ML23319A387
| ML23319A387 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 12/20/2023 |
| From: | Dennis Galvin NRC/NRR/DORL/LPL4 |
| To: | Peters K Comanche Peak Nuclear Power Co, Vistra Operations Company |
| Shared Package | |
| ML23319A374 | List: |
| References | |
| EPID L-2022-LLA-0171 | |
| Download: ML23319A387 (1) | |
Text
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION December 20, 2023 Mr. Ken J. Peters Senior Vice President and Chief Nuclear Officer Attention: Regulatory Affairs Vistra Operations Company LLC Comanche Peak Nuclear Power Plant 6322 N FM 56 P.O. Box 1002 Glen Rose, TX 76043
SUBJECT:
COMANCHE PEAK NUCLEAR POWER PLANT, UNIT NOS. 1 AND 2 -
ISSUANCE OF AMENDMENT NOS. 185 AND 185 REGARDING IMPLEMENTATION OF FULL SPECTRUM LOSS-OF-COOLANT ACCIDENT METHODOLOGY (EPID L-2022-LLA-0171)
Dear Mr. Peters:
The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 185 to Facility Operating License No. NPF-87 and Amendment No. 185 to Facility Operating License No. NPF-89 for Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2 (Comanche Peak, Units 1 and 2), respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated November 21, 2022, as supplemented by letter dated June 1, 2023.
The amendments revise TS 5.6.5, Core Operating Limits Report, to change the loss-of-coolant accident (LOCA) methodology to reflect the adoption of WCAP16996PA, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUMTM LOCA Methodology); TS Safety Limit (SL) 2.1.1.2 in Reactor Core SLs to reflect the peak fuel centerline melt temperature specified in the Westinghouse performance analysis and design model (PAD5); and TS 4.2.1, Fuel Assemblies, to remove the allowance for zircalloy cladding for Comanche Peak, Units 1 and 2.
The NRC staff has determined that the related safety evaluation (SE) contains proprietary information pursuant to Title 10 of the Code of Federal Regulations Section 2.390, Public inspections, exemptions, request for withholding. The proprietary information is indicated by bold text enclosed with ((double brackets)). The proprietary version of the SE is provided as enclosure 3. Accordingly, the NRC staff has also prepared a non-proprietary version of the SE, which is provided as enclosure 4.
to this letter contains proprietary information. When separated from Enclosure 3, this document is DECONTROLLED.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION A copy of the related SE is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Dennis J. Galvin, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-445 and 50-446
Enclosures:
- 1. Amendment No. 185 to NPF-87
- 2. Amendment No. 185 to NPF-89
- 3. Safety Evaluation (Proprietary)
- 4. Safety Evaluation (Non-Proprietary) cc w/o Enclosure 3: Listserv
COMANCHE PEAK POWER COMPANY LLC AND VISTRA OPERATIONS COMPANY LLC COMANCHE PEAK NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-445 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 185 License No. NPF-87
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Vistra Operations Company LLC (Vistra OpCo) dated November 21, 2022, as supplemented by letter dated June 1, 2023, 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 license 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 Facility Operating License No. NPF-87 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 185 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. Vistra OpCo shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance and shall be implemented prior to the Mode 4 entry for the Comanche Peak Unit 1 Cycle 25 (spring 2025).
FOR THE NUCLEAR REGULATORY COMMISSION
/RA - Jennifer Dixon-Herrity for/
Jennivine K. Rankin, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Facility Operating License and Technical Specifications Date of Issuance: December 20, 2023 COMANCHE PEAK POWER COMPANY LLC AND VISTRA OPERATIONS COMPANY LLC COMANCHE PEAK NUCLEAR POWER PLANT, UNIT NO. 2 DOCKET NO. 50-446 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 185 License No. NPF-89 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Vistra Operations Company LLC (Vistra OpCo) dated November 21, 2022, as supplemented by letter dated June 1, 2023, 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 license 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 Facility Operating License No. NPF-89 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 185 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. Vistra OpCo shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
The license amendment is effective as of its date of issuance and shall be implemented prior to the Mode 4 entry for the Comanche Peak Unit 2 Cycle 22 (fall 2024).
FOR THE NUCLEAR REGULATORY COMMISSION
/RA - Jennifer Dixon-Herrity for/
Jennivine K. Rankin, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Facility Operating License and Technical Specifications Date of Issuance: December 20, 2023
ATTACHMENT TO LICENSE AMENDMENT NO. 185 TO FACILITY OPERATING LICENSE NO. NPF-87 AND AMENDMENT NO. 185 TO FACILITY OPERATING LICENSE NO. NPF-89 COMANCHE PEAK NUCLEAR POWER PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-445 AND 50-446 Replace the following pages of Facility Operating License Nos. NPF-87 and NPF-89, and the 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.
Facility Operating License No. NPF-87 REMOVE INSERT 3
3 Facility Operating License No. NPF-89 REMOVE INSERT 3
3 Technical Specifications REMOVE INSERT 2.0-1 2.0-1 4.0-1 4.0-1 5.6-5 5.6-5
(3)
Vistra OpCo, 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, and described in the Final Safety Analysis Report, as supplemented and amended; (4)
Vistra OpCo, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use, at any time, any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)
Vistra OpCo, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required, any byproduct, source, and special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)
Vistra OpCo, 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 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 Vistra OpCo is authorized to operate the facility at reactor core power levels not in excess of 3458 megawatts thermal through Cycle 13 and 3612 megawatts thermal starting with Cycle 14 in accordance with the conditions specified herein.
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 185 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. Vistra OpCo shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
Unit 1 Amendment No. 185
(3)
Vistra OpCo, 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, and described in the Final Safety Analysis Report, as supplemented and amended; (4)
Vistra OpCo, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use, at any time, any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5)
Vistra OpCo, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required, any byproduct, source, and special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6)
Vistra OpCo, 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 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 Vistra OpCo is authorized to operate the facility at reactor core power levels not in excess of 3458 megawatts thermal through Cycle 11 and 3612 megawatts thermal starting with Cycle 12 in accordance with the conditions specified herein.
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 185 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. Vistra OpCo shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Antitrust Conditions DELETED Unit 2 Amendment No. 185
SLs 2.0 COMANCHE PEAK - UNITS 1 AND 2 2.0-1 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, the departure from nucleate boiling ratio (DNBR) shall be maintained the 95/95 DNB criterion for the DNB correlation(s) specified in Section 5.6.5.
2.1.1.2 In MODES 1 and 2, the peak fuel centerline temperature shall be maintained less than 5080°F, decreasing by 9°F per 10,000 MWD/MTU of burnup.
2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained 2735 psig.
2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
2.2.2 If SL 2.1.2 is violated:
2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.
Amendment No. 150, 185
Design Features 4.0 COMANCHE PEAK - UNITS 1 AND 2 4.0-1 4.0 DESIGN FEATURES 4.1 Site Location The site area is approximately 7,700 acres located in Somervell County in North Central Texas.
Squaw Creek Reservoir (SCR), established for station cooling, extends into Hood County. The site is situated along Squaw Creek, a tributary of the Paluxy River, which is a tributary of the Brazos River. The site is over 30 miles southwest of the nearest portion of Fort Worth and approximately 4.5 miles north-northwest of Glen Rose, the nearest community.
4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases.
A limited number of lead test assemblies that have not completed representative testing may be placed in non-limiting core regions.
4.2.2 Control Rod Assemblies The reactor core shall contain 53 control rod assemblies. The control material shall be silver-indium-cadmium as approved by the NRC.
