ML25213A183
| ML25213A183 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 08/27/2025 |
| From: | William Orders NRC/NRR/DORL/LPL4 |
| To: | Peters K Vistra Operations Company |
| Lee, Samson | |
| References | |
| EPID L-2024-LLA-0123 TSTF-51, TSTF-471-A, TSTF-571 | |
| Download: ML25213A183 (1) | |
Text
August 27, 2025 Mr. Ken J. Peters Executive 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. 193 AND 193, RESPECTIVELY, RE:
CHANGE TO TECHNICAL SPECIFICATIONS CONSISTENT WITH TSTF-51-A, TSTF-471-A AND TSTF-571 (EPID L-2024-LLA-0123)
Dear Mr. Peters:
The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued enclosed Amendment No. 193 to Renewed Facility Operating License No. NPF-87 and Amendment No. 193 to Renewed Facility Operating License No. NPF-89 for Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2 (Comanche Peak), respectively. The amendments consist of changes to the technical specifications (TSs) in response to your application dated September 12, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24256A088).
The amendments modify the Comanche Peak TSs to eliminate use of the defined term Core Alterations and revise requirements during handling of irradiated fuel. The amendments are consistent with NRC-approved Technical Specifications Task Force (TSTF) travelers TSTF-51-A, Revision 2, Revise containment requirements during handling irradiated fuel and core alterations; TSTF-471-A, Revision 1,Eliminate use of term CORE ALTERATIONS in ACTIONS and Notes; and TSTF-571-T, Revision 0, Revise Actions for Inoperable Source Range Neutron Flux Monitor.
K. Peters A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
William Orders, 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. 193 to NPF-87
- 2. Amendment No. 193 to NPF-89
- 3. Safety Evaluation cc: 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 RENEWED FACILITY OPERATING LICENSE Amendment No. 193 Renewed 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 September 12, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-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. 193 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this renewed 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 within 180 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Tony Nakanishi, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: August 27, 2025 TONY NAKANISHI Digitally signed by TONY NAKANISHI Date: 2025.08.27 09:54:20 -04'00'
COMANCHE PEAK POWER COMPANY LLC AND VISTRA OPERATIONS COMPANY LLC COMANCHE PEAK NUCLEAR POWER PLANT, UNIT NO. 2 DOCKET NO. 50-446 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 193 Renewed 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 September 12, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-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. 193 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this renewed license. Vistra OpCo shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 180 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Tony Nakanishi, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: August 27, 2025 TONY NAKANISHI Digitally signed by TONY NAKANISHI Date: 2025.08.27 09:54:46 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO.193 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-87 AND AMENDMENT NO.193 TO RENEWED 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 Renewed 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.
Renewed Facility Operating License No. NPF-87 REMOVE INSERT Renewed Facility Operating License No. NPF-89 REMOVE INSERT Technical Specifications REMOVE INSERT 1.1-2 1.1-2 3.3-53 3.3-53 3.3-55 3.3-55 3.3-57 3.3-57 3.7-26 3.7-26 3.7-28 3.7-28 3.7-29 3.7-29 3.8-19 3.8-19 3.8-20 3.8-20 3.8-27 3.8-27 3.8-36 3.8-36 3.8-40 3.8-40 3.8-41 3.8-41 3.9-1 3.9-1 3.9-2 3.9-2 3.9-4 3.9-4 3.9-6 3.9-6
(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 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 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. 193 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this renewed license. Vistra OpCo shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3) Antitrust Conditions DELETED (4) License Transfer The CP PowerCo Decommissioning Master Trust Agreement for the facility at the time the license transfers are effected and thereafter, is subject to the following:
Unit 1 Amendment No. 193
(2) CP PowerCo, pursuant to 10 CFR Part 50, to possess the facility at the designated location in Somervell County, Texas in accordance with the procedures and limitations set forth in this renewed license; (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 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 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. 193 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this renewed license. Vistra OpCo shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3) Antitrust Conditions DELETED (4) License Transfer Unit 2 Amendment No. 193
Definitions 1.1 1.1 Definitions (continued)
COMANCHE PEAK - UNITS 1 AND 2 1.1-2 CHANNEL OPERATIONAL TEST (COT)
CORE OPERATING LIMITS REPORT (COLR)
DOSE EQUIVALENT I-131 A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY so that the setpoints are within the necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.
DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes I-131, I-132, I-133, I-134, and I-135 actually present. The determination of DOSE EQUIVALENT I-131 shall be performed using thyroid dose conversion factors from Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites," or from Table E-7 of Regulatory Guide 1.109, Revision 1, NRC, 1977, or from ICRP-30, 1979, Supplement to Part 1, page 192-212, Table titled Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity, or from Table 2.1 of EPA Federal Guidance Report No. 11, 1988, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion.
Amendment No. 150,
Containment Ventilation Isolation Instrumentation 3.3.6 ACTIONS (continued)
COMANCHE PEAK - UNITS 1 AND 2 3.3-53 SURVEILLANCE REQUIREMENTS
NOTE--------------------------------------------------------------
Refer to Table 3.3.6-1 to determine which SRs apply for each Containment Ventilation Isolation Function.
CONDITION REQUIRED ACTION COMPLETION TIME C. --------------NOTE-------------
Only applicable during movement ofUHFHQWO\
irradiated fuelassemblies withincontainment.
Required Action and associated Completion Time for Condition A not met.
NOTE------------------------
The containment pressure relief valves may be opened in compliance with the gaseous effluent monitoring instrumentation requirements in Part I of the ODCM.
C.1 Place and maintain containment ventilation valves in closed position.
OR C.2 Enter applicable Conditions and Required Actions of LCO 3.9.4, "Containment Penetrations," for containment ventilation isolation valves made inoperable by isolation instrumentation.
Immediately Immediately SURVEILLANCE FREQUENCY SR 3.3.6.1 Perform CHANNEL CHECK.
In accordance with the Surveillance Frequency Control Program.
Amendment No. 156,
Containment Ventilation Isolation Instrumentation 3.3.6 COMANCHE PEAK - UNITS 1 AND 2 3.3-55 Table 3.3.6-1 (page 1 of 1)
Containment Ventilation Isolation Instrumentation
D Must satisfy Gaseous Effluent Dose Rate Requirements in Part I of the ODCM.
E During movement of UHFHQWO\irradiated fuel assemblies within containment.
FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS SURVEILLANCE REQUIREMENTS TRIP SETPOINT
- 1. Manual Initiation 1, 2, 3, 4 Refer to LCO 3.3.2 ESFAS Instrumentation, Functions 2.a and 3.a.1, respectively for all initiation functions and requirements.
- 2. Automatic Actuation Logic and Actuation Relays 1, 2, 3, 4 2 trains SR 3.3.6.2 SR 3.3.6.3 SR 3.3.6.5 NA
- 3. Containment Radiation
- a. Gaseous 1, 2, 3, 4, (b) 1 SR 3.3.6.1 SR 3.3.6.4 SR 3.3.6.7 (a)
- 4. Containment Isolation -
Phase A Refer to LCO 3.3.2, ESFAS Instrumentation, Function 3.a, for all initiation functions and requirements.
Amendment No. 156,
CREFS Actuation Instrumentation 3.3.7 ACTIONS (continued)
COMANCHE PEAK - UNITS 1 AND 2 3.3-57 CONDITION REQUIRED ACTION COMPLETION TIME B. One or more Functions with two channels or two trains inoperable.
B.1.1 Place one CREFS train in emergency recirculation mode.
AND B.1.2 Enter applicable Conditions and Required Actions for one CREFS train made inoperable by inoperable CREFS actuation instrumentation OR B.2 -----------------NOTE------------------
Applicable only to Functions 3a and 3b.
Secure the Control Room makeup air supply fan from the affected air intake.
Immediately Immediately Immediately C. Required Action and associated Completion Time for Condition A or B not met in MODE 1, 2, 3, or 4.
C.1 Be in MODE 3.
AND C.2 Be in MODE 5.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours D. Required Action and associated Completion Time for Condition A or B not met in MODE 5 or 6, or during movement of irradiated fuel assemblies.
D. Suspend movement of irradiated fuel assemblies.
Immediately Amendment No.
CREFS 3.7.10 ACTIONS (continued)
COMANCHE PEAK - UNITS 1 AND 2 3.7-26 CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and associated Completion Time of Condition A or B not met in MODE 1, 2, 3, or 4.
