ML24297A130
| ML24297A130 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 12/18/2024 |
| From: | Scott Wall Plant Licensing Branch III |
| To: | Lies Q Indiana Michigan Power Co |
| Wall S | |
| References | |
| EPID L-2023-LLA-0011 | |
| Download: ML24297A130 (1) | |
Text
December 18, 2024 Q. Shane Lies Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106
SUBJECT:
DONALD C. COOK NUCLEAR PLANT, UNIT NOS. 1 AND 2 ISSUANCE OF AMENDMENT NOS. 364 AND 345 REGARDING CHANGE TO THE TECHNICAL SPECIFICATION REQUIREMENT FOR NEUTRON FLUX INSTRUMENTATION (EPID L-2023-LLA-0011)
Dear Q. Shane Lies:
The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment Nos. 364 and 345 to Renewed Facility Operating License Nos. DPR-58 and DPR-74, for the Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, respectively. The amendments consist of changes to the technical specifications (TSs) in response to your application dated January 26, 2023, as supplemented by letters dated August 2, 2023, February 27, 2024, and August 15, 2024. The amendment revises TS 3.3.3, Post Accident Monitoring (PAM) Instrumentation to not require environmental qualification to be maintained for TS 3.3.3, Function 1, Neutron Flux, instrumentation.
A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Scott P. Wall, Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316
Enclosures:
- 1. Amendment No. 364 to DPR-58
- 2. Amendment No. 345 to DPR-74
- 3. Safety Evaluation
- 4. Notice and Environmental Finding cc: Listserv
INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-315 DONALD C. COOK NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 364 License No. DPR-58
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Indiana Michigan Power Company dated January 26, 2023, as supplemented by letters dated August 2, 2023, February 27, 2024, and August 15, 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. DPR-58 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 364, are hereby incorporated in this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: December 18, 2024 ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2024.12.18 15:25:49 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 364 DONALD C. COOK NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE DOCKET NO. 50-315 Renewed Facility Operating License No. DPR-58 Replace the following page of the Renewed Facility Operating License No. DPR-58 with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the area of change.
REMOVE INSERT Technical Specifications Replace the following pages of the Renewed Facility Operating License, Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT 3.3.3-4 3.3.3-4 3.3.3-5 3.3.3-5 Renewed License No. DPR-58 Amendment No: 363, 364 and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; and (5)
Pursuant to the Act and 10 CFR Parts 30 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 operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:
Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not to exceed 3304 megawatts thermal in accordance with the conditions specified herein.
(2)
Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 364, are hereby incorporated in this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Less than Four Loop Operation The licensee shall not operate the reactor at power levels above P-7 (as defined in Table 3.3.1-1 of Specification 3.3.1 of Appendix A to this renewed operating license) with less than four reactor coolant loops in operation until (a) safety analyses for less than four loop operation have been submitted, and (b) approval for less than four loop operation at power levels above P-7 has been granted by the Commission by amendment of this license.
(4)
Fire Protection Program Indiana Michigan Power Company shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensees amendment request dated July 1, 2011, as supplemented by letters dated September 2, 2011, April 27, 2012, June 29, 2012, August 9, 2012, October 15, 2012, November 9, 2012, January 14, 2013, February 1, 2013,
PAM Instrumentation 3.3.3 Cook Nuclear Plant Unit 1 3.3.3-4 Amendment No. 287, 313, 360, 364 Table 3.3.3-1 (page 1 of 2)
Post Accident Monitoring Instrumentation FUNCTION REQUIRED CHANNELS CONDITION REFERENCED FROM REQUIRED ACTION E.1 1.
Neutron Flux 2(a)
F 2.
Steam Generator Pressure (per steam generator) 2 F
3.
Reactor Coolant System (RCS) Hot Leg Temperature (Wide Range) 2 F
4.
RCS Cold Leg Temperature (Wide Range) 2 F
5.
RCS Pressure (Wide Range) 2 F
6.
Reactor Coolant Inventory Tracking System (Reactor Vessel Level Indication) 2 G
7.
Containment Water Level 2(b)
F 8.
Containment Pressure (Narrow Range) 2 F
9.
Penetration Flow Path Containment Isolation Valve Position 2 per penetration flow path(c)(d)
F 10.
Containment Area Radiation (High Range) 2 G
11.
Deleted 12.
Pressurizer Level 2
F 13.
Steam Generator Water Level (Wide Range) 4 F
14.
Condensate Storage Tank Level 1
G 15.
Core Exit Temperature - Quadrant 1 2(e)
F 16.
Core Exit Temperature - Quadrant 2 2(e)
F 17.
Core Exit Temperature - Quadrant 3 2(e)
F 18.
Core Exit Temperature - Quadrant 4 2(e)
F (a) Channels used to satisfy Function 1 are not required to be environmentally qualified.
(b) Up to one channel of Function 7 OPERABILITY requirements can be satisfied by an OPERABLE train of containment water level switches if both Containment Water Level channels are inoperable. This substitution is only allowed until the end of the current operating cycle when it is invoked.
(c)
Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.
(d) Only one position indication channel is required for penetration flow paths with only one installed control room indication channel.
(e) A channel consists of one core exit thermocouple (CET).
PAM Instrumentation 3.3.3 Cook Nuclear Plant Unit 1 3.3.3-5 Amendment No. 287, 299, 313, 360, 364 Table 3.3.3-1 (page 2 of 2)
Post Accident Monitoring Instrumentation FUNCTION REQUIRED CHANNELS CONDITION REFERENCED FROM REQUIRED ACTION E.1 19.
Secondary Heat Sink Indication (per steam generator) 2(f)
F 20.
Emergency Core Cooling System Flow (per train) 2(g)
F 21.
Containment Pressure (Wide Range) 2 F
22.
Refueling Water Storage Tank Level 2
F 23.
RCS Subcooling Margin Monitor 1(h)
F 24.
Component Cooling Water Pump Circuit Breaker Status 2
G 25.
Containment Recirculation Sump Water Level 2
F (f)
Any combination of two instruments per steam generator, including Steam Generator Water Level (Narrow Range) and Auxiliary Feedwater Flow, can be used to satisfy Function 19 OPERABILITY requirements.
(g) Any combination of two instruments per train, including Centrifugal Charging Pump Flow, Safety Injection Pump Flow, Centrifugal Charging Pump Circuit Breaker Status, and Safety Injection Pump Circuit Breaker Status, can be used to satisfy Function 20 OPERABILITY requirements.
(h) An OPERABLE plant process computer (PPC) subcooling margin readout can be used as a substitute for an inoperable Function 23, RCS Subcooling Margin Monitor.
INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-316 DONALD C. COOK NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 345 License No. DPR-74 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Indiana Michigan Power Company dated January 26, 2023, as supplemented by letters dated August 2, 2023, February 27, 2024, and August 15, 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. DPR-74 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 345, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Ilka Berrios, Acting Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: December 18, 2024 ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2024.12.18 15:26:18 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 345 DONALD C. COOK NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE DOCKET NO. 50-316 Renewed Facility Operating License No. DPR-74 Replace the following page of the Renewed Facility Operating License No. DPR-74 with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the area of change.
REMOVE INSERT Technical Specifications Replace the following pages of the Renewed Facility Operating License, Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
REMOVE INSERT 3.3.3-4 3.3.3-4 3.3.3-5 3.3.3-5 Renewed License No. DPR-74 Amendment No. 344, 345 and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)
Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; and (5)
Pursuant to the Act and 10 CFR Parts 30 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 operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:
Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not to exceed 3468 megawatts thermal in accordance with the conditions specified herein and in Attachment 1 to the renewed operating license. The preoperational tests, startup tests and other items identified in to this renewed operating license shall be completed. is an integral part of this renewed operating license.
(2)
Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 345, are hereby incorporated in this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Additional Conditions (a) Deleted by Amendment No. 76 (b) Deleted by Amendment No. 2 (c) Leak Testing of Emergency Core Cooling System Valves Indiana Michigan Power Company shall prior to completion of the first inservice testing interval leak test each of the two valves in series in the
PAM Instrumentation 3.3.3 Cook Nuclear Plant Unit 2 3.3.3-4 Amendment No. 269, 296, 342, 345 Table 3.3.3-1 (page 1 of 2)
Post Accident Monitoring Instrumentation FUNCTION REQUIRED CHANNELS CONDITION REFERENCED FROM REQUIRED ACTION E.1 1.
Neutron Flux 2(a)
F 2.
Steam Generator Pressure (per steam generator) 2 F
3.