Amendment No. 150, 185
Reporting Requirements 5.6 5.6 Reporting Requirements COMANCHE PEAK - UNITS 1 AND 2 5.6-4 5.6.5 Core Operating Limits Report (COLR) (continued) 11.
WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), November 2016.
12.
Not used.
13.
Not used.
14.
Not used.
15.
WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994.
16.
WCAP-17661-P-A, Revision 1, Improved RAOC and CAOC FQ Surveillance Technical Specifications, February 2019.
c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
d.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) a.
RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
1.
Specification 3.4.3, RCS Pressure and Temperature (P/T) Limits, and 2.
Specification 3.4.12, Low Temperature Overpressure Protection (LTOP) System.
b.
The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
Amendment No. 184, 185
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 185 TO FACILITY OPERATING LICENSE NO. NPF-87 AND AMENDMENT NO. 185 TO FACILITY OPERATING LICENSE NO. NPF-89 COMANCHE PEAK POWER COMPANY LLC AND VISTRA OPERATIONS COMPANY LLC COMANCHE PEAK NUCLEAR POWER PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-445 AND 50-446
1.0 INTRODUCTION
By letter dated November 21, 2022 (Reference 1), as supplemented by letter dated June 1, 2023 (Reference 2) to the U.S. Nuclear Regulatory Commission (NRC, the Commission), Vistra Operations Company LLC (Vistra OpCo, the licensee) submitted a license amendment request (LAR) for Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2 (Comanche Peak, Units 1 and 2). The requested amendment would revise the Comanche Peak Technical Specification (TS) 5.6.5, Core Operating Limits Report, by replacing four referenced reports with WCAP-16996-P-A, Revision 1, Realistic LOCA [Loss-of-Coolant Accident] Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology) (Reference 3), and revise TS Safety Limit (SL) 2.1.1.2 in Reactor Core SLs by changing the peak fuel centerline temperature to reflect that specified in Topical Report (TR)
WCAP-17642-P-A, Revision 1, Westinghouse Performance Analysis and Design Model (PAD5) (Reference 4). The proposed amendment also requests to revise TS 4.2.1, Fuel Assemblies, by removing the reference to zircalloy cladding.
The supplemental letter dated June 1, 2023, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register (FR) on March 7, 2023 (88 FR 14184).
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2.0 REGULATORY EVALUATION
2.1 Description of Proposed Technical Specification Changes 2.1.1 Proposed Change to TS SL 2.1.1.2 The Comanche Peak TS SL 2.1.1.2 will be revised as follows. Deletions are shown in strike-through.
In MODES 1 and 2, the peak fuel centerline temperature shall be maintained
< [less than] 4700°F [degrees Fahrenheit].In MODES 1 and 2, the peak fuel centerline temperature shall be maintained less than 5080°F, decreasing by 9°F per 10,000 MWD/MTU [megawatt day per metric ton of uranium] of burnup.
2.1.2 Proposed Change to Design Features TS 4.2.1 The licensee proposed to revise TS 4.2.1, to remove discussion of Zircaloy clad fuel. In addition, the licensee proposed to remove the lead test assembly allowance specifically for ZIRLOTM fuel rod cladding.
The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy or ZIRLOTM clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing or that contain Westinghouse ZIRLO' fuel rod cladding may be placed in non-limiting core regions.
2.1.3 Proposed Change to TS 5.6.5 The licensee proposed to revise TS 5.6.5, paragraph b to replace four existing NRC-approved LOCA methodologies with one new NRC-approved LOCA methodology contained in WCAP-16996-P-A, Revision 1. The new methodology would replace method 11. Methods 12 to 14 are deleted and will be designated as Not used.
- 11.
WCAP 10054 P A, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, August 1985.
WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology),
November 2016.
- 12.
WCAP 10054 P A, Addendum 2, Revision 1, Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model, July 1997.
Not Used.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION
- 13.
WCAP 10079 P A, NOTRUMP, A Nodal Transient Small Break and General Network Code, August 1985.
Not Used.
- 14.
WCAP 16009 P A, Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM), January 2005.
Not Used.
2.2 Description As described in chapter 1 of the Comanche Peak Updated Final Safety Analysis Report (UFSAR) (Reference 5), the Nuclear Steam Supply System (NSSS) for each unit consists of a Westinghouse Electric Company (Westinghouse) designed 4-loop pressurized water reactor (PWR) and supporting auxiliary systems. The rated thermal power output of each unit is 3612 megawatt thermal (MWt). Each unit has a large dry-ambient cylindrical steel lined reinforced concrete containment with a hemispherical dome.
2.3 Regulations and Guidance Documents Regulations The NRC staff applied the following regulations in its review of the proposed changes.
The regulations in Title 10 of the Code of Federal Regulations (10 CFR) 50.36(c) require that the TSs include items in the following specific categories: (1) SLs, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation (LCOs);
(3) surveillance requirements; (4) design features; and (5) administrative controls.
The proposed TS changes are related to the categories (1), (4), and (5) above since the licensee has proposed: (a) a change to the peak fuel centerline temperature safety limit, (b) a change to the reactor core fuel assemblies design feature, and (c) a change to the COLR reporting requirements administrative controls.
The regulation in 10 CFR 50.36(b) states, in part, that technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments.
The regulation in 10 CFR 50.46(a)(1)(i) states that an acceptable emergency core cooling system (ECCS) evaluation model (EM) must include sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor during LOCAs.
The regulations in 10 CFR 50.46(b) require, in part, that during a LOCA event, the following criteria are satisfied:
(1).
Peak cladding temperature. The calculated maximum fuel element cladding temperature shall not exceed 2200° F.
(2).
Maximum cladding oxidation. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION (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.
(5).
Long-term cooling. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.
The regulation in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, General Design Criterion (GDC) 10, Reactor design, states that [t]he reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.
The regulations in 10 CFR Part 50, Appendix K, ECCS Evaluation Models, Part II, Required Documentation, specify documentation requirements for the emergency core cooling performance EMs specified in 10 CFR 50.46(a)(1)(i).
Guidance Documents The NRC staff used the following documents to provide guidance on acceptable approaches to demonstrate that the above regulations are met.
NRC Regulatory Guide (RG) 1.157, Best-Estimate Calculations of Emergency Core Cooling System Performance, May 1989. (Reference 6)
NRC Regulatory Guide 1.203, Transient and Accident Analysis Methods, December 2005. (Reference 7)
NRC Information Notice (IN) 2011-21, Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation, December 13, 2011. (Reference 8)
NRC IN 1998-29, Predicted Increase in Fuel Rod Cladding Oxidation, dated August 3, 1998. (Reference 9)
NRC Generic Letter (GL) 1988-16, Removal of Cycle Specific Parameter Limits from Technical Specifications, dated October 4, 1988. (Reference 10)
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3.0 TECHNICAL EVALUATION
3.1 Evaluation of TS Changes 3.1.1 Evaluation of TS 5.6.5b Change In TS 5.6.5b, the proposed change replaces the four COLR references 11 to 14 with the WCAP-16996-P-A methodology, which includes analytical methods for determining the parameters FQ(Z), FH, and the K(Z) curve listed in TS 5.6.5a. GL 1988-16 states that it is acceptable for licensees to control these reactor physics parameter limits by specifying an NRC-approved calculation methodology. These parameter limits may be removed from the TS and placed in a cycle-specific COLR that is required to be submitted to the NRC every operating cycle or each time it is revised. The guidance in the GL 1988-16 will continue to be met because the proposed change in TS 5.6.5b identifies the NRC-approved methodologies that are used to determine the core operating limits. These changes are administrative in nature since the COLR references 11 through 14 of TS 5.6.5b (analytical methods topical reports) and are being replaced by WCAP-16996-P-A, Revision 1, as new reference 11. Identifying current references 12 through 14 as Not used does not impact any applicable requirements.