C.1 Be in MODE 3.
AND C.2 Be in MODE 5.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours D. Required Action and associated Completion Time of Condition A not met in MODE 5 or 6, or during movement of irradiated fuel assemblies.
D.1 Place OPERABLE CREFS train in emergency recirculation mode.
OR D.2 Suspend movement of irradiated fuel assemblies.
Immediately Immediately E. Two CREFS trains inoperable in MODE 5 or 6, or during movement of irradiated fuel assemblies.
OR One or more CREFS trains inoperable due to an inoperable CRE boundary in MODE 5 or 6, or during movement of irradiated fuel assemblies.
E. Suspend movement of irradiated fuel assemblies.
Immediately F. Two CREFS trains inoperable in MODE 1, 2, 3, or 4 for reasons other than Condition B.
F.1 Enter LCO 3.0.3.
Immediately Amendment No. 150,
CRACS 3.7.11 3.7 PLANT SYSTEMS COMANCHE PEAK - UNITS 1 AND 2 3.7-28 3.7.11 Control Room Air Conditioning System (CRACS)
LCO 3.7.11 Two CRACS trains shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, 3, 4, 5, and 6, During movement of irradiated fuel assemblies.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One CRACS train inoperable.
A.1 Restore CRACS train to OPERABLE status.
30 days B. Required Action and associated Completion Time of Condition A not met in MODE 1, 2, 3, or 4.
B.1 Be in MODE 3.
AND B.2 Be in MODE 5.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours C. Required Action and associated Completion Time of Condition A not met in MODE 5, or 6, or during movement of irradiated fuel assemblies.
C.1 Place OPERABLE CRACS train in operation.
OR C.2 Suspend movement of irradiated fuel assemblies.
Immediately Immediately Amendment No. 150,
CRACS 3.7.11 ACTIONS (continued)
COMANCHE PEAK - UNITS 1 AND 2 3.7-29 CONDITION REQUIRED ACTION COMPLETION TIME D. Two CRACS trains inoperable in MODE 5 or 6, or during movement of irradiated fuel assemblies.
D.1.1 Verify at least 100% of the required heat removal capability equivalent to a single OPERABLE train available.
AND D.1.2 Restore the CRACS trains to OPERABLE status.
OR D.2 Suspend movement of irradiated fuel assemblies.
Immediately 30 days Immediately E. Two CRACS trains inoperable in MODE 1, 2, 3, or 4.
E.1.1 Verify at least 100% of the required heat removal capability equivalent to a single OPERABLE train available.
AND E.1.2 Restore one CRACS train to OPERABLE status.
Immediately 30 days Immediately Amendment No. 150,
AC Sources - Shutdown 3.8.2 COMANCHE PEAK - UNITS 1 AND 2 3.8-19 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite circuit inoperable.
NOTE-----------------------
Enter applicable Conditions and Required Actions of LCO 3.8.10, with the required train de-energized as a result of Condition A.
A.1 Declare affected required feature(s) with no offsite power available inoperable.
OR A.2. Suspend movement of irradiated fuel assemblies.
AND A.2. Suspend operations involving positive reactivity additions that could result in loss of required SDM or boron concentration.
AND A.2. Initiate action to restore required offsite power circuit to OPERABLE status.
Immediately Immediately Immediately Immediately Amendment No. 150, 183
AC Sources - Shutdown 3.8.2 ACTIONS (continued)
COMANCHE PEAK - UNITS 1 AND 2 3.8-20 CONDITION REQUIRED ACTION COMPLETION TIME B. One required DG inoperable.
B. Suspend movement of irradiated fuel
assemblies.
AND B. Suspend operations involving positive reactivity additions that could result in loss of required SDM or boron concentration.
AND B. Initiate action to restore required DG to OPERABLE status.
Immediately Immediately Immediately Amendment No. 153,
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.2.1 SR 3.8.1.1 SR 3.8.1.2 SR 3.8.1.3 SR 3.8.1.
SR 3.8.1.4 SR 3.8.1.
