ML22046A233
| ML22046A233 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 06/21/2022 |
| From: | Scott Wall Plant Licensing Branch III |
| To: | Gebbie J Indiana Michigan Power Co |
| Wall S | |
| References | |
| EPID L-2021-LLA-0050 | |
| Download: ML22046A233 (26) | |
Text
June 21, 2022 Mr. Joel P. Gebbie 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. 360 AND 342 REGARDING CHANGE TO THE TECHNICAL SPECIFICATION REQUIREMENT FOR CONTAINMENT WATER LEVEL INSTRUMENTATION (EPID L-2021-LLA-0050)
Dear Mr. Gebbie:
The U S Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment Nos. 360 and 342 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 March 23, 2021, as supplemented by letter dated December 16, 2021.
The amendment revises TS 3.3.3, Post Accident Monitoring (PAM) Instrumentation to allow one channel of TS 3.3.3, Function 7, Containment Water Level, to be satisfied by a train of two operable containment water level switches in the event that both containment water level channels become inoperable. This alternate method of satisfying containment water level channel requirements would be limited to the remaining duration of the operating cycle each time it is invoked.
J. Gebbie 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. 360 to DPR-58 2.
Amendment No. 342 to DPR-74 3.
Safety Evaluation 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. 360 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 March 23, 2021, as supplemented by letter dated December 16, 2021, 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. 360, 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 Nancy L. Salgado, 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: June 21, 2022 Robert F.
Kuntz Digitally signed by Robert F. Kuntz Date: 2022.06.21 06:39:06 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 360 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: 359, 360 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. 360, 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 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
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(a)
F
- 8.
Containment Pressure (Narrow Range) 2 F
- 9.
Penetration Flow Path Containment Isolation Valve Position 2 per penetration flow path(b)(c)
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(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 (a) 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.
(b) 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.
(c) Only one position indication channel is required for penetration flow paths with only one installed control room indication channel.
(d) 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 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(e)
F
- 20.
Emergency Core Cooling System Flow (per train) 2(f)
F
- 21.
Containment Pressure (Wide Range) 2 F
- 22.
Refueling Water Storage Tank Level 2
F
- 23.
RCS Subcooling Margin Monitor 1(g)
F
- 24.
Component Cooling Water Pump Circuit Breaker Status 2
G
- 25.
Containment Recirculation Sump Water Level 2
F (e) 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.
(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.
(g) 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. 342 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 March 23, 2021, as supplemented by letter dated December 16, 2021, 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. 342, 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 the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Nancy L. Salgado, 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: June 21, 2022 Robert F.
Kuntz Digitally signed by Robert F. Kuntz Date: 2022.06.21 06:39:48 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 342 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. 341, 342 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. 342, 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 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
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(a)
F
- 8.
Containment Pressure (Narrow Range) 2 F
- 9.
Penetration Flow Path Containment Isolation Valve Position 2 per penetration flow path(b)(c)
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(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 (a) 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.
(b) 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.
(c) Only one position indication channel is required for penetration flow paths with only one installed control room indication channel.
(d) 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 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(e)
F
- 20.
Emergency Core Cooling System Flow (per train) 2(f)
F
- 21.
Containment Pressure (Wide Range) 2 F
- 22.
Refueling Water Storage Tank Level 2
F
- 23.
RCS Subcooling Margin Monitor 1(g)
F
- 24.
Component Cooling Water Pump Circuit Breaker Status 2
G
- 25.
Containment Recirculation Sump Water Level 2
F (e) 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.
(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.
(g) 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. 360 AND 342 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
1.0 INTRODUCTION
By application dated March 23, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21082A496), as supplemented by letter dated December 16, 2021 (ML21350A176), Indiana Michigan Power Company (I&M, the licensee),
requested changes to the technical specifications (TSs) for the Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2 (CNP).
The amendment revises TS 3.3.3, Post Accident Monitoring (PAM) Instrumentation to allow one channel of TS 3.3.3, Function 7, Containment Water Level, to be satisfied by a train of two operable containment water level switches in the event that both containment water level channels become inoperable. This alternate method of satisfying containment water level channel requirements would be limited to the remaining duration of the operating cycle each time it is invoked. Although the licensee also requested approval for a proposed change to the TS Bases, the proposed change to the TS Bases reflects the proposed change to the TS, and NRC approval of the proposed TS Bases change is not necessary.
On May 18, 2021, the NRC published a notice of consideration of approval of the application in the Federal Register (86 FR 26950). The supplemental letter dated December 16, 2021, provided additional information that expanded the scope of the application as originally noticed.