Reactor Coolant System (RCS) Hot Leg Temperature (Wide Range) 2 F
4.
RCS Cold Leg Temperature (Wide Range) 2 F
5.
RCS Pressure (Wide Range) 2 F
6.
Reactor Coolant Inventory Tracking System (Reactor Vessel Level Indication) 2 G
7.
Containment Water Level 2(b)
F 8.
Containment Pressure (Narrow Range) 2 F
9.
Penetration Flow Path Containment Isolation Valve Position 2 per penetration flow path(c)(d)
F 10.
Containment Area Radiation (High Range) 2 G
11.
Deleted 12.
Pressurizer Level 2
F 13.
Steam Generator Water Level (Wide Range) 4 F
14.
Condensate Storage Tank Level 1
G 15.
Core Exit Temperature - Quadrant 1 2(e)
F 16.
Core Exit Temperature - Quadrant 2 2(e)
F 17.
Core Exit Temperature - Quadrant 3 2(e)
F 18.
Core Exit Temperature - Quadrant 4 2(e)
F (a) Channels used to satisfy Function 1 are not required to be environmentally qualified.
(b) Up to one channel of Function 7 OPERABILITY requirements can be satisfied by an OPERABLE train of containment water level switches if both Containment Water Level channels are inoperable. This substitution is only allowed until the end of the current operating cycle when it is invoked.
(c)
Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.
(d) Only one position indication channel is required for penetration flow paths with only one installed control room indication channel.
(e) A channel consists of one core exit thermocouple (CET).
PAM Instrumentation 3.3.3 Cook Nuclear Plant Unit 2 3.3.3-5 Amendment No. 269, 282, 296, 342, 345 Table 3.3.3-1 (page 2 of 2)
Post Accident Monitoring Instrumentation FUNCTION REQUIRED CHANNELS CONDITION REFERENCED FROM REQUIRED ACTION E.1 19.
Secondary Heat Sink Indication (per steam generator) 2(f)
F 20.
Emergency Core Cooling System Flow (per train) 2(g)
F 21.
Containment Pressure (Wide Range) 2 F
22.
Refueling Water Storage Tank Level 2
F 23.
RCS Subcooling Margin Monitor 1(h)
F 24.
Component Cooling Water Pump Circuit Breaker Status 2
G 25.
Containment Recirculation Sump Water Level 2
F (f)
Any combination of two instruments per steam generator, including Steam Generator Water Level (Narrow Range) and Auxiliary Feedwater Flow, can be used to satisfy Function 19 OPERABILITY requirements.
(g) Any combination of two instruments per train, including Centrifugal Charging Pump Flow, Safety Injection Pump Flow, Centrifugal Charging Pump Circuit Breaker Status, and Safety Injection Pump Circuit Breaker Status, can be used to satisfy Function 20 OPERABILITY requirements.
(h) An OPERABLE plant process computer (PPC) subcooling margin readout can be used as a substitute for an inoperable Function 23, RCS Subcooling Margin Monitor.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 364 AND 345 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-58 AND DPR-74 INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-315 AND 50-316 Application:
Safety Evaluation Date:
January 26, 2023 ADAMS Accession No. ML23026A284 Supplements:
August 2, 2023 (ML23214A289)
February 27, 2024 (ML24058A357)
August 15, 2024 (ML24228A161)
December 18, 2024 Principal Contributors to Safety Evaluation:
David Rahn Hang Vu Santosh Bhatt Ahsan Sallman DaBin Gibbs Michelle Honcharik
1.0 INTRODUCTION
Indiana Michigan Power Company (the licensee) requested changes to the technical specifications (TSs) for Donald C. Cook Nuclear Plant (CNP), Unit Nos. 1 and 2, by license amendment request (application or LAR). The amendment revises TS 3.3.3, Post Accident Monitoring (PAM) Instrumentation to not require environmental qualification (EQ) to be maintained for TS 3.3.3, Function 1, Neutron Flux, instrumentation.
On March 21, 2023, the NRC published a notice of consideration of the application and a proposed no significant hazards consideration (NSHC) determination in the Federal Register (88 FR 17036). The supplemental letter of August 2, 2023, provided additional information that clarified the application, did not expand the scope of the application as noticed on March 21, 2023, and did not change the NRC staffs proposed no significant hazards consideration determination as published in the Federal Register on March 21, 2023 (88 FR 17036).The supplemental letter dated February 27, 2024, provided additional information that expanded the scope of the application as originally noticed. Therefore, the NRC published a notice of consideration of the application, as supplemented on August 2, 2023, and February 27, 2024, in the Federal Register on April 16, 2024 (89 FR 26946). The supplemental letter of August 15, 2024, provided additional information that clarified the application, did not expand the scope of the application as noticed on April 16, 2024, and did not change the NRC staffs proposed no significant hazards consideration determination as published in the Federal Register on April 16, 2024 (89 FR 26946).
2.0 REGULATORY EVALUATION
2.1 Regulatory Requirements and Guidance 2.1.1 Regulatory Requirements In Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.36, the U.S.
Nuclear Regulatory Commission (NRC, the Commission) established its regulatory requirements related to the content of TSs. Pursuant to 10 CFR 50.36(c), TSs for nuclear reactors are required to include items in the following categories related to station operation:
(1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements; (4) design features; and (5) administrative controls.
Under 10 CFR 50.36(c)(2), TSs must contain LCOs, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the LCO can be met. Typically, the TSs require restoration of equipment in a timeframe commensurate with its safety significance, along with other engineering considerations. Under 10 CFR 50.36(b), TSs must be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto.
The requirements in 10 CFR 50.36(c)(2)(ii) set forth four criteria to be used in determining whether a LCO is required to be included in TSs. These criteria are:
Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.
10 CFR 50.34(f)(2) requires applicants and licensees to provide instrumentation adequate for monitoring plant conditions following an accident that includes core damage.
10 CFR 50.49 requires holders or applicants for an operating license issued under Part 50 to establish a program for the EQ of electric equipment In accordance with 10 CFR 50.92(a), when determining whether to issue a license amendment, the Commission will be guided by the considerations which govern the issuance of initial licenses to the extent applicable and appropriate. In determining whether the proposed TS remedial actions should be granted, the staff typically applies the reasonable assurance standard derived from the requirements of 10 CFR 50.40(a) and 50.57(a)(3). The regulation at 10 CFR 50.40(a) states that in determining whether to grant the licensing request, the Commission will be guided by, among other things, consideration about whether the processes to be performed, the operating procedures, the facility and equipment, the use of the facility, and other technical specifications, or the proposals, in regard to any of the foregoing collectively provide reasonable assurance that the applicant will comply with the regulations in this chapter, including the regulations in Part 20 of this chapter, and that the health and safety of the public will not be endangered. The regulation at 10 CFR 50.57(a)(3) states that the Commission may issue an operating license when, in part, there is reasonable assurance that the activities authorized by the operating license can be conducted without endangering the health and safety of the public, and such activities will be conducted in compliance with the regulations in this chapter.
The NRC regulations codified in Appendix A, General Design Criteria for Nuclear Power Plants, of 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, contain, in pertinent part, the following General Design Criteria (GDC):
GDC 13, Instrumentation and control, states that [i]nstrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.
GDC 19, Control room, states that a control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for the prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.
The construction permits for CNP Units 1 and 2 were issued in 1969, before the 1971 final rule was issued, promulgating the GDCs. As discussed in the staff requirements memorandum (SRM)-SECY-92-223 (ML003763736), the Commission approved the staffs recommendation to continue its approach of not applying the GDC to plants with construction permits issued prior to May 21, 1971. The Commission also stated, in pertinent part, that:
At the time of promulgation of Appendix A to 10 CFR Part 50, the Commission stressed that the GDC were not new requirements and were promulgated to more clearly articulate the licensing requirements and practice in effect at that time. While compliance with the intent of the GDC is important, each plant licensed before the GDC were formally adopted was evaluated on a plant specific basis, determined to be safe, and licensed by the Commission.
Furthermore, current regulatory processes are sufficient to ensure that plants continue to be safe and comply with the intent of the GDC.
The plant-specific design criteria (PSDC) for CNP are discussed further below. Also, as discussed in Section 1.4.10 of the CNP updated final safety analysis report (ML22340A150), the CNP became obligated to conform with GDC 19 after initial licensing of CNP to the PSDCs.
2.1.2 Regulatory Guidance and Updated Final Safety Analysis Report (UFSAR) Criteria The NRC staff considered the following guidance, along with industry guidance endorsed by the NRC, during its review of the proposed changes:
NUREG-0737, Clarification of TMI [Three Mile Island] Action Plan Requirements, dated November 1980 (ML051400209), incorporates all TMl-related items approved for implementation by the Commission up to that point.