SR 3.8.1.5 SR 3.8.1.6 SR 3.8.1.1
SR 3.8.1.1
In accordance with applicable SRs 7KHIROORZLQJ65VDUHDSSOLFDEOHIRUAC sources required to be OPERABLE:
NOTE-----------------------------------------------
The following SRs are not required to be performed: SR 3.8.1.3, SR 3.8.1.9
SR 3.8.1.1, SR3.8.1.14, and SR 3.8.1.16.
DC Sources - Shutdown 3.8.5 3.8 ELECTRICAL POWER SYSTEMS COMANCHE PEAK - UNITS 1 AND 2 3.8-27 3.8.5 DC Sources -- Shutdown LCO 3.8.5 The Train A or Train B DC electrical power subsystem shall be OPERABLE to support one train of the DC electrical power distribution subsystems required by LCO 3.8.10, "Distribution Systems -- Shutdown."
APPLICABILITY:
MODES 5 and 6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required DC electrical power subsystems inoperable.
A.1 Declare affected required feature(s) inoperable.
OR A.2. Suspend movement of irradiated fuel assemblies.
AND A.2. Suspend operations involving positive reactivity additions that could result in loss of required SDM or boron concentration.
AND A.2. Initiate action to restore required DC electrical power subsystem to OPERABLE status.
Immediately Immediately Immediately Immediately Amendment No. 150,
Inverters - Shutdown 3.8.8 3.8 ELECTRICAL POWER SYSTEMS COMANCHE PEAK - UNITS 1 AND 2 3.8-36 3.8.8 Inverters Shutdown LCO 3.8.8 The Train A or Train B inverters shall be OPERABLE to support one train of the onsite Class 1E AC vital bus electrical power distribution subsystems required by LCO 3.8.10, "Distribution Systems -- Shutdown."
APPLICABILITY:
MODES 5 and 6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required inverters inoperable.
A.1 Declare affected required feature(s) inoperable.
OR A.2. Suspend movement of irradiated fuel assemblies.
AND A.2. Suspend operations involving positive reactivity additions that could result in loss of required SDM or boron concentration.
AND A.2. Initiate action to restore required inverters to OPERABLE status.
Immediately Immediately Immediately Immediately Amendment No. 156,
Distribution Systems - Shutdown 3.8.10 3.8 ELECTRICAL POWER SYSTEMS COMANCHE PEAK - UNITS 1 AND 2 3.8-40 3.8.10 Distribution Systems -- Shutdown LCO 3.8.10 The necessary portion of the Train A or Train B AC, DC, and AC vital bus electrical power distribution subsystems shall be OPERABLE to support one train of equipment required to be OPERABLE.
APPLICABILITY:
MODES 5 and 6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required AC, DC, or AC vital bus electrical power distribution subsystems inoperable.
A.1 Declare associated supported required feature(s) inoperable.
OR A.2. Suspend movement of irradiated fuel assemblies.
AND A.2. Suspend operations involving positive reactivity additions that could result in loss of required SDM or boron concentration.
AND Immediately Immediately Immediately Amendment No. 156,
Distribution Systems - Shutdown 3.8.10 ACTIONS COMANCHE PEAK - UNITS 1 AND 2 3.8-41 SURVEILLANCE REQUIREMENTS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued)
A.2. Initiate actions to restore required AC, DC, and AC vital bus electrical power distribution subsystems to OPERABLE status.
AND A.2. Declare associated required residual heat removal subsystem(s) inoperable and not in operation.
Immediately Immediately SURVEILLANCE FREQUENCY SR 3.8.10.1 Verify correct breaker alignments and voltage to required AC, DC, and AC vital bus electrical power distribution subsystems.
In accordance with the Surveillance Frequency Control Program.
Amendment No.
Boron Concentration 3.9.1 COMANCHE PEAK - UNITS 1 AND 2 3.9-1 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration LCO 3.9.1 Boron concentrations of all filled portions of the Reactor Coolant System, the refueling canal, and the refueling cavity, that have direct access to the reactor vessel, shall be maintained within the limit specified in the COLR.
NOTE----------------------------------------------
While this LCO is not met, entry into MODE 6 from MODE 5 is not permitted.
APPLICABILITY:
MODE 6.
ACTIONS SURVEILLANCE REQUIREMENTS CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration not within limit.
A. Suspend positive reactivity additions.
AND A. Initiate action to restore boron concentration to within limit.