Therefore, the NRC published a notice of consideration of approval of the application, as supplemented, in the Federal Register on April 19, 2022 (87 FR 23270).
2.0 REGULATORY EVALUATION
2.1 Background
The current TS 3.3.3 requires two channels of containment water level instrumentation to be operable. The associated TS Bases define these channels as the lower containment water level monitors, each supplemented by two level switches. In the event that both containment water level channels become inoperable, the limiting conditions for operation (LCO) for TS 3.3.3 would require the unit to shut down if one channel cannot be restored within seven days. The licensee states that due to the lead times involved in calibration, refurbishment, or replacement of lower containment water level instrumentation, and because certain portions of the instrumentation cannot be repaired or replaced online, it is possible that a containment water level channel may not be restored to be operable within the required seven days period.
The proposed change to the TS would allow one channel of TS 3.3.3, PAM instrumentation, Function 7, Containment Water Level, to be satisfied by a train of two operable containment water level switches in the event that both containment water level channels become inoperable, because the containment water level switches will provide the relevant PAM information required by control room operators. This alternate method of satisfying containment water level channel requirements would be limited to the remaining duration of the operating cycle each time it is invoked.
The proposed change to the TS would also reflect the instrumentation available to the control room operators should both the normal containment water level indicators become unavailable during the operating cycle and would allow the licensee to pursue resolution of an inoperable containment water level monitor without the increased risk associated with an unscheduled shut down.
2.2
System Description
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 the containment water level. Per 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), this variable is classified as Type A, Category 1. Type A, Category 1, instruments should be redundant and backed by uninterrupted power supply, and to meet the specified environmental qualifications. During a design basis accident, containment water level instrumentation provides information to control room operators in three situations: to check if the reactor coolant system (RCS) is intact; to determine if containment water level is greater than the minimum required to begin cold leg recirculation; and to indicate that water level is approaching containment flooding level.
The CNP TS Bases define the containment water level channels as Lower Containment Water Level Monitors NLl-320 and NLl-321. The lower containment water level monitors are capable of measuring from 599 feet (') 3 inches (") elevation to 614' elevation, which correspond to the containment floor level and maximum flood level, respectively. Each lower containment water level is supplemented by two level switches which provide indication in the control room when the water level exceeds the associated setpoint. One level switch actuates when the containment water level exceeds 602 2 3/4 (Nll-330 and NLl-331). The second level switch actuates when containment level reaches 613' 0" (NLl-340 and NLl-341). The low level switch indicates when sufficient water level exists in containment to switch the emergency core cooling system (ECCS) suction source from the refueling water storage tank to the containment recirculation sump, while the high level switch confirms whether or not the containment water level is approaching its design basis value.
2.3 Regulatory Requirements and Guidance 2.3.1 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 or 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: (1) safety limits, limiting safety system settings; (2) LCOs; (3) surveillance requirements; (4) design features; and (5) administrative controls.
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.
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 an 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.
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.
The NRC regulations codified in 10 CFR, Appendix A, General Design Criteria for Nuclear Power Plants, of 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, contain the following General Design Criteria (GDC):
GDC 13, Instrumentation and control, states, Instrumentation 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 the CNP units were issued in 1969, before the 1971 final rule originally 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:
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 below. Also, as discussed in Section 1.4.10 of the CNP updated final safety analysis report (ML21125A606), the CNP became obligated to conform with GDC 19 after initial licensing of CNP to the PSDCs.
2.3.2 Regulatory Guidance and Updated Final Safety Analysis Report Criteria NUREG-0737, Clarification of TMI [Three Mile Island] Action Plan Requirements, November 1980 (ML051400209), incorporates all TMl-related items approved for implementation by the Commission up to that point. Enclosure 1, Post-TMI Requirements for Operating Reactors, Item II.F.1, Attachment 5, pertains to containment water level.
RG 1.97, Revision 3, 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 Updated Final Safety Analysis Report (UFSAR), Section 1.4, Plant Specific Design Criteria (PSDC) (ML21125A606), 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.
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.
In accordance with PSDC 11, CNP is equipped with control rooms that meet the intent of GDC 19.
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.
In accordance with PSDC 12, instrumentation and controls are provided to monitor and control during normal as well as post-accident conditions which meets the intent of GDC 13.