NUREG-0800, Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, is intended to be a comprehensive and integrated document which provides guidance that describes methods or approaches that the staff has found acceptable for meeting NRC requirements.
Branch Technical Position (BTP) 7-10, Guidance on Application of Regulatory Guide 1.97, Revision 6, dated August 2016 (ML16019A169), provides additional guidelines for reviewing a licensees accident monitoring instrumentation. These guidelines are based on reviews of design submittals by applicants or licensees that contained approved interpretations and alternatives for the guidance identified in RG 1.97. Specifically, BTP 7-10, Table 2, For PWRs: Acceptable Deviations and Clarifications to Revisions 2 and 3 of Regulatory Guide 1.97, identifies guidelines for the range, design and qualification category, and purpose for PWR variables. In addition, Table 2, identifies acceptable deviations from RG 1.97 guidelines (e.g., deviations with respect to category, redundancy, range, direct measurement), and provides a summary of the acceptance guidelines or clarification associated with the deviations. Table 2 states, in part:
Variable Deviation Acceptance Guidelines/Clarification Neutron Flux Environmental qualification A non-environmentally qualified instrument is acceptable if qualified core exit thermocouples and reactor coolant system (RCS) hot and cold leg temperature indications are provided in conjunction with directions in emergency procedures for operator action to ensure that boric acid injection is occurring.
Regulatory Guide (RG) 1.97, Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident, Revision 3, May 1983 (ML003740282), describes a method that is acceptable to the staff for use in complying with the Commissions regulations to provide instrumentation to monitor plan variables and systems during and following an accident in a light-water-cooled nuclear power plant.
CNP UFSAR, Section 1.4, Plant Specific Design Criteria (PSDC) (ML24159A180), states that the CNP specific design is committed to meet the intent of the proposed GDC published in the Federal Register on July 11, 1967. The following PSDC criterion are applicable:
PSDC CRITERION 11, Control Room, states:
The facility shall be provided with a control room from which actions to maintain safe operational status of the plant can be controlled. Adequate radiation protection shall be provided to permit continuous occupancy of the control room under any credible post-accident condition or as an alternative, access to other areas of the facility as necessary to shutdown and maintain safe control of the facility without excessive radiation exposures of personnel.
PSDC CRITERION 12, Instrumentation and Control Systems, states:
Instrumentation and controls shall be provided as required to monitor and maintain within prescribed operating ranges essential reactor facility operating variables.
PSDC CRITERION 13, Fission Process Monitors and Controls, states:
Means shall be provided for monitoring or otherwise measuring and maintaining control over the fission process throughout core life under all conditions that can reasonably be anticipated to cause variations in reactivity of the core.
2.2
System Description
The instrumentation and control (I&C) systems of CNP Units 1 and 2 provide the reactor operator with the required information and control capability to operate the station in a safe and efficient manner. The I&C systems monitor all operationally important reactor operating parameters such as neutron flux, system pressures, flow rates, temperatures, levels, etc.
CNP Units 1 and 2 are Westinghouse Pressurized Water Reactors (PWRs). As such, their Emergency Operating Procedures (EOPs) are based on the Emergency Response Guidelines (ERGs) developed by Westinghouse Electric Company. In Generic Letter (GL) 83-22, Safety Evaluation of Emergency Response Guidelines dated June 3, 1983 (ML031080193), the NRC staff found the Westinghouse ERGs acceptable for EOP development.
PAM instrumentation is required to display the required plant variables after an accident to monitor the post-accident plant conditions. One of these variables is neutron flux. Per RG 1.97, Revision 3, this variable is classified as Type B, Category 1. Type B, Category 1 instruments should be redundant and backed by uninterrupted power supply, and should meet the specified environmental qualifications. During a design basis accident (DBA), wide range (WR) Neutron Flux instrumentation provides information to control room operators in two situations:
a) to check if the reactor is no longer critical; and b) to monitor the core for unexpected additions of reactivity after reactor shutdown has been achieved.
Following an accident in which energy is released within the containment, Neutron Flux instrumentation connections and cabling can be exposed to elevated temperature and humidity conditions, as well as other DBA conditions. Therefore, to fulfill the PAM function of monitoring the core for unexpected additions of reactivity after reactor shutdown has been achieved, the WR Neutron Flux monitoring equipment at CNP has been environmentally qualified for post-accident containment environmental conditions.
In the August 2, 2023 supplement, the licensee stated, in part, that:
In June of 1987, l&M submitted a letter indicating a decision to install environmentally qualified WR neutron flux instrumentation at CNP Unit 1 and Unit 2 during the refueling outages scheduled to occur in 1989, as part of compliance with RG 1.97 recommendations. The WR neutron flux instrumentation installed at CNP Unit 1 and Unit 2 (referred to as Gamma-Metrics or WR log power) is comprised of fission chambers located exterior to the reactor vessel, in a location similar to that of the existing Westinghouse nuclear instrumentation (source range, intermediate range, and power range detectors). The Gamma-Metrics detectors rely on thermal neutrons leaking from the core in order to provide an indication of in-core activity.
2.3 Reason for the Proposed Change 2.3.1 Background In the January 26, 2023, application, the licensee stated that during the most recent CNP Unit 2 refueling outage (RFO), while performing testing on the WR Neutron Flux instrumentation, small leaks were detected in the jacketing of the cable for 2-NRI-21 (Nuclear Instrumentation Channel 1 WR Radiation Detector). The jacket breach does not interfere with the ability of 2-NRI-21 to perform its required neutron flux monitoring under normal operating conditions.
However, the WR Neutron Flux channel may become unreliable under conditions where high temperature and high humidity are present, such as following a Loss of Coolant Accident (LOCA) event or a Main Steam Line Break (MSLB) event. During the CNP Unit 2 RFO, the neutron flux monitoring equipment vendor was not available for emergent support to repair or replace the portion of this cable within the required completion time. As a result, channel 2-NRI-21 does not meet EQ acceptance criteria and is considered inoperable.
The licensee indicated that during the RFO, it did attempt to engage with the instrument vendor to address the inoperable equipment. However, the vendor was not available for emergent support for repair or replacement during the CNP Unit 2 RFO. Additionally, the vendor notified the licensee of their intent to discontinue maintenance and support of these instruments at the end of 2024. As such, the licensee is unable to repair 2-NRI-21 to meet EQ acceptance criteria.
Further, as availability of spare parts and vendor support for these instruments are coming to an end, the licensee would be unable to repair other WR Neutron Flux channels.
2.3.2 Current PAM Instrumentation TS Requirements CNP TS 3.3.3, Table 3.3.3-1, Post Accident Monitoring Instrumentation, lists the PAM instrumentation channels which are required by the TS to be operable, and also references the relevant CONDITION listed in TS 3.3.3. The PAM instrumentation functions relevant to this LAR are listed in Table 1 below:
Table 1: PAM Instrumentation ACTION Requirements FUNCTION REQUIRED CHANNELS CONDITION REFERENCED FROM REQUIRED ACTION E.1
- 1.
Neutron Flux 2
F
- 3.
Reactor Coolant System (RCS) Hot Leg Temperature (Wide Range) 2 F
- 4.
RCS Cold Leg Temperature (Wide Range) 2 F
- 15. Core Exit Temperature - Quadrant 1 2(d)
F
- 16. Core Exit Temperature - Quadrant 2 2(d)
F
- 17. Core Exit Temperature - Quadrant 3 2(d)
F
- 18. Core Exit Temperature - Quadrant 4 2(d)
F
- 20. Emergency Core Cooling System Flow (per train) 2(f)
F (d)
A channel consists of one core exit thermocouple (CET).
(f)
Any combination of two instruments per train, including Centrifugal Charging Pump Flow, Safety Injection Pump Flow, Centrifugal Charging Pump Circuit Breaker Status, and Safety Injection Pump Circuit Breaker Status, can be used to satisfy Function 20 OPERABILITY requirements.
The LCO for TS 3.3.3, and required action and completion times of these FUNCTIONS, which are relevant to this LAR are listed in Table 2 below:
Table 2: TS 3.3.3 ACTIONS and COMPLETION TIMES CONDITION REQUIRED ACTION COMPLETION TIME C. ------------NOTE------------
Only applicable to Functions 14 and 23.
One or more Functions with one required channel inoperable C.1 Restore required channel to OPERABLE status 30 days D. On or more Functions with two or more required channels inoperable D.1 Enter the Condition referenced in Table 3.3.3-1 for the channel.