Immediately Immediately SURVEILLANCE FREQUENCY SR 3.9.1.1 Verify boron concentration is within the limit specified in the COLR.
In accordance with the Surveillance Frequency Control Program.
Amendment No. 150,
Unborated Water Source Isolation Valves 3.9.2 3.9 REFUELING OPERATIONS COMANCHE PEAK - UNITS 1 AND 2 3.9-2 3.9.2 Unborated Water Source Isolation Valves LCO 3.9.2 Each valve used to isolate unborated water sources shall be secured in the closed position.
APPLICABILITY:
MODE 6.
ACTIONS
NOTE---------------------------------------------------------------
Separate Condition entry is allowed for each unborated water source isolation valve.
CONDITION REQUIRED ACTION COMPLETION TIME A. --------------NOTE--------------
Required Action A. must be completed whenever Condition A is entered.
One or more valves not secured in closed position.
A. Initiate actions to secure valve in A. Perform SR 3.9.1.1.
Immediately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Amendment No.
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Nuclear Instrumentation 3.9.3 3.9 REFUELING OPERATIONS COMANCHE PEAK - UNITS 1 AND 2 3.9-4 3.9.3 Nuclear Instrumentation LCO 3.9.3 Two source range neutron flux monitors shall be OPERABLE.
APPLICABILITY:
MODE 6.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required source range neutron flux monitor inoperable.
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AND A.2 Immediately Immediately B. Two required source range neutron flux monitors inoperable.
B.1 Initiate action to restore one source range neutron flux monitor to OPERABLE status.
AND B.2 Perform S.R.3.9.1.1.
Immediately Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Amendment No.
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Containment Penetrations 3.9.4 3.9 REFUELING OPERATIONS COMANCHE PEAK - UNITS 1 AND 2 3.9-6 3.9.4 Containment Penetrations LCO 3.9.4 The containment penetrations shall be in the following status:
a.
The equipment hatch closed and held in place by four bolts, or if open, capable of being closed; b.
One door in the emergency air lock closed and one door in the personnel airlock capable of being closed; and c.
Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:
1.
closed by a manual or automatic isolation valve, blind flange, or equivalent, or 2.
capable of being closed by an OPERABLE containment ventilation isolation valve.
NOTE---------------------------------------------
Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls.
APPLICABILITY:
During movement of UHFHQWO\irradiated fuel assemblies within containment.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more containment penetrations not in required status.
A. Suspend movement of UHFHQWO\
LUUDGLDWHGIXHOassemblies within containment.
Immediately Amendment No.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 193 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-87 AND AMENDMENT NO. 193 TO RENEWED 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 application dated September 12, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24256A088), Vistra Operations Company LLC (Vistra OpCo, the licensee), requested changes to the technical specifications (TSs) for Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2 (Comanche Peak, CPNPP).
The proposed changes would revise the Comanche Peak TSs consistent with the following U.S.
Nuclear Regulatory Commission (NRC)-approved Technical Specifications Task Force (TSTF) travelers:
TSTF-51-A, Revision 2 (TSTF-51), Revise containment requirements during handling irradiated fuel and core alterations (ML040400343). TSTF-51 revises certain TSs to remove the requirements for certain engineered safety features systems to operate after sufficient radioactive decay of irradiated fuel has occurred following a plant shutdown.
TSTF-471-A, Revision 1 (TSTF-471), Eliminate use of term CORE ALTERATIONS in ACTIONS and Notes (ML19101A215). Similar to TSTF-51, TSTF-471 eliminates the remaining uses of the defined term CORE ALTERATIONS.
TSTF-571-T, Revision 0 (TSTF-571), Revise Actions for Inoperable Source Range Neutron Flux Monitor (pages 6 to 30 of the document in ADAMS Accession No. ML18221A561). TSTF-571 revises the Required Actions for an inoperable source range neutron flux monitor to prohibit the movement of fuel assemblies, sources, and reactivity control components when a core subcritical neutron flux monitor is inoperable.
A provision is included to allow such movement if it is needed to repair the core subcritical neutron flux monitor.