3.0 TECHNICAL EVALUATION
The primary purpose of PAM instrumentation is to display unit variables that provide information required by the control room operators during accident situations. In a design basis accident, containment water level instrumentation provides information to control room operators in three situations: to check if the RCS is intact, to determine if containment water level is greater than the minimum required to begin cold leg recirculation, and to indicate that water level is approaching containment flooding level.
3.1 Instrumentation To meet PSDC 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, and manually-initiated safety systems and other systems important to safety are performing their intended functions; and (3) provide information to the operators that will enable them to determine the potential for causing a gross breach of the barriers to radioactivity release and 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; two of these, Types A and B, are relevant to this amendment request. Type A variables are those variables to be monitored that provide primary information needed to permit the control room operators 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 design basis accident events. Type B variables are those variables that provide information to indicate whether plant safety functions are being accomplished. There are three categories (Category 1, 2, and 3) of instruments based on the design and qualification provisions. 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 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, (ML12100A228) 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.
The relevant instrumentation available to measure water level in containment is summarized in the table below:
Instrument ID Description Elevation RG 1.97 Function NLA-310, NLl-311 Containment Sump Water Level Monitors 589' 5" to 599' 8" Type B, Category 2 Used by operators to check if the RCS is intact; no LCO Function NLl-320, NLl-321 Containment Water Level Monitors 599' 3" to 614" Type A, Category 1 LCO 3.3.3 Function 7, Containment Water Level, used by operators to check if the RCS is intact NLl-330, NLl-331 Containment Water Level Switches (lower) 602' 2 3/4" Type A, Category 1 Used by operators to verify sufficient water in containment to switch to recirculation NLl-340, NLl-341 Containment Water Level Switches (upper) 613' 0" Type A, Category 1 Used by operators to identify containment water level approaching containment flooding Containment water level is a Type A, Category 1, variable provided for determination of adverse containment conditions. 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. In a design basis accident, containment water level instrumentation provides information to control room operators in three situations: to check if the RCS is intact, to determine if containment water level is greater than the minimum required to begin cold leg recirculation, and to indicate that water level is approaching containment flooding level. Both the containment water level monitors (NLI-320 and NLI-321) as well as the containment water level switches (NLI-330 and NLI-331 for lower level and NLI-340 and NLI-341 for upper level) for CNP are Type A and Category 1 and are fully qualified.
As noted in the table above, containment sump level monitors NLA-301 and NLI-311 are used by operators to check if the RCS is intact. These monitors are Type B, Category 2, per RG 1.97. Despite being considered Category 2 instruments, NLA-310 and NLl-311 are safety-related components and are environmentally and seismically qualified, with Class 1E power supplies. The containment sump water level monitors are subject to the design and qualification criteria associated with Category 1 instruments.
3.2 PAM - Check If RCS Is Intact 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 determine whether the RCS is intact or not. Containment radiation monitors, containment pressure, and containment water level, are primarily used make this determination. In order to determine if RCS is intact, the licensee simulated a larger break loss of coolant accident (LOCA) without a loss of offsite power (LOOP) as well as in conjunction with LOOP. One of the normal containment water level monitors NLI-320 was in inoperable mode prior to the start of simulation while the second monitor NLl-321 was set to fail prior to the step in the emergency operating procedures where operators diagnose whether the RCS is intact. In the LAR, the licensee stated that in the large break LOCA simulation, control room operators were able to identify that the RCS was not intact based on the containment radiation monitor response and containment pressure rise. Operators did not refer to containment water level indication for diagnosis. In the simulation of a large break LOCA with a loss of offsite power, containment pressure rise was used to transition from the diagnosis to the appropriate recovery procedure. Containment water level instruments were referenced, but containment sump water level (NLA-310 and NLl-311) were used for redundant indication. Loss of both containment water level instruments (NLl-320 and NLl-321) did not affect the diagnosis and response to a LOCA.
3.3 PAM - Minimum Recirculation Level If during the course of accident recovery, it becomes necessary to switch the ECCS suction source from the refueling water storage tank to the containment recirculation sump, then prior to switching to recirculation, operators can verify that adequate water level exists inside containment. After checking the availability of adequate water by checking the status of containment water level switches (NLl-330 and NLl-331), the control room operators can align the recirculation pump suction to the containment recirculation sump.
Containment water level switches NLl-330 and NLl-331 meet Category 1 design and qualification provisions. In the event both lower containment water level monitors become inoperable, indication of minimum recirculation level in containment will continue to be available to control room operators as discussed in Section 3.4 below.
