ML24225A002
| ML24225A002 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/03/2024 |
| From: | Scott Wall NRC/NRR/DORL/LPL3 |
| To: | Lies Q Indiana Michigan Power Co |
| Wall S | |
| References | |
| EPID L-2024-LLA-0040, TS 3.8.1 | |
| Download: ML24225A002 (1) | |
Text
September 3, 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. 363 AND 344 REVISING TECHNICAL SPECIFICATIONS SECTION 3.8.1, AC SOURCES-OPERATING, FOR A ONE-TIME EXTENSION OF A COMPLETION TIME (EPID L-2024-LLA-0040)
Dear Q. Shane Lies:
The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment Nos. 363 and 344 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 license and technical specifications (TSs) in response to your application dated April 3, 2024, as supplemented by letters dated May 16, 2024, July 2, 2024, July 15, 2024, and August 28, 2024.
The amendments revise TS 3.8.1, AC [Alternating Current] Sources - Operating, by adding a footnote for TS 3.8.1, Required Action A.3 to allow a one-time completion time extension from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br /> to support the replacement of the 12AB (Train B) Loop Feed Enclosure and associated bus for the Train B reserve feed preferred power source.
Q.
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. 363 to DPR-58
- 2. Amendment No. 344 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. 363 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 April 3, 2024, as supplemented by letters dated May 16, 2024, July 2, 2024, July 15, 2024, and August 28, 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. 363, 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.
FOR THE NUCLEAR REGULATORY COMMISSION Jeffrey A. Whited, 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: September 3, 2024 BRENT BALLARD Digitally signed by BRENT BALLARD Date: 2024.09.03 13:54:41 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 363 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 3.8.1-2 3.8.1-2 Renewed License No. DPR-58 Amendment No: 362, 363 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. 363, 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,
AC Sources - Operating 3.8.1 Cook Nuclear Plant Unit 1 3.8.1-2 Amendment No. 287, 291, 363 ACTIONS
NOTE-----------------------------------------------------------
LCO 3.0.4.b is not applicable to DGs.
CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite circuit inoperable.
A.1
NOTE--------------
Not applicable if a required Unit 2 offsite circuit is inoperable.
Perform SR 3.8.1.1 for required OPERABLE offsite circuit.
AND A.2 Declare required feature(s) with no offsite power available inoperable when its redundant required feature(s) is inoperable.
AND A.3 Restore required offsite circuit to OPERABLE status.
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> AND Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from discovery of no offsite power to one train concurrent with inoperability of redundant required feature(s) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />(a)
AND 17 days from discovery of failure to meet LCO 3.8.1.a or b (a) For Train B only, the Completion Time that Train B can be inoperable as specified by Required Action A.3 may be extended beyond the "72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />" up to "288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br />," to support modification of the Train B Reserve Feed 12AB Loop Feed Enclosure. Upon completion of the modification and restoration this footnote is no longer applicable. Prior to entry into the 288-hour extended Completion Time, the Supplemental Diesel Generators (SDGs) shall be verified as available. During the 288-hour extended Completion Time, the SDGs shall be verified as available once per shift. If the SDGs becomes unavailable after the initial 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> while in the extended 288-hour Completion Time period, it shall be made available within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or the unit shall 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 be in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This 24-hour period will be allowed only once within the single extended Completion Time. Compensatory measures described within CNP letter AEP-NRC-2024-02, dated April 3, 2024, and supplemented by AEP-NRC-2024-40, dated May 16, 2024, will remain in effect during the extended period. The one-time extension shall expire upon completion of the modification and restoration of operability for Train B.
INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-316 DONALD C. COOK NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 344 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 April 3, 2024, as supplemented by letters dated May 16, 2024, July 2, 2024, July 15, 2024, and August 28, 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. 344, 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.
