood, Units 1 and 2, and Byron, Units 1 and 2, was calculated incorrectly. This error resulted in a 0 percent to 5 percent deficiency in the modeled hot rod and hot assembly rod linear heat rates on a run-specific basis, depending on the as-sampled value for the uncertainty. The effect of the error correction was evaluated in the context of the application of the FSLOCA EM to Braidwood, Units 1 and 2, and Byron, Units 1 and 2. The licensee found that the error has only a limited impact on the power modeled for a single assembly in the core. Based on the above, the NRC staff found that there is a negligible impact of the error correction on the system thermal-hydraulic response during the postulated LOCA.
For Region I, the primary impact of the error correction is on the rate of cladding heatup above the two-phase mixture level in the core during the boil off phase. The PCT impact was assessed using run-specific PCT versus linear heat rate relationships and the run-specific hot rod and hot assembly linear heat rate increase that would result from the error correction. Using this approach, the correction of the error was estimated by the licensee to increase the Region I analysis PCT by 17 F for Braidwood, Unit 1 and Byron, Unit 1, leading to a final result of 1181 °F for the Region I analysis. Using this approach, the correction of the error was estimated by the licensee to increase the Region I analysis PCT by 1 °F for Braidwood, Unit 2 and Byron, Unit 2, leading to a final result of 1169 °F for the Region I analysis.
For Region II, parametric PWR sensitivity studies, derived from a subset of uncertainty analysis simulations covering various design features and fuel arrays, were examined by the licensee to determine the sensitivity of the analysis results to the error correction. The PCT impact from the error correction was found by the licensee to be different for the different transient phases (i.e.,
blowdown versus reflood) based on the PWR sensitivity studies and existing power distribution sensitivity studies. Based on the results from the PWR sensitivity studies, the correction of the error is estimated by the licensee to increase the Region II analysis PCT by 31 °F for Braidwood, Unit 1, and Byron, Unit No. 1, leading to an analysis result of 1643°F for the Region II analysis assuming loss-of-offsite power (LOOP) and 1641 °F for the Region II analysis assuming offsite power available (OPA). Based on the results from the PWR sensitivity studies, the correction of the error is estimated by the licensee to increase the Region II analysis PCT by 31 °F for Braidwood, Unit 2, and Byron, Unit 2, leading to an analysis result of 1711 °F for the Region II analysis assuming LOOP and 1752 °F for the Region II analysis assuming OPA.
Given that Braidwood, Units 1 and 2, and Byron, Unit 1 and 2, are Westinghouse-designed 4-loop PWRs with cold-side injection, the NRC staff finds that the FSLOCA EM is applicable to them. In addition, the NRC staff finds that the licensee has appropriately applied the FSLOCA EM with the changes or errors, as described above. Therefore, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 2.
Limitation and Condition 3 - Applicability for Containment Pressure Modeling For Region II, the containment pressure calculation will be executed in a manner consistent with the approved methodology (i.e., the COCO or LOTIC2 model will be based on appropriate plant-specific design parameters and conditions, and engineered safety features which can reduce pressure are modeled). This includes utilizing a plant-specific initial containment temperature, and only taking
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION credit for containment coatings which are qualified and outside of the break zone-of-influence.
The containment pressure calculation for the Braidwood, Units 1 and 2 and Byron, Units 1 and 2, analyses was performed consistent with the NRC-approved methodology. Appropriate design parameters and conditions were modeled, as were the engineered safety features which can reduce the containment pressure. A minimum initial temperature associated with normal full-power operating conditions was modeled, and no coatings were credited on any of the containment structures.
The NRC staff finds that the licensee used the NRC-approved methodology for the Region II containment pressure calculation with appropriate design parameters and conditions. Utilizing a plant-specific minimum initial temperature associated with normal full-power operating conditions and taking no credit for the coatings on any of the containment structures should reduce containment pressure, as required by the limitations and conditions. Therefore, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 3.
Limitation and Condition 4 - Decay Heat Modeling The decay heat uncertainty multiplier will be ((
)). The analysis simulations for the FSLOCA EM will not be executed for longer than 10,000 seconds following reactor trip unless the decay heat model is appropriately justified. The sampled values of the decay heat uncertainty multiplier for the cases which produced the Region I and Region II analysis results will be provided in the analysis submittal in units of sigma and absolute units.