The NRC staff considers the proposed TS 5.6.5 change to be acceptable because the new methodology, WCAP-16996-P-A, is an NRC-approved methodology found acceptable for Comanche Peak as discussed in section 3.2 of this SE, and therefore, would continue to provide administrative controls consistent with 10 CFR 50.36(c)(5).
3.1.2 Evaluation of TS SL 2.1.1.2 Change The NRC staff evaluation of TS SL 2.1.1.2 is provided in section 3.4 of this SE.
3.1.3 Evaluation of Design Feature TS 4.2.1 Change The licensee proposed to remove discussion of Zircaloy clad fuel because there is no Zircaloy clad fuel loaded in Comanche Peak. The licensee stated that all fuel loaded in Comanche Peak has ZIRLO cladding. In addition, the insertion of Zircaloy clad fuel would require additional analysis and calculations because the FSLOCA EM analysis only considered ZIRLO cladding.
Furthermore, the licensee proposed to remove the lead test assembly allowance specifically for ZIRLOTM fuel rod cladding because all fuel loaded in Comanche Peak has ZIRLO fuel rod cladding. As a result, the allowance specifically for ZIRLO cladding lead test assemblies is no longer required.
The NRC staff finds the changes acceptable because the changes are consistent with the current plant design and the newly approved FSLOCA methodology or are editorial clarifications that do not substantively change TS requirements, and therefore would continue to meet the requirements of 10 CFR 50.36(c)(4).
3.2 FSLOCA Analysis The FSLOCA EM methodology divides the break spectrum into two regions, termed as Region I and Region II. The Region I analysis is for small break LOCAs (SBLOCAs) and the Region II analysis is for large break LOCAs (LBLOCAs). According to the FSLOCA EM, the SBLOCA and
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION LBLOCA analyses are independent, separable, and do not influence each other. The licensee provided analyses for both SBLOCA and LBLOCA in the LAR.
Methodology The current NRC-approved Westinghouse best-estimate methodology used for the Comanche Peak LBLOCA analysis described in WCAP-16009-P-A, Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM), January 2005 (Reference 11), is termed as Automated Statistical Treatment of Uncertainty Method (ASTRUM) EM. This methodology is applicable to Westinghouse designed (a) 3-loop and 4-loop PWRs with ECCS injection into the reactor coolant system (RCS) cold legs, and (b) 2-loop PWRs with upper plenum injection. The ASTRUM EM uses the WCOBRA/TRAC as the analysis code and is applicable only for the LBLOCA analysis with a minimum break size of 1.0 ft2 (square feet). The ASTRUM EM was developed consistent with the criteria in RG 1.157.
As stated in section 1.2.2 of the NRC SE on WCAP-16996-P-A, the FSLOCA EM is built on the ASTRUM EM by extending the applicability of the WCOBRA/TRAC code to include the full spectrum of break sizes postulated in the RCS cold leg. The break sizes considered include any size in which the break flow is beyond the capacity of the normal charging pumps up to, and including, a double ended guillotine rupture in the RCS cold leg with a break flow area equal to two times the pipe flow area. For the development of FSLOCA EM, the licensee used the guidance in RGs 1.157 and 1.203.
The FSLOCA EM methodology uses the WCOBRA/TRAC-TF2 code to analyze the RCS thermal-hydraulic response for full spectrum of break sizes. This methodology is applicable to Westinghouse 3-and 4-loop plants with cold leg ECCS injection. Since Comanche Peak, Units 1 and 2 are Westinghouse designed 4-loop plants with cold leg ECCS injection, the FSLOCA EM is therefore applicable.
For Comanche Peak Units 1 and 2, two separate analyses were carried out using the FSLOCA EM methodology. A single composite vessel model was developed to represent both units. To ensure conservative application of the FSLOCA EM methodology, the plant designs and operating parameters were assessed to create the composite model. Moreover, to account for the difference in steam generator types, separate loop models were used.
The FSLOCA EM meets the regulation in 10 CFR 50.46(a)(1)(i) with respect to an acceptable ECCS evaluation because it is a best-estimate methodology and realistically describes the behavior of the reactor during LOCAs.
The FSLOCA EM meets the regulation in 10 CFR Part 50, Appendix K, that specifies documentation requirements for the emergency core cooling performance EMs specified in 10 CFR 50.46.
NRC IN 2011-21 notified addressees of concerns on the impact of irradiation on fuel thermal conductivity and its potential to result in significantly higher predicted peak cladding temperature (PCT) in realistic ECCS EMs. The Comanche Peak current LOCA analysis does not explicitly consider the fuel pellet thermal conductivity degradation (TCD) and peaking factor burndown. In a letter dated August 14, 2019 (Reference 12), the licensee proposed to use the FSLOCA EM to meet its regulatory commitment by updating its licensing basis analysis to account for the TCD
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION of fuel pellets in the FSLOCA analyses. The FSLOCA EM explicitly accounts for the effects of fuel pellet TCD by using the fuel rod performance input data generated by the PAD5 code described in NRC-approved TR WCAP-17642-P-A. The fuel pellet thermal conductivity input to the WCOBRA/TRAC-TF2 code used in the FSLOCA EM accounts for the TCD of fuel pellets, and therefore addresses IN 2011-21 and satisfies the licensees regulatory commitment made in the letter dated August 14, 2019.
For the calculation of the transient containment minimum back pressure as a boundary condition at the break for the FSLOCA analysis, the licensee used the NRC-accepted Westinghouse COCO (Containment Pressure Analysis Code) methodology described in TRs WCAP-8339, Westinghouse Emergency Core Cooling System Evaluation Model - Summary, (Reference 13) and WCAP-8327, Containment Pressure Analysis Code (COCO),
(Reference 14) by integrating it into the WCOBRA/TRAC-TF2 code. The mass and energy (M&E) release calculated by the WCOBRA/TRAC-TF2 code at the end of each timestep in the LOCA transient is transferred as an input to the COCO code for calculating the containment back pressure as a boundary condition at the break. The Westinghouse COCO code is an NRC-accepted code for calculating the containment pressure response.
The NRC staff finds that the licensee (1) appropriately applied the proposed FSLOCA EM for the FSLOCA analysis in order to satisfy 10 CFR 50.46(b)(1) through (b)(4), and (2) used NRC-accepted Westinghouse COCO code for calculating the transient minimum containment back pressure at the break.
Errors Discovered in WCOBRA/TRAC-TF2 Code Westinghouse letters to NRC (i.e., LTR-NRC-18-30, dated July 18, 2018 (Reference 15);
LTR-NRC-19-6, dated February 7, 2019 (Reference 16); LTR-NRC-20-5, dated March 11, 2020 (Reference 17); LTR-NRC-21-5, dated March 4, 2021(Reference 18); LTR-NRC-22-8, dated February 2022 (Reference 19); and LTR-NRC-23-5, dated March 10, 2023 (Reference 20))
reported errors and some changes to be made in the FSLOCA EM. The NRC staff noted that these letters did not report the error that impacts the gamma energy redistribution multiplier identified in Virginia Electric and Power Company letter to NRC dated August 31, 2020 (Reference 21). In the supplement dated June 1, 2023, the licensee stated that:
During the time that elapsed between the plant-specific application of the FSLOCA EM to Comanche Peak Units 1 and 2 and the submittal of the LAR to the NRC, some changes to or errors have been discovered in the FSLOCA EM.