FOR THE NUCLEAR REGULATORY COMMISSION Jeffrey A. Whited, 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: September 3, 2024 BRENT BALLARD Digitally signed by BRENT BALLARD Date: 2024.09.03 13:56:08 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 344 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 3.8.1-2 3.8.1-2 Renewed License No. DPR-74 Amendment No. 343, 344 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. 344, 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
AC Sources - Operating 3.8.1 Cook Nuclear Plant Unit 2 3.8.1-2 Amendment No. 269, 273, 344 ACTIONS
NOTE-----------------------------------------------------------
LCO 3.0.4.b is not applicable to DGs.
CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite circuit inoperable.
A.1
NOTE--------------
Not applicable if a required Unit 1 offsite circuit is inoperable.
Perform SR 3.8.1.1 for required OPERABLE offsite circuit.
AND A.2 Declare required feature(s) with no offsite power available inoperable when its redundant required feature(s) is inoperable.
AND A.3 Restore required offsite circuit to OPERABLE status.
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> AND Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from discovery of no offsite power to one train concurrent with inoperability of redundant required feature(s) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />(a)
AND 17 days from discovery of failure to meet LCO 3.8.1.a or b (a) For Train B only, the Completion Time that Train B can be inoperable as specified by Required Action A.3 may be extended beyond the "72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />" up to "288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br />," to support modification of the Train B Reserve Feed 12AB Loop Feed Enclosure. Upon completion of the modification and restoration this footnote is no longer applicable. Prior to entry into the 288-hour extended Completion Time, the Supplemental Diesel Generators (SDGs) shall be verified as available. During the 288-hour extended Completion Time, the SDGs shall be verified as available once per shift. If the SDGs becomes unavailable after the initial 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> while in the extended 288-hour Completion Time period, it shall be made available within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or the unit shall 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 be in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This 24-hour period will be allowed only once within the single extended Completion Time. Compensatory measures described within CNP letter AEP-NRC-2024-02, dated April 3, 2024, and supplemented by AEP-NRC-2024-40, dated May 16, 2024, will remain in effect during the extended period. The one-time extension shall expire upon completion of the modification and restoration of operability for Train B.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 363 AND 344 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 (i.e., initial and supplements)
Safety Evaluation Date April 3, 2024, ADAMS Accession No. ML24094A288 Supplements May 16, 2024 (ML24137A221)
July 2, 2024 (ML24184A140)
July 15, 2024 (ML24197A126)
August 28, 2024 (ML24241A161)
September 3, 2024 Principal Contributors to Safety Evaluation Edmund Kleeh Khoi Nquyen Thinh Dihn Charles Moulton Michael Swim Robert Elliott
1.0 INTRODUCTION
Indiana Michigan Power Company (I&M, the licensee) requested changes to the technical specifications (TSs) for Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2 (CNP), by license amendment request (application or LAR). The amendments would revise TS 3.8.1, AC
[Alternating Current] Sources - Operating, by adding a footnote for TS 3.8.1, Required Action A.3 to allow a one-time completion time (CT) extension from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br /> to support the replacement of the 12AB (Train B) Loop Feed Enclosure (LFE) and associated bus for the Train B reserve feed preferred power source. The replacement of the Train B LFE and associated bus does not require NRC approval.
The supplemental letters dated May 16, 2024, July 2, 2024, July 15, 2024, and August 28, 2024, provided additional information that clarified the application, did not expand the scope of the application as originally noticed and did not change the U.S. Nuclear Regulatory Commission (NRC, the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on June 11, 2024 (89 FR 49241).
2.0 REGULATORY EVALUATION
AND GUIDANCE 2.1
System Description
TS Bases B 3.8.1, AC Sources - Operating (ML22340A160 (Unit 1), and ML22340A202 (Unit 2)), states, in part:
The onsite Class 1E [alternating current (AC)] Distribution System is divided into redundant load groups (trains) so that the loss of any one load group does not prevent the minimum safety functions from being performed. Each train has connections to a preferred offsite power source, an alternate offsite power source, an auxiliary source (unit auxiliary transformer), and a single DG [Diesel Generator].