Consistent with the NRC-approved methodology, the decay heat uncertainty multiplier was
((
)) for the Braidwood, Units 1 and 2 and Byron, Units 1 and 2, analyses. The analysis simulations were all executed for less than 10,000 seconds following reactor trip. The sampled values of the decay heat uncertainty multiplier for the cases which produced the Region I and Region II analysis results have been provided in units of sigma and approximate absolute units in Tables 16 and 17 of Attachment 3 to the LAR.
The NRC staff finds that the licensee appropriately modeled decay heat per the limitation and condition and reported the resulting sampled values in units of sigma and absolute units for the limiting cases. Therefore, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 4.
Limitation and Condition 5 - Fuel Burnup Limits The maximum assembly and rod length-average burnup is limited to ((
)) respectively.
The maximum analyzed assembly and rod length-average burnup is less than or equal to
((
)) respectively, for Braidwood, Units 1 and 2 and Byron, Units 1 and 2.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Based on the above, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 5.
Limitation and Condition 6 - WCOBRA/TRAC-TF2 Interface with PAD 5.0 The fuel performance data for analyses with the FSLOCA EM should be based
[on the latest version of an NRC-approved fuel performance code, which is] the PAD5 code (at present), which includes the effect of thermal conductivity degradation. The nominal fuel pellet average temperatures and rod internal pressures should be the maximum values, and the generation of all the PAD5 fuel performance data should adhere to the NRC-approved PAD5 methodology.
PAD5 fuel performance data is utilized in the Braidwood, Units 1 and 2 and Byron, Units 1 and 2 analyses with the FSLOCA EM. The analyzed fuel pellet average temperatures bound the maximum values calculated in accordance with Section 7.5.1 of WCAP-17642-A, Revision 1 Westinghouse Performance Analysis and Design Model (PAD5), November 2017 (ADAMS Accession No. ML17338A396), and the analyzed rod internal pressures were calculated in accordance with Section 7.5.2 of WCAP-17642-A, Revision 1.
Given that the licensee used the latest NRC-approved fuel performance code (i.e., PAD5) and used appropriate conservative inputs, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 6.
Limitation and Condition 7 - lnterfacial Drag Uncertainty in Region I Analyses The YDRAG uncertainty parameter should be ((
))
Consistent with the NRC-approved methodology, the YDRAG uncertainty parameter was ((
)) for the Braidwood, Units 1 and 2 and Byron, Units 1 and 2, Region I analyses.
The NRC staff finds that the licensee appropriately used the specified interfacial drag uncertainty parameter as noted above and, therefore, meets the requirement of Limitation and Condition 7.
Limitation and Condition 8 - Biased Uncertainty Contributors in Region I Analyses The ((
))
Consistent with the NRC-approved methodology, the ((
)) for the Braidwood, Units 1 and 2 and Byron, Units 1 and 2, Region I analyses.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The NRC staff finds that the licensee appropriately used the specified biased uncertainty parameters as noted above and, therefore, meets the requirement of Limitation and Condition 8.
Limitation and Condition 9 - Effect of Bias in Applications for Region I For PWR designs which are not Westinghouse 3-loop PWRs, a sensitivity study will be executed to confirm that the ((
)) for the plant design being analyzed. This sensitivity study should be executed once, and then referenced in all applications to that particular plant class.
Braidwood, Units 1 and 2, and Byron, Units 1 and 2, are Westinghouse-designed 4-loop PWRs.
The requested sensitivity study was performed for a 4-loop Westinghouse-designed PWR and is discussed in Westinghouse letter No. LTR-NRC-18-50, Information to Satisfy the FULL SPECTRUM LOCA (FSLOCA) Evaluation Methodology Plant Type Limitations and Conditions for 4-loop Westinghouse Pressurized Water Reactors (PWRs), July 2018 (ADAMS Accession No. ML18198A041).
The NRC staff reviewed the attachment to Westinghouse letter No. LTR-NRC-18-50. This document describes the sensitivity studies done on the selected parameters and demonstrates that ((
)) Therefore, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 9.