7 days E. Required Action and associated Completion Time of Condition C or D not met.
E.1 Enter the Condition referenced in Table 3.3.3-1 for the channel.
Immediately F. As required by Required Action E.1 and referenced in Table 3.3.3-1.
F.1 Be in MODE 3.
AND F.2 Be in MODE 4.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 12 hours Given the vendor situation described above, and the fact that a large portion of the WR Neutron Flux channel cannot be repaired or replaced online, the seven-day Completion Time (CT) for certain failure modes would not provide sufficient time to restore a Neutron Flux channel.
Additionally, should the second Neutron Flux channel at CNP Unit 2 become inoperable, CNP Unit 2 would be required to shutdown if one of the channels cannot be restored within seven days.
2.3.3 RG 1.97, Revision 3, Pressurized Water Reactors Variable Types RG 1.97 defines three types of variables for the purpose of selecting accident monitoring instrumentation and applicable criteria. These types are:
Type A: Those variables that provide primary information needed to permit the control room operating personnel to take the specified manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for DBA events Type B: Those variables that provide information to indicate whether plant safety functions are being accomplished.
Type C: Those variables that provide information to indicate the potential for being breached or the actual breach of the barriers to fission products release (i.e. fuel cladding, primary coolant pressure boundary, and containment)
RG 1.97 also defines two additional types for monitoring the operation of systems important to safety and radioactive effluent release. Two additional variable types are:
Type D: Those variables that provide information to indicate the operation of individual safety systems and other systems important to safety.
Type E: Those variables to be monitored as required for use in determining the magnitude of the release of radioactive materials and for continuously assessing which releases.
RG 1.97 states that A variable included as Type A does not preclude it from being included as Type B, C, D, or E, or vice versa.
Table 3, PWR Variables of RG 1.97, Revision 3, provides design and qualification criteria for the instrumentation used to measure the various variables. The criteria are separated into three separate Categories that provide a graded approach to requirements depending on the importance to safety of the measurement of a specific variable.
Category 1 provides the most stringent requirements and is intended for key variables.
Category 2 provides less stringent requirements and generally applies to instrumentation designated for indicating system operating status.
Category 3 is intended to provide requirements that will ensure that high-quality, off-the-shelf instrumentation is obtained and applies to backup and diagnostic instrumentation.
Table 3 below summarizes the information provided in Table 3 of RG 1.97, Revision 3, for the PAM instrumentation variables relevant to this LAR:
Table 3: PAM Instrumentation Types, Categories, and Purpose Variable Type Category Purpose Neutron Flux B
1 Reactivity Control Function detection; Accomplishment of mitigation Variable Type Category Purpose RCS Hot Leg Water Temperature B
1 Core Cooling Function detection; Accomplishment of mitigation; Verification; Long-term surveillance RCS Cold Leg Water Temperature B
1 Core Cooling Function detection; Accomplishment of mitigation; Verification; Long-term surveillance Core Exit Temperature - Cooling B
3 Core Cooling Verification Core Exit Temperature - Cladding C
1 Fuel Cladding Verification 2.3.4 Initial Proposed Change In the January 26, 2023, application, the licensee proposed to reclassify the WR Neutron Flux instrumentation at CNP Units 1 and 2 as Category 3, and to completely remove this function from the CNP TS Table 3.3.3-1, Post Accident Monitoring Instrumentation. The licensee proposed that the RG 1.97, Revision 3, neutron flux safety function regarding reactivity control be fulfilled by the RCS hot leg water temperature, RCS cold leg water temperature, and CET instrumentation since these parameters are also Type B variables and meet the requirements of Category 1 instrumentation. The Neutron Flux instruments would be reclassified as a Category 3 instrumentation for the purpose of verification of reactivity control.
The licensee indicated that during an accident that involves adverse containment conditions, where the neutron flux monitoring equipment may be rendered inoperable, control room operators would still be able to monitor core and RCS temperature performance by CETs, RCS hot leg temperatures, and RCS cold leg temperatures. Additionally, the shutdown margin would be verified by measuring boron concentration. The EOPs would also direct the operators to assure that boric acid injection is taking place, adding negative reactivity to ensure that the core remains shut down.
NRC Staff Concerns Regarding the Initial Proposal The NRC staff previously disapproved attempts to change the categorization of this PAM Function by downgrading the neutron flux monitoring channel from Type B, Category 1 to Type B, Category 3. In the staffs review of Topical Report (TR) WCAP-15981-NP-A, Post Accident Monitoring Instrumentation Re-Definition for Westinghouse NSSS Plants, (ML103560687), the NRC staff stated, in part:
WCAP-15981-NP recommends that Neutron Flux (Source Range) be reclassified as Type B Category 3. In response to the NRC staff's Request for Additional Information (RAI) concerning providing an early indication of a return to criticality, the PWROG [Pressurized Water Reactor Owners Group] discussed monitoring boron concentration. The information provided does not support the determination that the monitoring of boron concentration via non-Category 1 instrumentation would provide adequate information about a potential return to criticality. Therefore, the NRC staff finds it unacceptable to reclassify Neutron Flux (Source Range) on a generic basis.
Further, the staff stated, in part:
TR WCAP-15981-NP does not discuss instrumentation to be used to provide an early indication of a return to criticality. Neutron Flux (Source Range) instrumentation provides this information. In a letter dated August 22, 2007
[ML072360096], the PWROG provided additional clarification to support the RAI responses documented in a letter dated June 28, 2007. The letter dated August 22, 2007 [ML072360096], states that RCS Boron concentration provides information to ensure adequate shutdown margin. However, TR WCAP-15981-NP has not proposed that RCS Boron Concentration be upgraded to a Category 1 variable in lieu of Neutron Flux (Source Range). Based on the information provided, the NRC staff does not agree with the proposed reclassification of Neutron Flux (Source Range) and concludes that Neutron Flux (Source Range) should be included in the PAM TS.
In the January 26, 2023, application, the licensee referenced safety evaluations (SEs) regarding conformance to RG 1.97, for R.E. Ginna Nuclear Power Plant (ML17264A259), Indian Point Nuclear Generating Station, Unit 2 (ML100351277), and Beaver Valley Power Station, Unit 1 (ML20094L410), as precedent for the initial request. The NRC staff notes that, in those SEs, it did not approve to change the categorization of an existing Type B, Category 1 variable to a Type B, Category 3 variable. Rather, the staff granted limited approvals of a PAM commitment plan submitted in response to GL 82-33, Supplement 1 to NUREG-0737 - Emergency Response Capability, dated December 1982 (ML051680455).
2.3.5 Final Proposed Change In the February 27, 2024, supplement, the licensee revised the scope of the LAR such that Neutron Flux monitoring remains as Function 1 in TS Table 3.3.3-1, but a footnote is added to reflect that Neutron Flux instruments are not required to meet EQ criteria.
The following text will be added to CNP, Unit Nos. 1 and 2, TS Table 3.3.31 (addition in bold):
FUNCTION REQUIRED CHANNELS CONDITION REFERENCED FROM REQUIRED ACTION E.1
- 1. Neutron Flux 2(a)
F (a) Channels used to satisfy Function 1 are not required to be environmentally qualified Existing footnotes (a) through (g) will be renumbered as footnotes (b) through (h) to reflect the addition of new footnote (a). The licensee also provided proposed changes to the TS Bases for information only. The proposed change to the TS Bases reflects the proposed change to the TS.
NRC approval of the proposed TS Bases change is not required.
3.0 TECHNICAL EVALUATION
The primary purpose of PAM instrumentation is to display unit variables that provide information necessary for the control room operators during accident situations. In a DBA, Neutron Flux instrumentation provides information to control room operators in two situations: (1) to verify that the reactor is no longer critical, and (2) to monitor the core for unexpected additions of reactivity after reactor shutdown has been achieved.
3.1 Instrumentation To meet PSDC Criterion 12, indications of plant variables are needed by the control room operators during accident situations to: (1) provide information needed to permit the operator to take pre-planned manual actions to accomplish safe shutdown; (2) determine whether the reactor trip, engineered safety feature systems, manually-initiated safety systems, and other systems important to safety are performing their intended functions; (3) provide information to the operators that will enable them to determine the potential for a gross breach of the barriers to radioactivity release; and (4) to determine if a gross breach of a barrier has occurred.