In Section 2.1, System Design and Operation, of the LAR, the licensee illustrates that the onsite AC electric power distribution system for each unit contains four 4.16 kilovolt (kV) non-safety-related buses (reactor coolant pumps (RCP) buses) with each RCP bus feeding a downstream safety-related 4.16 kV bus. The four safety-related buses are referred to as the T buses. With the main generator on-line, the RCP buses are normally fed from the Unit Auxiliary Transformers (UATs), which receive power from the main generator. When the main generator is offline, each units RCP and "T" buses are powered by the preferred offsite power sources two reserve auxiliary transformers (RATs). The preferred offsite power sources for each unit can be arranged so that Main Switchyard transformer No. 4 or No. 9 supplies one RAT and one engineered safety features (ESF) train, and Main Switchyard transformer No. 5 or No. 9 supplies the other RAT and ESF train. One train consists of two T buses and their ESF loads.
Under certain plant conditions, it is possible for transformer No. 4, No. 5, or No. 9 to feed the entire plant auxiliary load.
The alternate offsite circuit supplies a 69/4.16 kV transformer (TR 12EP-1) and the 4.16 kV EP Bus which in turn supplies each of the T buses. This alternate offsite source is 30 degrees out of phase with the plant 4.16 kV buses and is manually connected to the T buses. The alternate offsite power source has the necessary capacity to operate one train of the ESF loads in one unit while simultaneously operating one ESF train in the other CNP unit. Two supplemental diesel generators (SDGs) are provided for CNP to also power the EP bus and function the same as alternate offsite source in the number of trains per unit supplied.
The T buses can also be powered from the emergency diesel generators (EDGs). For each unit there are two EDGs with one EDG supplying two T buses per train. TS limiting condition for operation (LCO) 3.8.1.a requires two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System be operable.
2.2 Reason for the Proposed Change In Section 2.3, Reason for Proposed Change, of the LAR, the licensee stated that it intends to completely replace the Train B LFE and associated bus for the Train B reserve feed preferred power source for CNP with a cable bus design. TS Bases B 3.8.1 states, in part, that Unit 1 and Unit 2 RATs TR101AB and TR201AB supply Unit 1 and Unit 2 Train B safety-related loads.
Therefore, during replacement of the 12AB LFE, Train B in both units will be without power for the duration of the modification. The licensee indicates that the modification improves 12AB LFE reliability which has experienced moisture intrusion due to its inadequate design. Due to the scope of the modification, the 12-day duration will exceed the current CT of TS 3.8.1, Condition A of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for restoration of one offsite circuit. Therefore, the licensee requests a one-time extension of TS 3.8.1, Condition A CT from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 12 days (288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br />).
2.3 Proposed TS 3.8.1 Changes The licensee proposed to add a footnote to modify the Allowable Outage Times (AOTs) of TS 3.8.1, Required Action A.3. This action requires restoration of both offsite circuits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. For the action, a note (a) would be added. The footnote would state:
(a) For Train B only, the Completion Time that Train B can be inoperable as specified by Required Action A.3 may be extended beyond the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> up to 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br />, to support modification of the Train B Reserve Feed 12AB Loop Feed Enclosure. Upon completion of the modification and restoration this footnote is no longer applicable. Prior to entry into the 288-hour extended Completion Time, the Supplemental Diesel Generators (SDGs) shall be verified as available. During the 288-hour extended Completion Time, the SDGs shall be verified as available once per shift. If the SDGs becomes unavailable after the initial 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> while in the extended 288-hour Completion Time period, it shall be made available within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or the unit shall 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 be in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This 24-hour period will be allowed only once within the single extended Completion Time. Compensatory measures described within CNP letter AEP-NRC-2024-02, dated April 3, 2024, and supplemented by AEP-NRC-2024-40, dated May 16, 2024, will remain in effect during the extended period. The one-time extension shall expire upon completion of the modification and restoration of operability for Train B.