Limitation and Condition 10 - Boundary Between Region I and Region II Breaks For PWR designs which are not Westinghouse 3-loop PWRs, a sensitivity study will be executed to: 1) demonstrate that no unexplained behavior occurs in the predicted safety criteria across the region boundary, and 2) ensure that the
((
)) must cover the equivalent 2 to 4-inch break range using RCS
[reactor coolant system]-volume scaling relative to the demonstration plant. This sensitivity study should be executed once, and then referenced in all applications to that particular plant class.
Additionally, the minimum sampled break area for the analysis of Region II should be 1 ft2 [square foot].
Braidwood, Units 1 and 2 and Byron, Units 1 and 2, are Westinghouse-designed 4-loop PWRs.
The requested sensitivity study was performed for a 4-loop Westinghouse-designed PWR and is discussed in Westinghouse letter No. LTR-NRC-18-50.
The minimum sampled break area for the Braidwood, Units 1 and 2 and Byron, Units 1 and 2, Region II analyses was 1 ft2.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The NRC staff reviewed the attachment to Westinghouse letter No. LTR-NRC-18-50. This document describes the sensitivities performed to demonstrate that the boundary between Region I and Region II breaks is appropriate for a 4-loop Westinghouse-designed plant.
The limitation and condition specifies that plants with larger RCS fluid volumes than the 3-loop plant test example in WCAP-16996-P-A, Revision 1, should cover the same 2-to-4 inch range using break area to RCS volume scaling to ensure that the break range is preserved and not artificially truncated. The licensee applied ((
))
In order to demonstrate that no unexplained behavior occurs in the predicted safety criteria across the region boundary, the analysis examined breaks ((
)) were not considered, the NRC staff finds that these would be bounded by the LBLOCA (Region II) similar to the IBLOCAs.
In addition, the Region II analysis considers a minimum break area of 1.0 ft2 consistent with the requirement in the limitation and condition.
Therefore, the NRC staff finds that the requirements of Limitation and Condition 10 are met as the licensee performed the necessary sensitivity study to determine the appropriate break size range for Region I and boundary between Region I and Region II.
Limitation and Condition 11 - ((
)) in Uncertainty Analyses for Region II and Documentation of Reanalysis Results for Region I and Region II There are various aspects of this Limitation and Condition, which are summarized below:
- 1.
The ((
))
the Region I and Region II analysis seeds, and the analysis inputs will be declared and documented prior to performing the Region I and Region II uncertainty analyses. The ((
)) and the Region I and Region II analyses seeds will not be changed throughout the remainder of the analysis once they have been declared and documented.
- 2.
If the analysis inputs are changed after they have been declared and documented, for the intended purpose of demonstrating compliance with
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION the applicable acceptance criteria, then the changes and associated rationale for the changes will be provided in the analysis submittal.
Additionally, the preliminary values for PCT, maximum local oxidation (MLO), and core-wide oxidation (CWO) which caused the input changes will be provided. These preliminary values are not subject to Appendix B verification, and archival of the supporting information for these preliminary values is not required.
- 3.
Plant operating ranges which are sampled within the uncertainty analysis will be provided in the analysis submittal for both regions.
This Limitation and Condition was met for the Braidwood, Units 1 and 2 and Byron, Units 1 and 2 analyses as follows:
- 1.
The ((
)) the Region I and Region II analysis seeds, and the analysis inputs were declared and documented prior to performing the Region I and Region II uncertainty analyses. The ((
))
and the Region I and Region II analyses seeds were not changed once they were declared and documented.
- 2.
The analysis inputs were not changed once they were declared and documented.
- 3.
The plant operating ranges which were sampled within the uncertainty analyses are provided for Braidwood, Units 1 and 2 and Byron, Units 1 and 2, in Table 1 of to the LAR. Note that these operating ranges are consistent between all units.
The licensee declared and documented the appropriate inputs and did not change these values once declared and documented; therefore, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 11.
Limitation and Condition 12 - Steam Generator Heat Removal During SBLOCAs The plant-specific dynamic pressure loss from the steam generator secondary-side to the main steam safety valves must be adequately accounted for in analysis with the FSLOCA EM.
A bounding plant-specific dynamic pressure loss from the steam generator secondary-side to the main steam safety valves was modeled in the Braidwood, Units 1 and 2 and Byron, Units 1 and 2.
The licensee used a bounding dynamic pressure loss from the steam generator secondary-side to the main steam safety valves; therefore, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 12.