3.4 Use of Containment Water Level Monitors Containment water level is a Type A variable that provides primary information to control room operators to take specific manual actions for which there is no automatic control, and those actions are required for safety systems to achieve their safety function following a design basis accident. These level monitors are currently the only water level instrumentation that span the full potential range of post-accident containment water level. In addition to assisting in determining if the RCS is intact, the containment water level monitors are used to verify that the necessary water inventory has been retained in containment to support a subcooled recirculation recovery. If the inventory is not adequate, the operators would proceed with a saturated recovery. The high-level containment water level switches would provide backup indication that full containment water level is present, but the function would not be necessary because the plant could still be brought to cold shutdown conditions using the saturated recovery procedures.
The associated level switches that indicate low and high water level containment water level are used normally for switching the recirculation pump suction from refueling water storage tank to containment sump at low level and to determine if the containment water level may be approaching the high water level corresponding to the full design flood level. These switches are also Type A qualified level switches. These level switches are proposed to be used for the containment water level monitoring function if both of the containment water level monitors become unavailable or inoperable, and use of the level switches would only be allowed until the end of the operating cycle in which the switches are invoked.
The containment sump water level monitors are available to the operators to use as backup level indicators. In addition, the containment particulate and gaseous radiation monitors, containment pressure instruments, and containment sump water level monitor would continue to provide control room operators the information needed to evaluate whether the RCS is intact following a manual or automatic actuation of a reactor trip or safety injection.
3.5 Emergency Operating Procedures (EOPs)
The relevant instrumentation available to measure water level in containment are discussed in Section 3.1 above. In addition, the following instruments related to containment conditions are also available to allow for safe operation of the plant following a design basis accident:
Instrument ID Description Range RG 1.97 Function VRA-1310, VRA-1410 (Unit 1)
VRA-2310, VRA-2410 (Unit 2)
Containment Area Radiation - High Range Monitors 1 R/Hr to 1X107 R/Hr Type A, LCO 3.3.3 Function 10; Used by operators for determination of adverse containment PPP-300, PPP-301, PPP-302, PPP-303 Containment Pressure - Narrow Range Monitors
-5 psig to
+12 psig Type A, LCO 3.3.3 Function 8; Manually establish or trip containment spray PPA-310, PPA-312 Containment Pressure - Wide Range Monitors
-5 psig to
+36 psig Type B LCO 3.3.3 Function 21; Used by operators to check if the RCS is intact The containment water level switches provide continuous indication in the sense that the lights associated with the containment water level switches stay on as long as the water level remains above the setpoint and go off when the water level falls below the setpoint.
In the event that both containment water level monitors (NLI-320/NLI-321) are inoperable:
a.
Containment water level up to elevation 599 8 would still be available to control room operators from the containment sump water level monitors (NLA-310/NLI-311). The availability of these instruments is administratively controlled by the licensees Technical Requirements Manual. A containment water level at or above 599 8 would be sufficient to indicate significant water on the containment floor, an indication to control room operators that the RCS is no longer intact in the event of a LOCA b.
Containment area radiation monitors (VRA-1310/VRA-1410/VRA-2310/VRA-2410) and containment pressure instruments (PPP-300/PPP-301/PPP-302/PPP-303), would continue to provide control room operators the information needed to evaluate whether the RCS is intact following a manual or automatic actuation of a reactor trip or safety injection. Availability of these instruments is controlled by TSs.
c.
Containment level switches (NLI-330, NLI-331) indicate when sufficient water level exists in containment to switch the ECCS suction source from the refueling water storage tank to the containment recirculation sump.
d.
Containment level switches (NLI-340, NLI-341) confirm whether or not the containment flood level is approaching its design basis value: 614 (Unit 1), 613.5 (Unit 2).
The above discussion indicates that while both the lower containment water level monitors are inoperable:
The PAM instrumentation provides operators a sufficient understanding of containment water level, from the containment floor to the maximum design basis value, following a design basis accident.
The PAM instrumentation provides operators sufficient information to allow for safe operation of the plant following a design basis accident.