RG 1.97 describes five types of variables (Types A, B, C, D, and E) for the purpose of aiding in selecting accident-monitoring instrumentation and applicable criteria. Types A, B and C, are relevant to this amendment request. There are three categories of instruments based on the design and qualification provisions (i.e., Category 1, 2 and 3). Design and qualification provisions for Category 1 and Category 2 instruments are detailed in Table 1 of RG 1.97.
Category 2 instrument provisions are similar to the provisions for Category 1 instruments with some noted differences, such as seismic qualification and redundancy for Type B instruments not being specified. Power supply for Category 2 instruments should be from a highly reliable power supply but does not have to be from a standby power source (e.g., diesel generators).
The guidelines in NUREG-1431, Standard Technical Specifications - Westinghouse Plants, Volume 2, Bases, Revision 5, September 2021 (ML21259A159) state that PAM instruments classified as Type A or Category 1 should be included in limiting conditions for operation.
Type A instruments satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) because they are linked to operator actions in mitigating an accident, and Category 1 instruments that are not Type A may satisfy Criterion 4 of 10 CFR 50.36(c)(2)(ii) because they are important to reducing public risk.
During an accident that involves adverse containment conditions, where the Neutron Flux instruments may be rendered inoperable, the licensee states that the control room operators would still be able to monitor core and RCS temperature performance by CETs, RCS hot leg temperatures and RCS cold leg temperatures. At CNP, the CETs and RCS hot and cold leg temperatures (Wide Range) instruments are Type A, Category 1, variable. Type A and Category 1 variables at CNP meet RG 1.97, Category 1, design and qualification provisions for seismic and environmental qualification, single failure criterion, utilization of emergency standby power, immediately accessible display, continuous readout, and recording of display.
3.2 PAM - Verification of Shutdown After an accident, a reactor trip is initiated either automatically or manually, and emergency core cooling system (ECCS) systems initiate automatically in response to the accident. One of the post-accident requirements is to verify that the reactor is no longer critical. Currently, the WR Neutron Flux Instrumentation is the Category 1 variable provided to verify reactor shutdown.
In the January 26, 2023, application, the licensee states, in part:
EOP E-0, Reactor Trip or Safety Injection, is the entry point for all postulated and analyzed accident conditions, with the single exception of a complete loss of all alternating current electrical power. The Loss of Power scenario does not have an adverse impact on the containment atmosphere, and thus would not challenge the availability of the Gamma-Metrics wide range neutron flux instrumentation.
During an accident that involves normal containment conditions, the Neutron Flux instrumentation is expected to be available for use, regardless of whether it meets EQ criteria.
Therefore, for events that do not result in a harsh environment inside of containment (e.g., normal reactor trip, steam generator tube rupture, anticipated transient without SCRAM, etc.), the non-environmentally qualified Neutron Flux instrumentation will continue to provide information to control room operators to support verification that the reactor is no longer critical.
The NRC staff concurs that this is a reasonable conclusion because the inside-of-containment portions of the Neutron Flux instrumentation would not be subjected to a harsh environment condition that would affect its operability.
During an accident that involves adverse containment conditions, where the Neutron Flux instrumentation may be rendered inoperable, focus is placed on the completion time of EOP E-0, Step 1, Check Reactor Trip. In the January 26, 2023, application, the licensee states, in part:
The first four steps of EOP E-0, including Check Reactor Trip, are Immediate Actions, and are performed by control room operators from memory.
Performance of the Immediate Actions is not a timed evolution per the Time Critical Operator Actions (TCOA) program. The shortest defined TCOA is timed from the reactor trip until auxiliary feedwater is isolated, which occurs at Step 8 of E-0. The required time for this action is six minutes, and validation of the actions have shown that the action has historically occurred between four and five minutes. This timing includes the Immediate Actions and a subsequent review of each of those actions by the unit supervisor using the hard copy of the procedure. As such, a one minute expectation for completion of the first immediate action step is reasonable and conservative.
For events that do involve the release of energy into containment (e.g., MSLB Inside Containment Accident or LOCA), it is expected that non-environmentally qualified Neutron Flux instrumentation would not be adversely impacted within the first minute and will continue to provide information to control room operators to support verification that the reactor is no longer critical. The NRC staff concurs that this is a reasonable conclusion because confirmation that a significant reduction in neutron flux generation has occurred during reactor shutdown following control rod insertion when using the Neutron Flux instruments should reasonably take less than a minute. During this period, it is reasonable to expect that a significant degradation of the neutron flux detectors and cabling would not have had time to occur.
3.3 PAM - Monitoring Potential Return to Criticality After an accident, a reactor trip is initiated either automatically or manually, and ECCS systems initiate automatically in response to the accident. One of the post-accident requirements is to maintain subcriticality. Currently, the WR Neutron Flux instrumentation is the Category 1 variable used to monitor the core for unexpected additions of reactivity after reactor shutdown has been achieved.
As discussed in Section 3.2 of this safety evaluation, during an accident that involves normal containment conditions, the Neutron Flux instrumentation is expected to be available for use, regardless of whether it meets EQ criteria. Therefore, the Neutron Flux instrumentation will continue to provide information to control room operators to support monitoring the core for unexpected additions of reactivity after reactor shutdown. The NRC staff concurs that this is a reasonable conclusion since the harsh environmental conditions that could degrade performance of the portion of the Neutron Flux instrumentation that is inside the containment would not have occurred.
The licensee has proposed to not require the WR Neutron Flux instruments to meet EQ criteria.
As such, during an accident that involves adverse containment conditions, the NRC staff assumed that WR Neutron Flux instrumentation would be rendered inoperable after the initial verification of reactor shutdown described in section 3.2 above and cannot be relied upon to function for the duration of the event. The licensee indicates that the only credible accident conditions that would result in adverse containment conditions are a MSLB and LOCA (inside containment). In the August 15, 2024, supplement, the licensee indicated that:
the MSLB accident is the only DBA where the reactor core is assumed and analyzed for a return to criticality; the postulated return to criticality during a MSLB event is short-lived and there is no consequential damage to the core; the MSLB analysis assumes no credit for operator action; in all LOCA scenarios the core remains subcritical following the reactor trip and automatic injection of boric acid by the ECCS; and none of the DBAs described in the CNP USFAR, including MSLBs and LOCAs, credit the use of the WR Neutron Flux instruments.
The NRC staff reviewed CNP USFAR Section 14.2 (ML22340A148 (Unit 1); ML22340A193 (Unit 2)) for the quantitative MSLB analyses, as well as USFAR Section 14.3 (ML22340A216 (Unit 1); ML22340A133 and ML22340A164 (Unit 2)) for the quantitative LOCA analyses, and concurs that the WR Neutron Flux instruments are not credited for use in the DBAs.
Nevertheless, to meet the requirements of GDC 19 and the guidelines of RG 1.97, Revision 3, the licensee is required to have the ability to monitor the core for unexpected additions of reactivity after reactor shutdown has been achieved. For events that result in an adverse containment environment, the licensee proposed that CET temperature and RCS hot and cold leg temperatures should be the key variables used to monitor subcriticality following initial reactor shutdown.
3.4 Use of CET Temperature and RCS Hot and Cold Leg Temperatures CNP UFSAR, Section 7.8, Post-Accident Monitoring Instrumentation (ML22340A198),
Table 7.8-1 (ML22340A197) lists the display instrumentation provided to the control room operators to enable them to perform required manual safety functions, and to assess plant and environmental conditions during, and following, an accident. Table 7.8-1 lists Neutron Flux (B-1), RCS hot leg water temperature (B-5), RCS cold leg water temperature (B-6), and Core Exit Temperature (B-8) as RG 1.97, Revision 3, Type B variables.
Currently, the WR Neutron Flux instrumentation is the Category 1 variable used to: (1) verify that the reactor is no longer critical, and (2) monitor the core for unexpected additions of reactivity after reactor shutdown has been achieved. The licensee proposes that the WR Neutron Flux instrumentation only be required to verify that the reactor is no longer critical, which the staff understands would apply in the initial period following shutdown. The RCS hot and cold leg temperatures are Category 1 variables used to determine RCS subcooling margin.
CETs are Category 1 variables normally relied upon to determine RCS subcooling margin.
The licensee proposes that the requirement of monitoring the core for unexpected additions of reactivity after reactor shutdown has been achieved be fulfilled by the CETs and the RCS hot and cold leg temperatures. Since CETs and the RCS hot and cold leg temperatures are also Category 1 variables, it would also be necessary for them to meet the stringent EQ requirements that are currently required for the WR Neutron Flux instruments.