1.1 Proposed TS Changes
TSTF-51 Changes Proposed Comanche Peak TS changes consistent with TSTF-51 include deletion of the use of CORE ALTERATIONS in several TSs and the replacement of the term irradiated with recently irradiated in TS applicability, conditions, and required actions. The following TSs would be affected:
TS 3.3.6, Containment Ventilation Isolation Instrumentation
TS 3.3.7, Control Room Emergency Filtration System (CREFS) Actuation Instrumentation
TS 3.7.10, Control Room Emergency Filtration/Pressurization System (CREFS)
TS 3.7.11, Control Room Air Conditioning System (CRACS)
TS 3.8.2, AC [Alternating Current] Sources - Shutdown
TS 3.8.5, DC [Direct Current] Sources - Shutdown
TS 3.8.8, Inverters - Shutdown
TS 3.8.10, Distribution Systems - Shutdown
TS 3.9.4, Containment Penetrations TSTF-471-A Changes Proposed Comanche Peak TS changes consistent with TSTF-471-A include deletion of the definition and the remaining uses of CORE ALTERATIONS in several TSs. The following TSs would be affected:
TS 1.1, Definitions
TS 3.8.2, AC Sources - Shutdown
TS 3.8.5, DC Sources - Shutdown
TS 3.8.8, Inverters - Shutdown
TS 3.8.10, Distribution Systems - Shutdown
TS 3.9.2 Unborated Water Source Isolation Valves TSTF-571-T Changes Proposed Comanche Peak TS changes consistent with TSTF-571-T would revise a required action in TS 3.9.3, Nuclear Instrumentation, to prohibit the movement of fuel assemblies, sources, and reactivity control components when a source range neutron flux monitor is inoperable. In addition, a Note to the proposed required action would allow movement as needed to support source range neutron flux monitor repair activities.
2.0 REGULATORY EVALUATION
2.1 Regulatory Requirements Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, Technical specifications, provides the regulatory requirements for the content of TSs. The categories of items required to be in the TSs are provided in 10 CFR 50.36(c). Specifically, the regulations in 10 CFR 50.36(c)(2)(i) requires, in part, that the TSs to include limiting conditions for operation (LCOs),
which are the lowest functional capability or performance levels of equipment required for safe operation of a nuclear reactor facility. In the case when an LCO is not met, the licensee shall shut down the reactor or follow the remedial actions allowed by the TSs until the condition can be met.
The regulation in 10 CFR 50.36(a)(1) states, in part, that [a] summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications.
Accordingly, the licensee submitted TSs Bases changes that correspond to the proposed TSs changes for information only. Changes to the Comanche Peak TS bases shall be implemented by the licensee in accordance with TS 5.5.14, Technical Specifications (TS) Bases Control Program.
2.2 Regulatory Guidance An NRC letter addressed to the TSTF contained the NRC staffs resolution of concerns regarding potential operating issues in plant-specific implementation of TSTF-51, Revision 2; TSTF-471, Revision 1; and TSTF-286, Revision 2. The \
letter dated October 4, 2018 (ML17346A587), contains descriptions of additional radiological information or analysis that licensees should include in applications to support review and adoption of these travelers. An NRC letter addressed to the TSTF dated November 7, 2013 (ML13246A358), documented the staffs initial concerns, which were later resolved in the \
Traveler TSTF-286, Operations Involving Positive Reactivity Additions, was approved by an NRC \
letter dated July 6, 2000 (ML003730788). TSTF-286 changes the standard TS (STS) by providing the flexibility necessary to provide for continued safe reactor operations, while also limiting any potential for excess positive reactivity addition. (NRC staff concerns with TSTF-286 led to the development of TSTF-571.)
Travelers TSTF-51-A, TSTF-471-A, and TSTF-571-T provide acceptable methods of modifying core monitoring instrumentation and dose consequences requirements contained in the STS.
Though travelers TSTF-51, TSTF-471, and TSTF-571 were developed based on changes to NUREG-1431, Revisions 1, 3, and 4, respectively, the NRC staffs review of this license amendment request (LAR) includes consideration of whether the proposed changes are consistent with the latest revision. NUREG-1431, Revision 5, Standard Technical Specifications for Westinghouse Plants, Volume 1, Specifications (ML21259A155), provides example TS LCOs and acceptable remedial actions that meet the requirements in 10 CFR 50.36(c)(2)(i) for a standard plant design.