Each of these changes has been separately reported, via LTR-NRC-19-6, LTR-NRC-20-5 (ADAMS Accession No. ML20086F461), LTR-NRC-21-5 (ADAMS Accession No. ML21063A564), LTR-NRC-22-8 (ADAMS Accession Nos. ML22054A120 and ML22054A121), and LTR-NRC-23-5, pursuant to 10 CFR 50.46. Plant-specific PCT variations are not addressed in these letters, but are treated, as appropriate, on a plant-specific basis in accordance with the applicable sections of 10 CFR 50. All changes or errors have been assessed to have a negligible effect on the PCT, except for one, which is explicitly addressed in LAR, Attachment 6. Precedents for including non-zero PCT assessments in the LAR include ADAMS Accession No. ML20063L282, ADAMS Accession No. ML19266A657, and ADAMS Accession No. ML20244A336.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION In attachment 61 to the LAR, the licensee reported an error in the gamma redistribution multiplier on hot rod and hot assembly power (FGAMMA) applied in the FSLOCA EM resulting in an under-estimation of the hot rod and hot assembly power by up to 5 percent. This could result in the under-prediction of the PCT and potentially the oxidation results obtained by FSLOCA EM methodology.
SBLOCA (Region I) Analysis In attachment 3 to the LAR, section 1.3.1, Region I Analysis, the licensee described the following five phenomena that occur during the time periods from the initiation of an SBLOCA event: (a) Blowdown, (b) Natural Circulation, (c) Loop Seal Clearance, (d) Boil-Off, and (e) Core Recovery. The data used by the licensee in the SBLOCA event analysis is provided in the following tables in the LAR, attachment 3:
Table 1 - Plant Operating Range Analyzed and Key Parameters for Comanche Peak Unit 1 and Unit 2 Table 4 - Safety Injection Flow Used for Region I Calculation for Comanche Peak Unit 1 and Unit 2 Table 6 - Steam Generator Main Steam Safety Valve Parameters for Comanche Peak Unit 1 and Unit 2 Table 8A - Comanche Peak Unit 1 Sequence of Events for Region I Analysis PCT Transient Table 8B - Comanche Peak Unit 2 Sequence of Events for Region I Analysis PCT Transient Table 10A - Comanche Peak Unit 1 Sampled Value of Decay Heat Uncertainty Multiplier, DECAY_HT, for Region I and Region II Analysis Cases Table 10B - Comanche Peak Unit 2 Sampled Value of Decay Heat Uncertainty Multiplier, DECAY_HT, for Region I and Region II Analysis Cases The licensees analysis used the following assumptions:
Most limiting single failure of one ECCS train.
The control rod drop is modeled for the small breaks assuming a 2-second signal delay time and a 2.7-second rod drop time.
Reactor coolant pump trip is modeled coincident with reactor trip on the low pressurizer pressure setpoint for loss of offsite power (LOOP) transients.
1 The attachments to the LAR are listed in the transmittal letter; however, the attachments are not labeled as such in the LAR.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION When the low pressurizer pressure safety injection (SI) setpoint is reached, the SI is initiated into the RCS after a delay time that accounts for the emergency diesel generators start-up and filling of headers.
The licensee stated that the transient that produced the most limiting PCT result is a cold leg break with a break diameter of 3.7-inches for Unit 1 and 3.6-inches for Unit 2.
LBLOCA Analysis In attachment 3 to the LAR, section 1.4.1, Region II Analysis, the licensee described the phenomena occurring during a LBLOCA event by dividing the transient into the following phases: (a) Blowdown - Critical Heat Flux (CHF) Phase, (b) Blowdown - Upward Core Flow Phase, (c) Blowdown - Downward Core Flow Phase, (d) Refill Phase, and (e) Reflood Phase.
The data used by the licensee in the LBLOCA event analysis is provided in the following tables in the LAR, Attachment 3:
Table 1 - Plant Operating Range Analyzed and Key Parameters for Comanche Peak Unit 1 and Unit 2 Table 2 - Containment Data Used for Region II Calculation of Containment Pressure for Comanche Peak Unit 1 and Unit 2 Table 3 - Containment Heat Sink Data Used for Region II Calculation of Containment Pressure for Comanche Peak Unit 1 and Unit 2 Table 5 - Safety Injection Flow Used for Region II Calculation for Comanche Peak Unit 1 and Unit 2 Table 6 - Steam Generator Main Steam Safety Valve Parameters for Comanche Peak Unit 1 and Unit 2 Table 9A - Comanche Peak Unit 1 Sequence of Events for Region II Analysis PCT Transient Table 9B - Comanche Peak Unit 2 Sequence of Events for Region II Analysis PCT Transient Table 10A - Comanche Peak Unit 1 Sampled Value of Decay Heat Uncertainty Multiplier, DECAY_HT, for Region I and Region II Analysis Cases Table 10B - Comanche Peak Unit 2 Sampled Value of Decay Heat Uncertainty Multiplier, DECAY_HT, for Region I and Region II Analysis Cases In the supplement dated June 1, 2023, the licensee referred to section 30 of WCAP-16996-P-A for the method used for the LBLOCA uncertainty analysis. This method is based on Monte Carlo sampling of all uncertainty contributors which led to the results from which upper tolerance limits are derived for the PCT, maximum local oxidation (MLO) and core-wide oxidation (CWO).
In the supplement dated June 1, 2023, the licensee stated, in part, the following:
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION
((
))
The licensee performed the LBLOCA analysis for both offsite power available (OPA) and LOOP cases assuming the most limiting single failure using the FSLOCA EM as described in WCAP-16996-P-A, with the exception that the licensee made the changes and error corrections reported in Westinghouse letters: LTR-NRC-18-30, LTR-NRC-19-6, LTR-NRC-20-5, LTR-NRC-21-5, LTR-NRC-22-8, and LTR-NRC-23-5.
Results Tables 1A and 1B below show the licensees SBLOCA and LBLOCA analysis results for Comanche Peak Units 1 and 2 for the PCT, MLO, and CWO. In attachment 6 to the LAR, the licensee stated that the values in tables 7A and 7B of attachments 3 and 4 to the LAR should be replaced with re-calculated values presented in attachment 6. Therefore, the values presented in tables 1A and 1B are those reflected in tables 7A and 7B of attachment 6 to the LAR.
The maximum values of PCT, MLO, and CWO for Comanche Peak Unit 1 are 1,566°F (based on the LOOP and OPA cases from the LBLOCA analysis), 8.96 percent (based on the SBLOCA analysis), and 0.02 percent (based on the LOOP and OPA cases from the LBLOCA analysis),
respectively.
The maximum values of PCT, MLO, and CWO for Comanche Peak Unit 2 are 1,610°F (based on the LOOP case from the LBLOCA analysis), 8.82 percent (based on the LOOP and OPA cases from the LBLOCA analysis), and 0.04 percent (based on the LOOP and OPA cases from the LBLOCA analysis), respectively.
to the LAR, figures 22A and 22B show the containment pressure for the transient that produced the limiting PCT result for Comanche Peak, Units 1 and 2.
In a letter dated June 1, 2023, the licensee provided the following:
Figures 29A and 32A and figures 29B and 32B show PCT versus effective break for the LOOP and OPA configurations for Comanche Peak, Units 1 and 2, respectively. These graphs reflect the combined effect of the break size and break flow model uncertainties.
Figures 30A and 33A and figures 30B and 33B, which show the transient MLO (or transient equivalent cladding reacted (ECR)) versus PCT, is provided for the LOOP and OPA configurations for Comanche Peak, Units 1 and 2 uncertainty analyses, respectively.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Figures 31A and 34A and figures 31B and 34B, which show the CWO versus PCT, is provided for the LOOP and OPA configurations for Comanche Peak Units 1 and 2 uncertainty analyses, respectively.