Additionally, NUREG-0800, BTP 7-10, indicates that the use of non-environmentally qualified WR Neutron Flux instruments is acceptable provided that: (1) qualified CETs and RCS hot and cold leg temperature indications are provided in conjunction with directions in EOPs for operator action in order to ensure that boric acid injection is occurring (discussed in Section 3.5 of this Safety Evaluation, below), and (2) RCS conditions following DBAs with energy release to the containment would enable proper process measurement capabilities of the CETs and RCS hot and cold leg temperature indications. Since the licensee has proposed the use of qualified CETs and RCS hot and cold leg temperatures for monitoring design basis events that are accompanied by energy releases inside the containment and has shown that for some types of events the CETs will be functional, and for other types of events, the RCS hot leg and cold leg temperature indications will be functional, the NRC staff has determined that the guidance in BTP 7-10 has been adequately addressed. The NRC staff bases this conclusion on its evaluation of the licensees analysis that for RCS LOCA events the CETs would still be functional and would detect increases in core heat due to re-criticality in the core, and that for non-RCS LOCA events, the RCS hot leg and cold leg temperature instruments would still be functional and would be able monitor increases in RCS temperature.
3.5 Emergency Operating Procedures (EOPs)
During an accident that involves adverse containment conditions, where the WR Neutron Flux instrumentation may be rendered inoperable, the licensee indicated that the control room operators would still be able to monitor core and RCS temperature performance by CETs, RCS Hot Leg Temperatures and RCS Cold Leg Temperatures. Additionally, the shutdown margin would be verified by measuring boron concentration. The EOPs would also direct the operators to assure that boric acid injection is taking place, adding negative reactivity to ensure that the core remains shut down.
In the January 26, 2023, application, the licensee stated, in part:
While temperature and boron concentration are not direct measurements of neutron flux, it is important to consider that the CSFSTs [Critical Safety Function Status Trees] are only intended to determine if an imminent threat to a critical safety function exists. The CSFSTs act as a decision point in order to determine if the operator should immediately suspend the performance of optimum recovery procedures in order to address this challenge. A continuous challenge does not exist unless the core is producing enough power that a temperature rise is indicated on the RCS temperature instruments.
In the August 2, 2023, supplement, the licensee stated that the CSFST for subcriticality (procedure F-0.1), directs operators to use the WR Neutron Flux instruments to monitor for unexpected additions of reactivity after reactor shutdown has been achieved. Per procedure F-0.1, operators must enter function restoration procedure OHP-4023-FR-S.1, Response to Nuclear Power Generation (FR-S.1) and OHP-4023-FR-S.2, Response to Loss of Core Shutdown, (FR-S.2) to add negative reactivity to a core that is observed to be critical when expected to be shutdown (including via emergency boration). Both FR-S.1 and FR-S.2 direct operators to verify boration flow using instruments QFl-410 and QFI-411. Both QFl-410 and QFI-411 instruments are powered from vital instrument power supplies and will be available during all postulated accident conditions, including the boric acid transfer pump running lights.
In the August 2, 2023 supplement, the licensee also stated, in part:
Operators are trained to establish stable conditions for long-term heat removal in the event of an accident scenario. Once a stable balance of temperatures and flows has been established, any changes that occur that were not initiated or expected by control room operators will be promptly investigated and corrected, per established training and the guidance of procedures in progress.
In addition, in the August 2, 2023 supplement, the licensee described the impact on operators if the WR Neutron Flux instruments are not available (i.e., during an event that results in a harsh environment in containment). Specifically, the licensee stated, in part:
During the evaluation of the current conditions at CNP Unit 2, with one train of WR log power not able to meet its environmental qualification requirements, it was observed that when performing FR-S.1, operators rely on Gamma-Metrics as the sole input in the decision to transition back to optimal recovery procedures. In the event that both Gamma-Metrics are rendered inoperable, this procedural requirement could result in operators not exiting FR-S.1 even if all other relevant parameters indicate that the reactor is shut down. As such, training has been provided to operators, and a procedure change is being processed, outside of the requested change, to allow control room operators to determine reactor condition using an aggregate evaluation of several instruments, including core exit thermocouple temperatures, hot and cold leg temperatures, any indicating nuclear instruments, and reactor boron concentration. If an aggregate indication review determines that the reactor is subcritical, control room operators will be directed to continue boration to maintain adequate shutdown margin and return to the procedure and step in effect In an accident scenario that results in an adverse containment, WR neutron flux instrumentation, if not environmentally qualified, cannot be relied upon to function for the duration of the event. In an accident scenario where WR neutron flux instruments are rendered inoperable, operators would be directed to enter FR-S.1 since at that point WR log power less than 5% cannot be confirmed.
Control room operators, as directed by FR-S.1, would initiate actions to add negative reactivity to the core, including emergency boration, and would remain in FR-S.1 until it can be determined, using an aggregate indication review, that the reactor is subcritical.
Based on the above, the staff concludes that if neutron flux indication is available, and it indicates that an unexpected addition of negative reactivity is in process, procedure F-0.1 will direct the operators to enter FR-S.1 to initiate actions to add negative reactivity to the core. The operators will need to remain in that procedure until they determine the reactor is subcritical. If neutron flux indication is not available, then the crew cannot use it to determine whether the criteria of F-0.1 are met for reactivity control, and thus the operators will be directed by F-0.1 to enter FR-S.1 to initiate actions to add negative reactivity to the core. They will need to remain in that procedure until they determine the reactor is subcritical. Accordingly, operators must enter FR-S.1 to ensure the reactor is shutdown if either: (1) neutron flux indication is available and indicates that an unexpected addition of positive reactivity is in process, or (2) neutron flux indication is not available.
As part of implementation of the license amendment, the licensee will revise FR-S.1 to direct the operators to use core exit thermocouple temperatures, RCS hot and cold leg temperatures, any indicating nuclear instruments, and reactor boron concentration levels to evaluate whether the reactivity control safety function is met. These indications are already provided in the control room, and thus operators are already familiar with them. Additionally, in the January 26, 2023, application, the licensee stated that an internal training request has been generated for refresher training for licensed operators to discuss the various other indications that can be used in the absence of both channels of wide range nuclear instrumentation. The NRC staff concludes that providing training on what operators should use when Neutron Flux instrumentation is unavailable, as directed by FR-S.1, is reasonable to ensure that operators will know what indications to use and what actions to take when Neutron Flux instrumentation is not available.
In the August 2, 2023, supplement, the licensee provided more information about the ECCS and stated, in part:
If the emergency core cooling system (ECCS) is in service via a safety injection signal, then the flow rates of each set of injection pumps are available for the operator to monitor for boric acid injection as follows:
Boron injection flowpath (flow instruments) - 1FI-51, 1FI-52, 1FI-53, 1FI-54 (high head, per loop injection)
Safety injection flowpath (flow instruments) - 1FI-260, 1FI-266 (medium head)
Residual heat removal flowpath (flow instruments)- 1FI-310, 1FI-311, 1FI-320, 1FI-321 (low head)
Each of these flow rates is monitored (by plant operators while operating) within the EOPs following a safety injection actuation. Additionally, monitor lights are provided at the top of the control panels to indicate valve positions for the various signals to ensure that ECCS flowpaths are properly aligned.
The NRC staff notes that when the ECCS is in service, the operators can monitor the boric acid injection by observing the presence of a safety injection signal, as well as monitoring the flow rates of each set of injection pumps follow the boron injection flowpath (high head, per loop injection), the safety injection flowpath (medium head), and the residual heat removal flowpath (low head).
Main Steam Line Break The licensee indicated that the fundamental characteristic of a MSLB is a rapid cooldown and depressurization of the intact RCS due to the uncontrolled heat removal via the high blowdown steam flow out of the break.
In the August 2, 2023, supplement, the licensee indicates that the operator training would allow control room operators to rely on an aggregate indication of plant instruments, including core exit thermocouples, RCS hot and cold leg temperatures, and reactor boron concentration, as they continue to evaluate for subcriticality., The licensee also stated that [w]ith the CETs located inside the reactor vessel and mounted directly above the active fuel region within the coolant flow path, any rise in heat output from the fuel assemblies will be seen by the CETs within seconds of the change.