3.0 TECHNICAL EVALUATION
3.1 Applicable Accidents and Transients When the reactor vessel head is unbolted and removed, core alterations take place during operating Mode 6 (refueling operation). There are only two accidents considered during Mode 6.
These accidents are: (1) a fuel handling accident (FHA), and (2) a boron dilution accident. An FHA is initiated by the dropping of an irradiated fuel assembly, either in the containment or in the spent fuel pool. There are no mitigation actions, except for taking credit for ventilation systems to reduce the dose consequences. Thus, the suspension of core alterations, except for suspension of movement of irradiated fuel, will not prevent or impair the mitigation of an FHA.
The analysis for an FHA assumes that a fuel assembly is dropped during fuel handling in the containment or the spent fuel pool. Interlocks and procedural, and administrative controls make such an event highly unlikely. However, if an assembly were damaged to the extent that one or more fuel rods were broken, the accumulated fission product gases and iodines in the fuel element gap would be released to the surrounding water. Release of the solid fission products in the fuel would be negligible because of the low fuel temperature during refueling, which greatly limits their diffusion.
A boron dilution accident is initiated by a dilution source that results in the boron concentration dropping below the value required to maintain the shutdown margin. TS 3.9.1 applies in Mode 6, and the refueling boron concentration limit is specified in the core operating limits report. This accident is mitigated by stopping the dilution. The suspension of core alterations has no effect on the mitigation of a boron dilution accident. Also, the control rods or fuel do not affect the initial conditions of a boron dilution accident.
3.2 Review of Comanche Peak FHA Analysis In section 3.1, Current Licensing Basis and Accident Analysis, of the enclosure to the LAR, the licensee stated, in part:
This submittal does not modify the fuel handling accident analysis as previously approved by the NRC in CPNPP License Amendment 146, Section 2.9.8,
[ML081510173].
Further, in section 4.1, Applicable Regulatory Requirement, of the LAR enclosure, the licensee stated:
The proposed amendment does not alter the design or operation of the control room envelope or the CREFS or CRACS. The FHA analysis is not changed by this submittal and continues to show the radiological dose to the MCR [main control room] personnel to be within requirements.
The NRC staff reviewed the initial conditions, inputs, and assumptions for the accident dose consequence analysis supporting the FHA in both the LAR and in Amendment Nos. 146 and 146, Comanche Peak, Units 1, and 2-Issuance of Amendment Nos. 146 and 146, Stretch Power Uprate, Revision to TS 1.0, Use and Application, to Revise Rated Thermal Power from 3458 to 3612 MWt (TAC Nos. MD6615 and MD6616), dated June 27, 2008 (ML081510173). The review also included Comanche Peaks updated final safety analysis report. The staffs review independently verified that there is no change to the current licensing basis FHA analysis contained in this application. As there is no change in the FHA dose consequence analysis, and the analysis of record shows that the dose estimates at the exclusion area boundary, low population zone, and in the control room continue to meet applicable regulatory requirements, this application is acceptable from an accident dose consequence analysis perspective.
3.3 Evaluation of Proposed TSTF-51 and TSTF-471 TS Changes In an NRC letter to the TSTF dated October 4, 2018, the NRC staff states, in part:
After considerable review and analysis, the NRC staff concludes that for certain facilities, LARs adopting TSTF-51 and TSTF-471 could result in exceeding the bounding licensing basis Fuel Handling Accident (FHA) analysis of record dose for the control room and is therefore considered an unanalyzed condition.
The NRC staff determined that following approval of LARs to adopt TSTF-51 and TSTF-471, the containment purge and exhaust isolation instrumentation and the containment penetrations will no longer require to be operable during core alterations. The NRC staff identified that dropping a source, fuel assembly, or component during core alterations could damage a recently irradiated fuel assembly creating a radioactive source term that may result in exceeding the resultant radiological doses calculated by the licensing basis FHA analysis of record. Therefore, the NRC staff recommended in the \
October 4, 2018, letter that when licensees request to adopt TSTF-51 and TSTF-471, one of the following discussions be provided to support removal of the defined term core alterations from the TS applicability:
Confirm that the length of time defined as recently is less than the time required to remove the reactor vessel head and internals and expose the irradiated fuel after a shutdown; Provide an analysis that demonstrates that the dropping of any unirradiated fuel assembly, sources, reactivity control component, or other component affecting reactivity within the reactor vessel onto irradiated fuel assemblies prior to the period of time defined as recently will not result in a radioactive release from the irradiated fuel; Describe the limitations or controls that would prevent movement of any unirradiated fuel assembly, source, reactivity control component, or other component affecting reactivity within the reactor vessel capable of damaging a fuel assembly prior to the time period defined as recently; or Provide an analysis that demonstrates that the dose consequences of a failure of a single irradiated fuel assembly with no technical specification-required mitigation systems available remain below the regulatory limits and the regulatory guidance limits for a fuel handling accident.