Limitation and Condition 13 - Upper Head Spray Nozzle Loss Coefficient In plant-specific models for analysis with the FSLOCA EM: 1) the ((
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION
)) and 2) the ((
))
The NRC staff finds that the ((
)) in the analyses for Braidwood, Units 1 and 2 and Byron, Units 1 and 2. The NRC staff finds that the ((
)) in the analyses.
Based on the above, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 13.
Limitation and Condition 14 - Correlation for Oxidation For analyses with the FSLOCA EM to demonstrate compliance against the current 10 CFR 50.46 oxidation criterion, the transient time-at-temperature will be converted to an equivalent cladding reacted (ECR) using either the Baker-Just or the Cathcart-Pawel correlation. In either case, the pre-transient corrosion will be summed with the LOCA transient oxidation. If the Cathcart-Pawel correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 13 percent limit. If the Baker-Just correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 17 percent limit.
For the Braidwood, Units 1 and 2 and Byron, Units 1 and 2, analyses, the Baker-Just correlation was used to convert the LOCA transient time-at-temperature to an ECR. The resulting LOCA transient ECR was then summed with the preexisting corrosion for comparison against the 10 CFR 50.46 local oxidation acceptance criterion of 17 percent.
The NRC staff finds that by using the Baker-Just correlation, converting to an ECR, and accounting for preexisting corrosion, the licensee has met the requirements of Limitation and Condition 14.
Limitation and Condition 15 - LOOP versus OPA Treatment in Uncertainty Analyses for Region II The Region II analysis will be executed twice; once assuming loss-of-offsite power (LOOP) and once assuming offsite power available (OPA). The results from both analysis executions should be shown to be in compliance with the 10 CFR 50.46 acceptance criteria.
The ((
))
The Region II uncertainty analyses for Braidwood, Units 1 and 2 and Byron, Units 1 and 2, were performed twice; once assuming a LOOP and once assuming OPA. The results from both analyses that were performed are in compliance with the 10 CFR 50.46 acceptance criteria.
The LOOP configuration is the limiting configuration for Braidwood, Unit 1, and Byron, Unit 1, and the OPA configuration is the limiting configuration for Braidwood Unit 2, and Byron, Unit 2 (See Tables 8 and 9 of Attachment 3 to the LAR).
The ((
))
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Given that the licensee has performed the Region II analysis for both LOOP and OPA and that the results from both are in compliance with the acceptance criteria in 10 CFR 50.46(b)(1) through (b)(4), the NRC staff finds that the licensee has met the requirements of Limitation and Condition 15.
3.3 Results and Compliance with 10 CFR 50.46 The licensee presented the results for PCT, MLO, and CWO in Tables 8 and 9 of Attachment 3 to the LAR for Braidwood, Units 1 and 2 and Byron, Units 1 and 2, respectively. To demonstrate compliance with 10 CFR 50.46(b)(1) through (b)(4), the following criteria must be met:
- 1. PCT;
- 2. Maximum cladding oxidation;
- 3. Maximum hydrogen generation; and
- 4. Coolable geometry.
Each of the above four 10 CFR 50.46 criteria is discussed below.
Note that the FSLOCA EM does not address 10 CFR 50.46 (b)(5), Long-term cooling.
Long-term cooling is dependent on the demonstration of the continued delivery of cooling water to the core. The licensee stated that the actions that are currently in place to maintain long-term cooling are not impacted by the application of the NRC-approved FSLOCA EM.
Peak Cladding Temperature The requirement of 10 CFR 50.46(b)(1) states that The calculated maximum fuel element cladding temperature shall not exceed 2200 °F. The licensee stated that the analysis for PCT corresponds to a bounding estimate of the 95th percentile PCT at the 95-percent confidence level, and given that the resulting PCT is less than 2200 °F, the analyses with the FSLOCA EM confirm that 10 CFR 50.46 acceptance criterion (b)(1) is satisfied. The licensee presented the results in Tables 8 and 9 of Attachment 3 to the LAR for Braidwood, Unit 1 and Byron, Unit 1, and Braidwood, Unit 2, and Byron, Unit 2, respectively.
Given that the maximum calculated PCT is below the 2200 °F PCT limit, the NRC staff finds that the acceptance criterion of 10 CFR 50.46(b)(1) is met.