Table 1A: Comanche Peak Unit 1 Analysis Results with the FSLOCA EM Results SBLOCA Analysis Value LBLOCA Analysis Value With LOOP LBLOCA Analysis Value With OPA Acceptance Criteria in 10 CFR 50.46 95/95 PCT (Note 1) 1,017+15 =
1,032°F 1,546+20 =
1,566°F 1,546+20 =
1,566°F (b)(1) 2200°F 95/95 MLO 8.96%
8.64%
8.64%
(b)(2) 17%
95/95 CWO 0.00%
0.02%
0.02%
(b)(3) 1%
Note:
- 1. The PCT presented in the table is the sum of the uncertainty analysis result plus the impact of the energy redistribution uncertainty error correction.
Table 1B: Comanche Peak Unit 2 Analysis Results with the FSLOCA EM Results SBLOCA Analysis Value LBLOCA Analysis Value With LOOP LBLOCA Analysis Value With OPA Acceptance Criteria in 10 CFR 50.46 95/95 PCT (Note 1) 1,113+3 =
1,116°F 1,579+31 =
1,610°F 1,569+31 =
1,600°F (b)(1) 2200°F 95/95 MLO 8.70%
8.82%
8.82%
(b)(2) 17%
95/95 CWO 0.00%
0.04%
0.04%
(b)(3) 1%
Note:
- 1. The PCT presented in the table is the sum of the uncertainty analysis result plus the impact of the energy redistribution uncertainty error correction.
NRC Staff Evaluation of the SBLOCA and LBLOCA Analysis and Results The NRC staff finds that the licensees SBLOCA and LBLOCA analysis is acceptable based on the following:
The assumptions and key inputs which include core parameters, RCS parameters, and containment parameters are consistent with the plant configuration and current licensing basis as described in the UFSAR.
The LBLOCA analysis performed by the licensee supports the peaking factor margin and burndown summarized in UFSAR tables 15.6-8, Peak Clad Temperature Including All Penalties and Benefits, Best-Estimate Large Break LOCA (BE LOCA) for Comanche Peak Unit 1, and 15.6-10, Peak Clad Temperature Including All Penalties and Benefits, Best-Estimate Large Break LOCA (BE LOCA) for Comanche Peak Unit 2.
The licensee provided the Comanche Peak, Units 1 and 2 operating range over which the uncertainty values are defined for LBLOCA analysis, which are consistent with UFSAR Tables 15.6-4, Plant Operating Range Allowed by the Best-Estimate Large
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Break LOCA Analysis for Comanche Peak Unit 1, and 15.6-5, Plant Operating Range Allowed by the Best-Estimate Large Break LOCA Analysis for Comanche Peak Unit 2.
The uncertainty analysis is consistent with the FSLOCA EM.
The licensee used the updated FSLOCA EM after making changes and correcting errors reported in Westinghouse 10 CFR 50.46 letters LTR-NRC-18-30, LTR-NRC-19-6, LTR-NRC-20-5, LTR-NRC-21-5, LTR-NRC-22-8, and LTR-NRC-23-5 to NRC.
For the calculation of a minimum containment back pressure as a boundary condition at the break, which is conservative for the PCT calculation, the licensee integrated the COCO code into the WCOBRA/TRAC-TF2 thermal-hydraulic code and used appropriate input parameters to include the effects of the installed pressure reducing systems (i.e.,
assuming all trains of containment spray system and fan cooler are in operation).
The licensee performed the LBLOCA analysis for both OPA and LOOP cases, assuming the most limiting single failure of one ECCS train.
For the LBLOCA uncertainty analysis, the licensee used a ((
)) (as mentioned in licensees evaluation of limitation and condition 15) simulations which is
((
)) to simultaneously bound the 95th percentile values of PCT, MLO, and CWO with a 95 percent confidence level as predicted by the statistical theory.
In attachment 6 to the LAR, the error identified on hot rod and hot assembly power (FGAMMA) used for the FSLOCA' EM, which results in an underestimation of the hot rod and hot assembly power by up to 5 percent, is extensively discussed and the licensee appropriately applied the FSLOCA EM for FSLOCA analysis after making changes and correcting errors.
The licensee calculated the MLO as the sum of pre-transient corrosion and transient oxidation consistent with the position in IN 1998-29.
The licensee has satisfied all limitations and conditions for the analysis as evaluated by the NRC staff in section 3.5 below.
3.3 Compliance with 10 CFR 50.46 The results in attachment 4 to the LAR, tables 7A and 7B for the PCT, MLO, and CWO show significant margins relative to the regulatory requirements given in 10 CFR 50.46(b)(1), (b)(2),
and (b)(3). Regarding compliance with 10 CFR 50.46(b)(4), the NRC staff finds that coolable core geometry is maintained because of the following:
The acceptance criteria 10 CFR 50.46(b)(1) through (b)(3) are satisfied, Based on section 32.1 of the FSLOCA EM, the effects of LOCA and seismic loads on the core geometry do not need to be considered unless the fuel assembly grid deformation extends beyond the core periphery (i.e., deformation in a fuel assembly with
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION no sides adjacent to the core baffle plates). In a letter dated June 1, 2023, the licensee stated:
The FSLOCA EM analysis does not affect the existing calculations that support the analysis of record related to combined LOCA and seismic loads, and the conclusion is retained from prior calculations and is credited in the current LOCA design basis analyses. That is, the previous calculations on grid deformation due to combined LOCA and seismic loads remain valid. As described in Section 4.2.2.2.4 of the Final Safety Analysis Report (FSAR) regarding the combined LOCA and seismic loads, A coolable geometry is, therefore, assured of the IFM
[intermediate flow mixing] grid elevation, as well as at the structural grid elevation.
As stated in UFSAR section 15.6.5.2.6:
10 CFR 50.46 acceptance criterion (b)(4) requires that the calculated changes in core geometry are such that the core remains amenable to cooling. This criterion has historically been satisfied by adherence to criteria (b)(1) and (b)(2), and by assuring that fuel deformation due to combined LOCA and seismic loads is specifically addressed. It has been demonstrated that the PCT and maximum cladding oxidation limits remain in effect for Best-Estimate LOCA applications. The approved methodology (WCAP-12945-P-A) specifies that effects of LOCA and seismic loads on core geometry do not need to be considered unless grid crushing extends beyond the 44 assemblies in the low-power channel.
This situation has not been calculated to occur for Comanche Peak Units 1 and 2. Therefore, acceptance criterion (b)(4) is satisfied.
For compliance with 10 CFR 50.46(b)(5), the licensee did not utilize the FSLOCA EM because it is not NRC-approved for the long-term analyses. Therefore, for continued compliance with 10 CFR 50.46(b)(5), the current licensing basis is maintained and is not affected by the implementation of the FSLOCA EM.
The NRC staff finds the proposed LBLOCA analysis results for Comanche Peak, Units 1 and 2 are in compliance with 10 CFR 50.46(b)(1) through (b)(4) and acceptable. For the LBLOCA long term cooling, the NRC staff finds it acceptable that the licensee proposes to maintain the current licensing basis, which was previously found to be acceptable by the NRC.
3.4 TS SL 2.1.1.2 Change Evaluation The licensee requests to change the Comanche Peak, Units 1 and 2 TS to reflect the peak fuel centerline temperature found in TR WCAP-17642-P-A. The current Comanche Peak, Units 1 and 2 TS SL 2.1.1.2 limits the peak fuel centerline temperature to less than 4700°F. For normal operation and anticipated operational occurrences, the reactor protection system is designed to ensure that the peak fuel centerline temperature does not exceed the fuel melt temperature criterion. The intent of this criterion is to avoid gross fuel melting.