In the December 16, 2021, supplement, the licensee reviewed the CNP EOPs that use instrument indications from the containment water level monitors (NLl-320, NLl-321) as a decision input to identify manual actions during Scenario 1, Check If RCS Is Intact, and Scenario 2, Subcooled Versus Saturated Recovery. The Scenario 1 function of checking if the RCS is intact is a qualitative check of several diverse instruments, with no specific parameter value driving an operator action. The normal containment sump water level monitors (NLA-310, NLl-311) are currently one of the specified diverse indicators and, along with the gaseous and particulate radiation monitors, the most likely to indicate a minor loss in RCS integrity. The containment pressure instruments provide a back-up indication for the containment water level monitor for larger losses of RCS integrity necessary for containment water level to come within range of the level monitors (NLl-320, NLl-321). The Scenario 2 determination of recovery path uses the containment water level monitor in combination with the RWST level indications to determine if a subcooled or saturated recovery should be used, either of which are acceptable procedural methods for placing the plant in cold shutdown following a loss of coolant accident.
Although the high containment water level switches (NLI-340, NLI-341) would not enter into the operator decision, they would support a determination that the water inventory had been retained within containment. The one EOP function requiring a specific containment water level value for initiation of an operator action is the switching from RWST injection to containment sump recirculation. The licensee stated that the lower containment water level switches (NLI-330, NLI-331) rather than the containment water level monitors (NLl-320, NLl-321) are used to verify that containment water level is sufficiently high due to the increased accuracy of the level switches. Therefore, the NRC staff is satisfied that available operable instrumentation would be sufficient to support determination of appropriate EOP manual actions when both containment level monitors (NLI-320/NLI-321) are inoperable for the remainder of an operating cycle.
Based on above analysis and evaluation, the capability of PAM instruments and EOPs provide reasonable assurance that the necessary post-accident functions will be accomplished while the containment level channels (NLI-320 and NLI-321) are inoperable. When a train of containment water level switches (NLl-330 and NLl-340 or NLl-331 and NLl-341) is used in place of one containment water level channel, the staff found that adequate instrumentation would be available to allow safe operation of the plant following a design basis accident.
3.6 Technical Specifications In the December 16, 2021, supplement, the licensee proposed to modify CNP TS Table 3.3.3-1, Post Accident Monitoring Instrumentation, to add a footnote to Function 7, Containment Water Level, to read as follows:
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.
The current CNP TS 3.3.3, Post Accident Monitoring (PAM) Instrumentation, states that if two channels of containment water level 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 />. The proposed footnote to Function 7 would provide a remedial action to allow the use of an OPERABLE train of containment water level switches (NLl-330 and NLl-340 or NLl-331 and NLl-341) to satisfy 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)(i) by providing a remedial action to allow continued safe operation using an operable train of containment water level switches until the end of the current operating cycle.
Therefore, the NRC staff finds the proposed change to TS 3.3.3, Function 7, acceptable.
3.7 Technical Conclusion The proposed changes permit use of level switches should the TS-specified level monitors become unavailable during plant operation. This change is only permitted until the end of the operating cycle in which the switches are invoked. The lower containment water level switches are the normal instruments supporting the operator action to initiate sump recirculation, and the high containment water level switch provides information that design containment post-accident water inventory has been retained. The containment water level monitors directly support a qualitative check of RCS integrity that includes consideration of several diverse instrument inputs that provide redundancy. The water level monitor also provides information for a decision point regarding specific post-accident recovery actions where acceptable recovery actions would still be implemented with the instrument unavailable. Both the normal water level monitors and the water level switches are classified as Type A, Category 1, and subject to the qualification provisions for such components. The NRC staff has determined that the necessary post-accident containment water level monitoring can be met by the level switches and credited operator actions for the post-accident monitoring of RCS integrity, minimum recirculation level, and manual actions subsequent to a design basis accident. The NRC staff reviewed the proposed changes as well as the licensees justifications for the changes. The staff finds that the proposed change meets the criteria of 10 CFR 50.36(c)(2) and RG 1.97, Revision 3. The NRC staff, therefore, finds the proposed addition of the footnote to TS Table 3.3.3-1 acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the State of Michigan official was notified of the proposed issuance of the amendments on January 21, 2022. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change the requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration as published in the Federal Register on June 16, 2020 (85 FR 36432) and April 19, 2022 (87 FR 23270), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: H. Vu, NRR N. Chien, NRR S. Jones, NRR T. Sweat, NRR Date of issuance: June 21, 2022
- via memo OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DEX/EICB/BC*
NRR/DSS/SCPB/BC(A)*
NAME SWall SRohrer MWaters SJones DATE 02/15/2022 02/16/2022 11/09/2021 01/03/2022 OFFICE NRR/DSS/STSB/BC OGC - NLO NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME VCusumano MSpencer NSalgado (RKuntz for)
SWall DATE 02/15/2022 06/07/2022 06/08/2022 06/21/2022