Maximum Cladding Oxidation The requirements of 10 CFR 50.46(b)(2) state, in part, that The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation. The licensee stated that the analysis for MLO corresponds to a bounding estimate of the 95th percentile MLO at the 95-percent confidence level. Since the resulting MLO is less than 17 percent when converting the time-at-temperature to an ECR using the Baker-Just correlation and adding the pretransient corrosion, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(2) is satisfied. The licensee presented the results in Tables 8 and 9 of to the LAR for Braidwood, Unit 1 and Byron, Unit 1, and Braidwood, Unit 2, and Byron, Unit 2, respectively.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Given that the resulting MLO is below the 17 percent limit, the NRC staff finds that the acceptance criterion of 10 CFR 50.46(b)(2) is met.
Maximum Hydrogen Generation The requirement of 10 CFR 50.46(b)(3) states that The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
The licensee stated that the analysis for CWO corresponds to a bounding estimate of the 95th percentile CWO at the 95-percent confidence level. Since the resulting CWO is less than 1 percent, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(3) is satisfied. The licensee presented the results in Tables 8 and 9 of Attachment 3 to the LAR for Braidwood, Unit 1, and Byron, Unit 1, and Braidwood, Unit 2, and Byron, Unit 2, respectively.
Given that the resulting CWO is below the 1 percent limit, the NRC staff finds that the acceptance criterion of 10 CFR 50.46(b)(3) is met.
Coolable Geometry The requirement of 10 CFR 50.46(b)(4) states that Calculated changes in core geometry shall be such that the core remains amenable to cooling. The licensee stated that this criterion is met by demonstrating compliance with criteria 10 CFR 50.46(b)(1), (b)(2), and (b)(3), and by ensuring that fuel assembly grid deformation due to combined LOCA and seismic loads is specifically addressed.
Section 32.1 of the NRC-approved FSLOCA EM documents that the effects of LOCA and seismic loads on the core geometry do not need to be considered unless fuel assembly grid deformation extends beyond the core periphery (i.e., deformation in a fuel assembly with no sides adjacent to the core baffle plates). Inboard grid deformation due to the combined LOCA and seismic loads was calculated to not occur for Braidwood, Units 1 and 2, and Byron, Units 1 and 2, based on performance of the fuel assembly structural analysis that, in turn, was based on the NRC-approved Licensing Topical Reports by Westinghouse, Safety Analysis of the Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident, October 1978 (ADAMS Accession No. ML19284A422), as referenced in the plants Updated Final Safety Analysis Report. The FSLOCA EM analyses did not invalidate the existing seismic/LOCA analysis.
Given that the criteria in 10 CFR 50.46(b)(1), (b)(2), and (b)(3) are met and that fuel assembly grid deformation due to combined LOCA and seismic loads is specifically addressed, the NRC staff finds that the acceptance criterion of 10 CFR 50.46(b)(4) is met.
3.4 Technical Evaluation Summary The licensee proposed to modify TS 5.6.5, Core Operating Limits Report (COLR), to replace the existing three NRC-approved LOCA methodologies listed in TS 5.6.5 b. 6., 7, and 8 with the NRC-approved LOCA methodology contained in WCAP-16996-P-A, Revision 1. The NRC staff concludes that the existing three methodologies are no longer needed to address the full spectrum of LOCA break sizes and the proposed TS change is acceptable as it changes from
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION one set of NRC-approved methods to another NRC-approved method, which addresses the full spectrum of LOCAs. The NRC staff's review has determined that the licensee appropriately applied the FSLOCA EM to Braidwood, Units 1 and 2 and Byron, Units 1 and 2, and finds that the resulting analysis meets the criteria in 10 CFR 50.46(b)(1) through (4).
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Illinois State official was notified of the proposed issuance of the amendments on October 28, 2020. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration in the Federal Register on May 5, 2020 (85 FR 26722), and there has been no public comment on such finding. 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 Commission's 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: M. Razzaque Date: December 28, 2020
ML20315A516 (proprietary); ML20317A001 (non-proprietary)
OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/SNSB/BC NRR/DSS/STSB/BC NAME JWiebe SRohrer (JBurkhardt for) SKrepel VCusumano DATE 11/16/2020 11/12/2020 9/30/2020 12/10/2020 OFFICE OGC - NLO NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME JMcManus NSalgado (RKuntz for)
JWiebe DATE 12/8/2020 12/18/2020 12/28/2020