The scope of the NRC staffs review is limited to the proposed change to the TS SL for peak fuel centerline temperature be maintained less than 5080°F, decreasing by 9°F per
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 10,000 MWD/MTU of burnup. The proposed limit reflects the limit in NRC approved TR WCAP-17642-P-A.
The empirically derived fuel centerline melt temperature described in TR WCAP-17642-P-A is based on fuel properties described in open literature. The description of the fuel properties can be found in:
1-Thermophysical Properties of MOX and UO2 Fuels Including the Effects of Irradiation, S.G. Popov; J.J. Carbajo; V.K. Ivanov; and G.L. Yoder, ORNL/TM-2000/3S1 (2000),
(Reference 22); and 2-A Review of the Thermophysical Properties of MOX and UO2 Fuels, J.J. Carbajo; G.L. Yoder; S.G. Popov; and V.K. Ivanov, Journal of Nuclear Materials, 299, 181 (2001)
(Reference 23)
The above-mentioned two references provide the data describing the fuel properties. The specific burnup dependence is provided in section 6.1.5, Melting Point, of TR WCAP-17642-P-A based on an assessment of these data. The NRC staff determined that this burnup dependence was acceptable as described in section 3.7.12, Pellet Overheating Melting, of the NRC staffs SE (included in WCAP-17642-P-A) for TR WCAP-17642-P-A.
As noted beneath the caption for figure 59, on page 92 of the NRC staff SE for TR WCAP-17642-P-A, the NRC staff determined that this melting limit is acceptable.
Therefore, the NRC staff finds the change in TS SL 2.1.1.2 acceptable because the burnup dependent fuel centerline melt temperature is based on inherent fuel properties and does not depend on any specific calculational methodology.
The NRC staff determined that its analysis and technical justification for the fuel centerline melting temperature in the SE for WCAP-17642-P-A is applicable to Comanche Peak, Units 1 and 2 because the properties of the fuel design in use at Comanche Peak Units 1 and 2 are consistent with the fuel property data used to generate the fuel centerline melting temperature.
Therefore, the NRC staff determined that the proposed SL is conservative for Comanche Peak, Units 1 and 2.
As stated, in part, by the licensee in the LAR, section 3.0, Technical Evaluation:
The peak fuel centerline temperature SL is independent of the PAD5 methodology. The current licensing basis safety analyses use the existing SL 2.1.1.2 for fuel melt as an acceptance criterion as required by the current methodology. Thus, Vistra OpCo will continue to meet the existing SL when using its current licensing basis safety analyses even with the implementation of the proposed SL. Since the existing SL for peak fuel centerline temperature is more restrictive than the proposed limit, the current licensing basis safety analyses remain conservative with respect to the proposed SL.
Based on the above, including the LAR information that supports the use of the empirically derived and less restrictive temperature, the NRC staff concludes that the proposed increase in the peak fuel centerline temperature TS SL is consistent with the 10 CFR 50.36(b) requirement that TSs be derived from the licensees safety analyses, as amended.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The licensees proposed TS change meets the requirements of GDC 10 because (1) the peak fuel centerline temperature is based on a conservative evaluation of test data that is applicable to the fuel design used at Comanche Peak, Units 1 and 2, and (2) the proposed limit will ensure that fuel melt is precluded during conditions of normal operations and under anticipated operational occurrences. The proposed TS change also meets 10 CFR 50.36(c)(1)(i)(A) because the revised TS limits an important variable that is necessary to reasonably protect the integrity of a physical barrier that guards against the uncontrolled release of radioactivity.
Therefore, the NRC staff concludes that the proposed change is acceptable.
3.5 Limitations and Conditions The NRC SE for WCAP-16996-P-A, table 22 provides a list of limitations and conditions required to be satisfied in order for the licensee to implement the NRC-approved FSLOCA EM.
The licensee summarized each limitation and condition in the LAR. The NRC staff reviewed the licensees summaries against the actual limitations and conditions documented in table 22 of the NRC SE and finds the summaries to be acceptable because the licensee appropriately described the specific requirement of each limitation and condition in its summary. The NRC staff evaluation of how each limitation and condition is satisfied for the Comanche Peak, Units 1 and 2 LBLOCA analysis is provided below.
3.5.1 Limitation and Condition 1: FSLOCATM EM Applicability with Regard to LOCA Transient Phases Summary:
The FSLOCA EM is not approved to demonstrate compliance with 10 CFR 50.46 acceptance criterion (b)(5) related to the long-term cooling.
Evaluation:
The licensee did not use the FSLOCA EM to show compliance with 10 CFR 50.46(b)(5) because it is not an NRC-approved methodology to analyze LBLOCA long term cooling.
Therefore, the NRC staff finds this limitation and condition is satisfied.
3.5.2 Limitation and Condition 2: FSLOCATM EM Applicability with Regard to Type of PWR Plants Summary:
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.
Evaluation:
Comanche Peak, Units 1 and 2 are Westinghouse-designed 4-loop PWRs with cold-side injection. The licensee used the NRC-approved FSLOCA EM, which included changes and error corrections reported to the NRC pursuant to 10 CFR 50.46, as described and justified in Westinghouse letters LTR-NRC-18-30, LTR-NRC-19-6, LTR-NRC-20-5, LTR-NRC-21-5, LTR-NRC-22-8, and LTR-NRC-23-5. Additionally, the licensee provided supplemental
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION information and analysis for evaluation of the Gamma Energy Redistribution Error for Comanche Peak, Units 1 and 2 in attachment 6 to the LAR. The NRC staff therefore finds this limitation and condition is satisfied.
3.5.3 Limitation and Condition 3: FSLOCATM EM Applicability for Containment Pressure Modeling Summary:
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 credit for containment coatings which are qualified and outside of the break zone-of-influence.
Evaluation:
The NRC staff reviewed the information provided in attachment 3 to the LAR and finds that the LOCA containment pressure response analysis is performed using the NRC-accepted Westinghouse COCO methodology. For this analysis, the licensee integrated the COCO code into the WCOBRA/TRAC-TF2 thermal-hydraulic code. The licensee stated that appropriate input parameters, including the effects of all the installed pressure reducing systems (i.e., assuming all trains of the containment spray system in operation), were used for a calculation of a conservatively low containment back pressure as a boundary condition at the break. A plant specific minimum initial temperature associated with normal full-power operating conditions was modeled, and conservatively no coatings were credited on any of the containment structures.
The NRC staff finds this limitation and condition is satisfied because (a) the licensee used an acceptable plant specific initial temperature for the containment pressure response, (b) the licensee used NRC-approved methodology for the LBLOCA containment pressure calculation using inputs that minimized the containment back pressure as a boundary condition to the break for a conservative PCT calculation, and (c) conservatively, the licensee did not credit any coatings on any of the containment structures.
3.5.4 Limitation and Condition 4: Decay Heat Modeling in FSLOCATM EM Applications Summary:
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.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Evaluation:
The NRC staff reviewed tables 10A and 10B in attachment 3 of the LAR and confirmed that for both Region I and II analysis the licensee used sampled values of the decay heat uncertainty multiplier (DECY_HT), which are absolute in units of [sigma] and are ((
)) The analysis simulations were all executed for less than 10,000 seconds following reactor trip.
The NRC staff finds that that the licensee appropriately modeled decay heat and correctly reported the resulting sampled values in units of and absolute units for the limiting cases.
Therefore, the NRC staff finds that the licensee has satisfied the requirements for this limitation and condition.