Loss of Coolant Accident In the August 2, 2023, supplement, the licensee provided an overview of the expected response of the CET instruments regarding their adequacy to enable operators to take appropriate mitigative actions to recover from an accident, and to avoid further adverse consequences of the event. The licensee described the fundamental characteristics of a LOCA, and stated that:
, throughout a LOCA event, WR log power alone is typically inadequate for early detection of criticality due to the unreliable indication caused by a voided RCS. Control room operators are aware of this limitation and are trained to monitor multiple indications and to assess for aggregate for indications of criticality. These include CET trends, RCS hot and cold leg temperature indications, and boron sample results. ECCS flows and temperatures will be monitored by control room operators, and, due to the position of the CETs, a change in core exit temperatures would be observed quickly. Control room staff would conclude that a rising temperature trend that is not accompanied by a reduction in ECCS flow or ECCS cooling flow would be considered an indicator of a reactivity concern requiring operator action during all accident conditions.
In reviewing the licensees August 2, 2023, supplement, the NRC staff did not have sufficient information regarding how: (1) the RCS hot and cold leg temperature instruments response would enable plant operators to take timely actions (especially in a case where the reactor coolant pumps fail to operate following an accident with energy release to the containment); and (2) the combination of a) the boron concentration and the assurance of boron injection instrumentation indication; b) the response time characteristics of instrumentation used for responding to a change in boron concentration and for assuring boron injection is taking place, and c) the use of instrumentation measuring boron concentration would enable plant operators to take timely mitigative action in an event of a return to criticality following a LOCA event.
Accordingly, on November 17, 2023, the staff sent a request for additional information to the licensee (ML23321A122).
In response to the staffs request, in the February 27, 2024, supplement, the licensee stated that boron concentration would not be considered as a key variable with regard to subcriticality but would be used by control room operators to assess shutdown margin as part of an aggregate indication review. The licensee also stated that continuous indication of boron concentration via instrumentation is not available at CNP Unit 1 or Unit 2; rather RCS boron concentration is measured by sampling the RCS. Specifically, the licensee stated, in part:
The operator follows Step 7 of Emergency Operating Procedure OHP-4023-E-0, Reactor Trip or Safety Injection, to perform Attachment A of this procedure which systematically reviews expected equipment response to start pumps and align equipment that is required for the accident conditions. This includes ensuring that ECCS pumps are operating with a flow path from the RWST
[Refueling Water Storage Tank] to the RCS through their respective injection flow paths. This is confirmed by observing pump running currents, valve positions, and injection flow indication on the control panels.
While Emergency Core Cooling Flow is a Type A, Category 1 variable included in CNP Unit 1 and Unit 2 TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, as Function 20, it is not considered a key variable with regards to subcriticality.
The NRC staff finds the licensees explanation regarding ECCS flow not being a key variable with regard to subcriticality is acceptable since the boron injection from the accumulators and the RWST occurs through three intact loops, with one loop conservatively assumed to be unavailable due to a break for all DBAs.
In the February 27, 2024, supplement, the licensee also stated, in part:
During post-accident recovery with the RCS intact, and in a situation where Gamma-Metrics instruments are not available, the control room operators are trained and directed by emergency operating procedures to monitor RCS temperature indication as a key variable to identify any postulated return to criticality and rising core power level. One or more indications of RCS Temperature, including CET Temperature, RCS Hot Leg Temperature, and RCS Cold Leg Temperature, would be available to control room operators. These key parameters are all included in CNP Unit 1 and Unit 2 TS Table 3.3.3-1, PAM Instrumentation, and conform to the design and qualification requirements of a Category 1 variable, in accordance with the requirements of RG 1.97, Revision 3.
In the February 27, 2024, supplement, the licensee described the process or procedure that plant operators would use to interpret potentially anomalous neutron source range readings by using information from the reactor vessel level indication system (RVLIS) regarding the location of potential reactor vessel voiding. The licensee also provided a description of the process or procedure that plant operators would use to interpret potentially anomalous neutron source range readings by using information from the RVLIS regarding the location of potential reactor vessel voiding. However, the NRC staff found that it did not have sufficient information to make a determination as to whether the RCS hot leg and cold leg instruments would be sufficiently responsive to provide the information to plant operators in a timely manner. Accordingly, on June 26, 2024, the staff sent a request for additional information to the licensee (ML24178A043).
In response to the staffs request, in the August 15, 2024, supplement, the licensee indicated that for the quantitative analyses of the CETs, RCS hot leg and cold leg temperature instrument responses during postulated post-accident conditions are not considered feasible. The licensee instead clarified the physical arrangement of the CET, WR RCS hot leg Resistance Temperature Detectors (RTDs), and WR RCS cold leg RTDs for CNP Unit 1 and Unit 2, and further discussed their expected response during the MSLB and LOCA accidents. Additionally, the licensee indicated that the proposed change to TS Table 3.3.3-1 does not impact the availability or functionality of the PAM WR Neutron Flux instruments for accidents that do not result in adverse containment conditions. The licensee identified MSLB Inside Containment and LOCA as the two accidents from its UFSAR that can cause adverse conditions in the containment where non-environmentally qualified WR Neutron Flux instruments cannot be relied upon. The licensee further stated that the accident analysis in the CNP UFSAR for the SLB and LOCA events does not credit WR Neutron Flux instruments.
As discussed in Section 3.3 of this SE, the NRC staff finds that in all design basis LOCA scenarios the core will remain subcritical following the reactor trip and injection of boric acid by the ECCS. The NRC staff also reviewed the sample calculation performed by the licensee in the supplement dated February 27, 2024, which demonstrated that the dilution rate for postulated boron dilution event following a LOCA event would be relatively slow. In the February 27, 2024, supplement, the licensee states, in part:
Based on the conservative assumptions described above, and a 20 gpm [gallons per minute] leak of unborated water, it is calculated that it would take nearly 300 gallons of water to dilute the recirculation inventory by 1 ppm [part per million], and the dilution would occur at roughly 4.1 ppm per hour. At this rate of boron dilution, the time duration to raise power from 0% [percent] to 5% [percent]
on a hypothetical return to criticality would be on a time scale of one to several hours.
The NRC staff finds the sample calculation provided to be acceptable because any such dilution would require very large leaks and would be a relatively slow event for a hypothetical return to criticality. The NRC staff further concludes that such a slow dilution rate would also provide plant operators significant time to diagnose and mitigate the dilution prior to a return to criticality occurring. The NRC staff also notes that a postulated boron dilution event concurrent with a LOCA is not a DBA assumed in CNP UFSAR.
With regard to the MSLB analysis, where the reactor core is analyzed for return to criticality for post-accident scenarios, the CNP UFSAR analysis shows that a postulated return to criticality for such events is short lived, and that there is no consequential damage to the core or its cooling capabilities. The NRC staff notes that for such an event, long term subcriticality is maintained using boron injection from the ECCS system subsequent to an automatic reactor trip. The NRC staff finds the licensees response to be acceptable based on the analysis presented in the CNP UFSAR, and the fact that any postulated return to power event is self-limiting due to temperature feedback (i.e., cold water injection would cause power to increase, which would cause the hot leg temperature to increase, thereby causing cold leg temperatures to increase and warm water to be returned to the reactor, leading to negative reactivity in the core, and causing power to decrease).
In the August 15, 2024, supplement, the licensee described the applicable RCS Temperature instrument arrangements that are listed in TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, and stated, in part:
CNP Unit 1 and Unit 2 both have a four-loop RCS with a single WR hot leg RTD and single WR cold leg RTD in each of the four loops. The WR RCS RTDs are located in thermowells that extend into the cold and hot leg piping. The WR cold leg RTDs are downstream of the Reactor Coolant Pump (RCP) discharge. While each of the four loops has WR RTDs, only the hot leg and cold leg RTDs in Loops 1 and 3 are credited by CNP Unit 1 and Unit 2 TS 3.3.3 for post-accident monitoring. In addition to WR RTDs, CNP Unit 1 and Unit 2 also have narrow range hot leg and cold leg RTDs installed in each of the four loops. While these narrow range RTDs do support the reactor protection system, they are not credited by TS 3.3.3 for post-accident monitoring.
The CNP Unit 1 and Unit 2 in-vessel instrumentation systems provides up to 65 CETs, positioned to measure fuel assembly coolant outlet temperature at preselected locations. Thermocouples are threaded into guide tubes that penetrate the reactor vessel head through seal assemblies and terminate in the upper core support assembly above the exit flow end of the fuel assemblies. The CETs are approximately 1.3 feet above the active fuel line, which is below the bottom of the cold leg and hot leg vessel nozzles. Note that CNP Unit 1 and Unit 2 TS 3.3.3 for post-accident monitoring only require the use of eight total CETs at a given time (two for each of the four core quadrants).