The licensee stated in its LAR that the term recently, as in recently irradiated fuel, is defined as fuel that has occupied part of a critical reactor core within the previous 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />. Consistent with the first bullet above, the licensee determined that after a reactor shutdown, 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> is approximately half the time needed to remove the reactor vessel head and internals, exposing the irradiated fuel. Also, the licensee stated in the LAR that the Comanche Peak Technical Requirements Manual technical requirement 13.9.31 Decay Time, will prohibit movement of irradiated fuel in the reactor vessel unless the reactor is subcritical for a minimum of 75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br />.
Based on this, the NRC staff has determined that dropping a source, fuel assembly, or component during core alterations will not damage a recently irradiated fuel assembly. Without the creation of a radioactive source term from a core alterations event, the radiological doses will remain bounded by the current FHA with recently irradiated fuel analysis, which assumes the FHA will occur 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> after plant shutdown. The LAR has addressed the staffs concerns with adoption of TSTF-51-A and TSTF-471-A. Therefore, the NRC staff finds acceptable the proposed changes consistent with TSTF-51-A and TSTF-471-A because they meet 10 CFR 50.36(c)(2); and the revised TS applicability, conditions, and remedial actions will allow for safe operation of the plant.
3.4 Evaluation of Proposed TSTF-571 TS Changes During its conversion to improved STS, Comanche Peak TS 3.9.3, Required Action A.2, incorporated changes related to TSTF-286 that were issued in the license amendments dated February 26, 1999 (ML021820213).
In the letter to the TSTF dated November 7, 2013, NRC staff raised concerns with the TSTF-286 changes to required actions in the nuclear instrumentation specification when one source range neutron flux monitor is inoperable. As stated in section 4 of TSTF-571-T, the concern was that during the movement of fuel assemblies, sources, and reactivity, control components with one source range neutron flux monitor inoperable, there was potential for the redundant operable monitor to become effectively decoupled from the core reactivity condition (hereafter referred to as decouple-effect). For example, if one source range neutron flux monitor were inoperable and certain strategically located fuel assemblies were removed, then the operable monitor may no longer be capable of monitoring the reactivity condition of fuel assemblies located in the far half of the core. As a result of the TSTF-286 changes, a situation may be created in which a reactivity increase in the reactor core might not be detected.
In addressing the NRCs concern regarding the decouple-effect, the licensee proposed to revise the requirement in Comanche Peak TS 3.9.3, Required Action A.2, to read [s]uspend movement of fuel, sources, and reactivity control components within the reactor vessel, as modified by a note which would state, [f]uel assemblies, sources, and reactivity control components may be moved if necessary to restore an inoperable source range neutron flux monitor or to complete movement of a component to a safe condition. The NRC staff determined that this TSTF-571 revision that would affect TS 3.9.3, Required Action A.2, for one source range neutron flux monitor inoperable would avoid a reactivity change and adequately addresses the NRC staffs concern regarding the decouple-effect.
The proposed changes to Comanche Peak TS 3.9.3, Required Action A.2, are acceptable TS remedial actions that would continue to assure that the boron concentration limit requirements are met; the LAR has addressed the NRC staffs concerns with the previous adoption of TSTF-286. Therefore, the NRC staff finds that the proposed changes to TS 3.9.3, consistent with TSTF-571-T, will meet 10 CFR 50.36(c)(2) in that following the proposed required action will allow for safe operation of the plant and is, therefore, 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 August 21, 2025. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION The amendments change a requirement 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 November 26, 2024 (89 FR 93365), 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 Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: K. West, NRR J. Daniel, NRR S. Meighan, NRR Date: August 27, 2025
- eConcurrence
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