3.5.5 Limitation and Condition 5: Fuel Burnup Limits in FSLOCATM EM Applications Summary:
The maximum assembly and rod length-average burnup is limited to ((
)) respectively Evaluation:
In response to this limitation and condition, in attachment 3 to the LAR, the licensee confirmed that the maximum analyzed assembly and rod length-average burnups for Comanche Peak Units1 and 2 FSLOCA analysis were less than or equal to ((
)) respectively, which are the limits contained in the FSLOCA EM.
Based on the above, the NRC staff finds that the licensee has satisfied the requirement for this limitation and condition.
3.5.6 Limitation and Condition 6: WCOBRA/TRAC-TF2 Interface with PAD 5.0 in the FSLOCATM EM Summary:
The fuel performance data for analyses with the FSLOCA EM should be based on 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.
Evaluation:
The NRC staff reviewed attachment 3 to the LAR and confirmed that the licensee used PAD5 in the analysis along with the FSLOCA EM. PAD5 is the latest version of the fuel performance code, which explicitly models TCD and is benchmarked to high burnup data in TR WCAP-17642-P-A. The licensee stated that the FSLOCA EM considers the effects of fuel pellet TCD and other burnup-related effects by initializing to fuel rod performance data input generated by the PAD5 code. In the analysis, the fuel pellet average temperatures
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION conservatively bounded the maximum values calculated in accordance with section 7.5.1, Maximum Fuel Temperatures, of TR WCAP-17642-P-A. The analyzed rod internal pressures were calculated in accordance with section 7.5.2, Rod Internal Pressure, of TR WCAP-17642-P-A.
The NRC staff finds that the Comanche Peak, Units 1 and 2 analysis for FSLOCA satisfies the requirements of this limitation and condition because the licensee used PAD5, which is the latest version of the NRC-approved fuel performance code and explicitly includes the effect of TCD using conservative inputs.
3.5.7 Limitation and Condition 7: Interfacial Drag Uncertainty in FSLOCATM EM Region I Analyses Summary:
The YDRAG [bubbly flow drag multiplier] uncertainty parameter should be ((
))
Evaluation:
In response to this limitation and condition, in attachment 3 to the LAR, the licensee confirmed that the interfacial drag multiplier YDRAG was ((
)) established in the FSLOCA EM as described in WCAP-16996-P-A, section 29.1.5. The lower value of the YDRAG reduces the two-phase mixture producing a lesser swell and therefore results in the calculation of a conservative rod heat-up. The NRC staff finds this limitation and condition is satisfied.
3.5.8 Limitation and Condition 8: Biased Uncertainty Contributors in FSLOCATM EM Region I Analyses Summary:
The ((
))
Evaluation:
The licensee stated in attachment 3 to the LAR, that the ((
)) for the Comanche Peak Units 1 and 2 Region I analyses. The NRC staff finds that the limitation and condition is satisfied because in the Region I analysis, the biasing of (a) ((
)) is consistent with WCAP-16996-P, section 29.1.6, and (b) ((
)) is consistent with WCAP-16996-P, section 29.1.7.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.5.9 Limitation and Condition 9: Effect of Bias in FSLOCATM EM Applications for Region I Summary:
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.
Evaluation:
In attachment 3 to the LAR, the licensee stated that Comanche Peak, Units 1 and 2 are both Westinghouse-designed 4-loop PWRs. The requested sensitivity study was performed for a 4-loop Westinghouse-designed PWR and is discussed in Westinghouse letter LTR-NRC-18-50 (Reference 24). The NRC-staff finds this limitation and condition is satisfied.
3.5.10 Limitation and Condition 10: Boundary Between FSLOCATM EM Region I and Region II Breaks Summary:
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-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.
Evaluation:
The licensee stated in attachment 3 to the LAR, that Comanche Peak Units 1 and 2 are both Westinghouse-designed 4-loop PWRs. The requested sensitivity study was performed for a 4-loop Westinghouse-designed PWR and is discussed in Westinghouse letter, LTR-NRC-18-50.
The minimum sampled break area for the Comanche Peak Units 1 and 2 Region II analyses was 1 ft2. The NRC staff finds that this limitation and condition is satisfied.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.5.11 Limitation and Condition 11: ((
)) in FSLOCATM EM Uncertainty Analyses for Region II and Documentation of Reanalysis Results for Region I and Region II Summary:
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 analysis 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 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 peak cladding temperature (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.
Evaluation:
In response to this limitation and condition, in attachment 3 to the LAR, the licensee confirmed that in the Comanche Peak, Units 1 and 2 models for analysis with the FSLOCA EM 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 and were not changed after they were declared and documented. The licensee provided the Comanche Peak, Units 1 and 2 plant operating ranges, which were sampled within the uncertainty analysis in attachment 3 to the LAR, table 1. Based on this information, the NRC staff finds this limitation and condition is satisfied.
3.5.12 Limitation and Condition 12: Steam Generator Heat Removal During SBLOCAs Summary:
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.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Evaluation:
In response to this limitation and condition, in attachment 3 to the LAR, the licensee confirmed that in the Comanche Peak, Units1 and 2 models for analysis with the FSLOCA EM, a conservatively high plant-specific dynamic pressure loss from the steam generator secondary side to the main steam safety valves was modeled. The NRC staff finds this limitation and condition is satisfied.
3.5.13 Limitation and Condition 13: Upper Head Spray Nozzle Loss Coefficient Summary:
In plant-specific models for analysis with the FSLOCA EM, specific modeling considerations for the upper head spray nozzles should be followed as required by the NRC-approved methodology.
In plant-specific models for analysis with the FSLOCA EM: 1) the ((
)) and 2) the ((
))
Evaluation:
In response to this limitation and condition in attachment 3 to the LAR, the licensee confirmed that in the Comanche Peak, Units 1 and 2 models for analysis with the FSLOCA EM, the
((
)) and the ((
)) The NRC staff finds this limitation and condition is satisfied.
3.5.14 Limitation and Condition 14: Correlation for Oxidation Summary:
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.
Evaluation:
In response to this limitation and condition, in attachment 3 to the LAR, the licensee stated that in the Comanche Peak, Units 1 and 2 models for analysis with the FSLOCA EM, the Baker-Just correlation was used to convert the LOCA transient time-at-temperature into an ECR. The licensee met the 10 CFR 50.46(b)(2) MLO acceptance criterion of 17 percent after summing up the pre-existing corrosion with the resulting LOCA transient ECR. Therefore, the NRC staff finds this limitation and condition is satisfied.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.5.15 Limitation and Condition 15: LOOP versus OPA Treatment in FSLOCATM EM Uncertainty Analyses for Region II Summary:
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 ((
))
Evaluation:
The NRC staff review of attachment 3 to the LAR, confirmed that the Region II analysis for Comanche Peak, Units 1 and 2 was performed twice; once assuming a LOOP and once assuming OPA. The licensee performed the statistical analysis for both cases in accordance with the FSLOCA EM. The results from both analyses are in compliance with the 10 CFR 50.46 acceptance criteria as evaluated in section 1.5 of the LAR. The licensee used a ((
)) Based on the above evaluation, the NRC staff finds this limitation and condition is satisfied.
3.6 Technical Evaluation Summary Based on the evaluations of the proposed change, the NRC staff concludes the following:
The 10 CFR 50.36(c)(5) requirement is satisfied because the licensee added the approved FSLOCA EM Methodology (WCAP-16996-P-A) to the TS 5.6.5b reference list as a provision for administrative controls.
The deletion of the obsolete COLR references in TS 5.6.5b is acceptable because they will no longer be used.