In the August 15, 2024, supplement, the licensee also discussed the RCS Temperature Instrument response during a MSLB and stated, in part:
the WR RCS hot leg and cold leg RTDs are judged to be responsive to a return to power given that the bulk fluid circulating from the core through the RCS loops is subcooled during post-SLB conditions and the observation that the narrow range RCS loop RTDs are responsive for post-[MSLB] OP T [Overpower Delta-Temperature] // OT T [Overtemperature Delta-Temperature] reactor protection. CETs are also judged to be as responsive or more responsive to changes in core temperature during post-[MSLB] conditions than the WR RCS cold leg and hot leg RTDs In the August 15, 2024, supplement, the licensee also discussed the RCS Temperature Instrument response following a LOCA, and stated, in part:
Initially during LTCC [Long Term Core Cooling], ECCS injection flow is directed through any intact cold legs, into the downcomer and upwards through the core which is covered by the two-phase mixture flow. The CETs, which are positioned closely above the top of the fuel (the CET tips are at an elevation lower than the bottom of the cold leg and hot leg vessel nozzles), are judged to be thermally responsive to changes in the two-phase core exit temperature and steaming, given that the CETs will be effectively submerged in the two-phase mixture.
Between 5.5 and 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the initiation of the LOCA, the ECCS is realigned to supply water to the RCS hot legs in order to control boric acid concentration in the reactor vessel. During this LTCC hot leg recirculation, the flow direction of the core is effectively reversed, and therefore CETs are not expected to be as responsive to changes in core temperature, given that at this point the CETs would be primarily measuring the temperature of the recirculated ECCS flow going into the reactor core. If a rising core power due to a dilution event is postulated, then it is expected to impact the entirety or a majority of the core, due to the core being a well-mixed environment in this scenario. Therefore, the CETs would be expected to respond to steam released from bulk boiling in the core.
Based on the above analysis, the NRC staff determined that the CETs are expected to respond to steam released from bulk boiling in the core for most postulated events (i.e., MSLB and LOCA). However, it would be expected that the RCS hot leg temperature and cold leg temperature instruments would be responsive to events that do not involve a large break LOCA affecting the integrity of the hot legs or cold legs within the reactor coolant pressure boundary.
Therefore, the NRC staff concluded that the RCS hot leg and cold leg instruments could be relied on to provide post-accident monitoring information for MSLB events, but would not be consistently reliable for all types of LOCA events.
Additionally, the NRC staff determined that at CNP, the control room operators are trained and directed by EOPs to monitor RCS temperature indication as a key variable to identify any postulated return to criticality and rising core power level in a situation where WR Neutron Flux instruments are not available. The CETs would provide useful PAM information for most types of events, and for other types of events, the RCS hot leg and cold leg temperature instruments would be capable of providing information to the plant operators to take timely action for accident events that do not result in large LOCAs affecting the integrity of the RCS hot legs and cold legs.
Therefore, the NRC staff is satisfied that available operable instrumentation (either CETs, or RCS hot and cold leg temperatures) would be sufficient for operators to monitor the core for unexpected additions of reactivity after reactor shutdown has been achieved for design basis events that result in energy release to the primary containment. Depending on the type of event, either the CETs or the RCS hot and cold leg temperature instruments would have the capability to support determination of appropriate EOP manual actions when the WR Neutron Flux instruments are inoperable due to the adverse environment. Based on above evaluation, the NRC staff concludes that the capability of PAM instruments provide reasonable assurance that the necessary post-accident functions will be accomplished while WR Neutron Flux instruments are inoperable due to the existence of a harsh environment. In sum, the NRC staff found that when the CETs and RCS hot and cold leg temperatures are used in place of the WR Neutron Flux instruments, adequate instrumentation would be available to monitor the core for unexpected additions of reactivity after reactor shutdown has been achieved following a DBA.
3.6 Technical Specifications The current CNP TS 3.3.3, Post Accident Monitoring (PAM) Instrumentation, states that if two channels of WR Neutron Flux instruments become inoperable, Required Action D.1 is to restore all but one channel to operable status within seven days. If that required action is not completed within the associated completion time, Required Action F.1 is to be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in Mode 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. In the February 27, 2024, supplement, the licensee proposed to modify CNP TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, to add a footnote to Function 1, Neutron Flux, to read as follows:
Channels used to satisfy Function 1 are not required to be environmentally qualified The licensees proposed footnote to Function 1 would provide a remedial action to allow the use of OPERABLE CETs and RCS hot and cold leg temperature to satisfy the PAM requirement of monitoring the core for unexpected additions of reactivity after reactor shutdown has been achieved, and would avoid the Required Action F.1. The NRC staff reviewed the proposed footnote and determined that it continues to meet the requirements of 10 CFR 50.36(c)(2)(ii) because, consistent with NUREG-0800, BTP 7-10, the action assigns PAM requirements to other Category 1 variables. Therefore, the NRC staff finds the proposed change to TS 3.3.3, Function 1, is acceptable.
3.7 Technical Conclusion The proposed changes contained in this LAR permit the use of CETs and RCS hot and cold leg temperature instruments to monitor the core for unexpected additions of reactivity after reactor shutdown has been achieved should the WR Neutron Flux instruments become unavailable following a DBA with energy release to the containment. As discussed above. the NRC staff has determined that the WR Neutron Flux instruments will still be required to be available in order to verify that the reactor is no longer critical immediately following a reactor trip or safety injection regardless of the containment environment. For events that do not result in an adverse containment environment, WR Neutron Flux instruments will still be required to be available in order to monitor subcriticality following the initial reactor shutdown.
For events that result in an adverse containment environment, the NRC has determined that, consistent with NUREG-0800, BTP 7-10, either the CET temperature or the RCS hot and cold leg temperatures will be the key variables used to monitor subcriticality following initial reactor shutdown, depending on whether the event is a MSLB or a LOCA. The CETs and RCS hot and cold leg temperature instruments are classified as Type A, Category 1, and are subject to the qualification provisions for such components. The NRC staff has determined that monitoring the core for unexpected additions of reactivity after reactor shutdown has been achieved can be satisfied by either the CET temperature or RCS hot and cold leg temperature instrument channels, depending on the type of design basis event (MSLB or LOCA) that has occurred.
These instrument channels would be used along with credited operator actions identified in the emergency operating procedures for the post-accident monitoring of RCS integrity.
Based on the above evaluation, the NRC staff finds that the licensees proposed addition of the footnote to TS Table 3.3.3-1 continues to provide the capability to detect, and indicate in the control room, unexpected increases in reactor core activity which could lead to a significant abnormal degradation of the reactor coolant pressure boundary, is consistent with CNP PSDC 11, 12, and 13, and the TS continue to meet 10 CFR 50.36(c)(2) and 10 CFR 50.34(f)(2).
Therefore, the NRC staff concludes that the proposed changes are acceptable.
4.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
NOTICES AND ENVIRONMENTAL FINDINGS RELATED TO AMENDMENT NOS. 364 AND 345 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-58 AND DPR-74 INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-315 AND 50-316 Application:
Safety Evaluation Date:
January 26, 2023 ADAMS Accession No. ML23026A284 Supplements:
August 2, 2023 (ML23214A289)
February 27, 2024 (ML24058A357)
August 15, 2024 (ML24228A161)
December 18, 2024
1.0 INTRODUCTION
Indiana Michigan Power Company requested changes to the technical specifications for Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, by license amendment request. The amendment revises TS 3.3.3, Post Accident Monitoring (PAM) Instrumentation to not require environmental qualification to be maintained for TS 3.3.3, Function 1, Neutron Flux, instrumentation.
2.0 STATE CONSULTATION
In accordance with the Commissions regulations, the State of Michigan official was notified of the proposed issuance of the amendment on October 17, 2024. The State official had no comments.
3.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in Title 10 of the Code of Federal Regulations (10 CFR) Part 20. The NRC staff has determined that the amendment involves 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 as published in the Federal Register on March 21, 2023 (88 FR 17036) and April 16, 2024 (89 FR 26946), 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.
ML24297A130 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DEX/EICB/BC NAME SWall SRohrer FSacko DATE 10/23/2024 10/24/2024 10/24/2024 OFFICE NRR/DSS/SCPB/BC NRR/DSS/SNSB/BC NRR/DSS/STSB/BC NAME MValentin (HWagage for)
PSahd SMehta DATE 11/12/2024 10/24/2024 11/11/2024 OFFICE NRR/DRO/IOLB/BC OGC - NLO NRR/DORL/LPL3/BC(A)
NAME JAnderson STurk IBerrios DATE 11/01/2024 12/10/2024 12/16/2024 OFFICE NRR/DORL/LPL3/PM NAME SWall DATE 12/18/2024