The guidance in GL 1988-16 continues to be implemented because the proposed change specifies the NRC-approved methodology for the determination of core operating limits.
The FSLOCA EM is NRC-approved and satisfies the requirements of 10 CFR 50.46(a)(1)(i),
and 10 CFR Part 50, Appendix K, Part II documentation requirements.
The licensee appropriately applied the FSLOCA EM for FSLOCA analysis after making changes and correcting errors reported in Westinghouse 10 CFR 50.46 letters LTR-NRC-18-30, LTR-NRC-19-6, LTR-NRC-20-5, LTR-NRC-21-5, LTR-NRC-22-8, and LTR-NRC-23-5 and evaluating the gamma energy redistribution error (The gamma energy redistribution error is summarized in attachment 6 of the LAR).
The FSLOCA analysis results satisfies the requirements of 10 CFR 50.46(b)(1) through (b)(4).
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION FSLOCA EM is not used to confirm compliance with 10 CFR 50.46(b)(5) because it is not NRC-approved for LOCA long term core cooling analysis.
The current compliance with 10 CFR 50.46(b)(5) is not affected by the FSLOCA EM implementation, so the current licensing basis remains as previously approved by the NRC.
IN 2011-21 on TCD for fuel pellets is appropriately considered while confirming compliance with 10 CFR 50.46(b)(1).
IN 1998-29 regarding predicted increase in fuel rod cladding oxidation is appropriately considered while confirming compliance with 10 CFR 50.46(b)(2) on the MLO.
All limitations and conditions listed in the NRC SE for WCAP-16996-P-A for the FSLOCA analysis are satisfied.
Based on the above conclusions, the NRC staff finds the proposed changes acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Texas State official was notified of the proposed issuance of the amendments on November 13, 2023. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration published in the Federal Register on March 7, 2023 (88 FR 14184), 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
Based on the considerations discussed above, the NRC staff concludes 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.
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION
7.0 REFERENCES
- 1.
Sewell, S. K., Vistra OpCo, letter to NRC, Application to Revise Technical Specifications to Apply the Westinghouse Full Spectrum Loss of Coolant Accident Evaluation Model LAR 22-002, dated November 21, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22325A324).
- 2.
Lloyd, J. J., Vistra OpCo, letter to NRC, Response to Request for Information for Application to Revise Technical Specifications to Apply the Westinghouse Full Spectrum Loss of Coolant Accident Evaluation Model LAR 22-002, dated June 1, 2023 (ML23152A266).
- 3.
Gresham, J. A. Westinghouse Electric Company, letter to NRC, Submittal of WCAP-16996-P-A/WCAP-16996-NP-A, Volumes I, II, III and Appendices, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUMTM LOCA Methodology) (TAC No. ME5244) (Proprietary/Non-Proprietary), dated October 2, 2017 (Package ML17277A130).
- 4.
Gresham, J. A., Westinghouse, letter to NRC, Submittal of WCAP-17642-P-A / WCAP-17642-NP-A, Revision 1 Westinghouse Performance Analysis and Design Model (PAD5), (Proprietary / Non-Proprietary), dated November 27, 2017 (Package ML17335A334).
- 5.
Vistra OpCo, Comanche Peak Nuclear Power Plant Units 1 and 2, FSAR Certified Amendment 111, Text and Tables - Redacted, dated February 2022 (ML22277A825).
- 6.
NRC, Best Estimate Calculations of Emergency Core Cooling System Performance, RG 1.157, dated May 1989 (ML003739584).
- 7.
NRC, Transient and Accident Analysis Methods, RG 1.203, dated December 2005 (ML053500170).
- 8.
NRC, Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation, NRC IN 2011-21, dated December 13, 2011 (ML113430785).
- 9.
NRC, Predicted Increase in Fuel Rod Cladding Oxidation, NRC IN 1998-29, dated August 3, 1998 (ML031050107).
- 10.
NRC, Removal of Cycle Specific Parameter Limits from Technical Specifications, GL 1988-16, dated October 4, 1988 (ML031200485).
- 11.
Westinghouse, Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM), WCAP-16009-P-A, dated January 2005 (ML050910159 and ML050910161).
- 12.
McCool, T. P., Vistra OpCo, letter to the NRC, ECCS Reanalysis Schedule, dated August 14, 2019 (ML19248B765).
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION
- 13.
Westinghouse, Westinghouse Emergency Core Cooling System Evaluation Model -
Summary, WCAP-8339, datedJune 1974 (ML092430562; not publicly available, proprietary information).
- 14.
Westinghouse, Containment Pressure Analysis Code (COCO), WCAP-8327, dated July 1974 (ML092460709; not publicly available, proprietary information).
- 15.
Gresham, J. A., Westinghouse, letter to J. Whitman, NRC, LTR-NRC-18-30, U.S.
Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2017, dated July 18, 2018 (ML19288A174).
- 16.
Hosack, K. L., Westinghouse, letter to NRC, LTR-NRC-19-6, U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2018, dated February 7, 2019 (Package ML19042A378).
- 17.
Hosack, K. L, Westinghouse, letter to NRC, LTR-NRC-20-5, U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2019, dated March 11, 2020 (Package ML20086F461)
- 18.
Harper, Z. S., Westinghouse, letter to NRC, LTR-NRC-21-5, U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2020, dated March 4, 2021 (ML21063A564)
- 19.
Westinghouse letter to NRC, LTR-NRC-22-8 U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2021, dated February 2022.
- 20.
Harper, Z. S., Westinghouse, letter to NRC, LTR-NRC-23-5, U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2022, dated March 10, 2023 (ML23072A071).
- 21.
Bischof, G. T., Virginia Electric and Power Company, letter to NRC, Virginia Electric and Power Company North Anna Power Station Units 1 and 2 Proposed License Amendment Request Addition of Analytical Methodology to the Core Operating Limits Report for a Full Spectrum Loss of Coolant Accident (FSLOCA) Gamma Energy Redistribution Information, dated August 31, 2020 (ML20244A336).
- 22.
Thermophysical Properties of MOX [Mixed Oxide] and UO2 Fuels Including the Effects of Irradiation, S.G. Popov; J.J. Carbajo; V.K. Ivanov; and G.L. Yoder, ORNL/TM-2000/3S1 (2000).
- 23.
A Review of the Thermophysical Properties of MOX and UO2 Fuels, J.J. Carbajo; G.L.
Yoder; S.G. Popov; and V.K. Ivanov, Journal of Nuclear Materials, 299, 181 (2001).
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION
- 24.
Westinghouse, Information to Satisfy the FULL SPECTRUM LOCA (FSLOCA)
Evaluation Methodology Plant Type Limitations and Conditions for 4-loop Westinghouse Pressurized Water Reactors (PWRs) (Proprietary/Non-Proprietary), LTR-NRC-18-50, dated July 2018 (ML18198A039 (letter) and ML18198A041 (redacted)).
Principal Contributors: Noushin Amini Clinton Ashley Ahsan Sallman Date: December 20, 2023
Package: ML23319A374 Proprietary: ML23319A392 Non-Proprietary: ML23319A387
- via email NRR-058 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA*
NRR/DSS/SNSB/BC*
NAME DGalvin PBlechman PSahd DATE 11/20/2023 11/20/2023 8/2/2023 OFFICE NRR/DSS/STSB/BC*
OGC NRR/DORL/LPL4/BC*
NAME VCusumano BVaisey JRankin (JDixon-Herrity for)
DATE 7/27/2023 12/6/2023 12/20/2023 OFFICE NRR/DORL/LPL4/PM*
NAME DGalvin DATE 12/20/2023