2.4 Applicable Regulatory Requirements and Guidance The NRC staff applied the following regulatory requirements and guidance documents for review of the LAR.
Title 10 of the Code of Federal Regulations (10 CFR) section 50.36, Technical specifications, paragraph (c)(2) requires, in part, that the applicants for a license authorizing operation of a production or utilization facility include in their application proposed TSs that specify LCOs. The regulation at 10 CFR 50.36(c)(2)(i) states, in part, that LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.
Section 10 CFR 50.63, Loss of all alternating current power, requires, in part, that a nuclear power plant shall be able to withstand for a specified duration and recover from a complete loss of offsite and onsite alternating current (AC) sources (i.e., a station blackout (SBO)).
Section 10 CFR 50.92(a) states, in part, that in determining whether to grant an amendment to a license, the Commission will be guided by the considerations which govern the issuance of initial licenses or construction permits to the extent applicable and appropriate. Both the common standards for licenses and constructions permits in 10 CFR 50.40(a), and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public.
Section 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 17, Electric Power Systems, states, in part:
An onsite electric power system and an offsite electric power system shall be provided to permit functioning of structures, systems, and components important to safety. The safety function for each system (assuming the other system is not functioning) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents. The onsite electric power supplies, including the batteries, and the onsite electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.
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 CNP Updated Final Safety Analysis Report (UFSAR), Section 1.4, Plant Specific Design Criteria (PSDC) (ML22340A150), 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 39, Emergency Power, states:
An emergency power source shall be provided and designed with adequate independency, redundancy, capacity, and testability to permit the functioning of the engineered safety features and protection systems required to avoid undue risk to the health and safety of the public. This power source shall provide this capacity assuming a failure of a single active component.
In accordance with PSDC 39, power supplies are arranged to meet the intent of GDC 17.
The NRC staff used the following guidance documents while reviewing the LAR:
NUREG-0800 (Standard Review Plan) Branch Technical Position (BTP) 8-8, Onsite (Emergency Diesel Generators) and Offsite Power Sources Allowed Outage Time Extension (ML113640138). The BTP 8-8 provides guidance for reviewing requests for AOT extensions for the onsite and offsite electrical power sources to perform online maintenance of the power sources.
Regulatory Guide (RG) 1.177, Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, January 2021, Revision 2 (ML20164A034), describes an acceptable risk-informed approach for assessing proposed changes to TS AOTs, also known as CTs.
RG 1.200, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, December 2020, Revision 3 (ML20238B871), describes one acceptable approach for determining whether the technical adequacy of the probabilistic risk assessment (PRA), in total or the parts that are used to support an application, is sufficient to provide confidence in the results such that the PRA can be used in regulatory decisionmaking for light-water reactors.
The terms CT and AOT are used interchangeably in the LAR and the TSs to describe the amount of time for which a limiting condition for operation may not be met as long as the prescribed remedial actions in the TSs are followed until the condition can be met, in accordance with 10 CFR 50.36(c)(2)(i).
3.0 TECHNICAL EVALUATION
The NRC staff evaluated the proposed TS changes considering both deterministic and risk-insights to determine if the licensee justified continued operation until the completion of the maintenance, or 12 days (288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br />) from the start of the AOT, whichever occurs earliest.
3.1 Deterministic Evaluation The NRC staff used the guidance in BTP 8-8 to perform the deterministic review of the CNP LAR and its supplements. The licensee supplemented its application to address certain aspects of BTP 8-8 that were not discussed in the LAR. The staff evaluated those supplements as shown below:
3.1.1 Supplemental Power Source Capacity The guidance in NUREG-0800, BTP 8-8, states, in part:
The supplemental source must have the capacity to bring a unit to safe shutdown (cold shutdown) in case of a loss of offsite power (LOOP) concurrent with a single failure during plant operation (Mode 1).
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Multi-unit sites that have installed a single AAC [alternate AC] power source for SBO [station blackout] cannot substitute it for the inoperable diesel when requesting AOT extensions unless the AAC source has enough capacity to carry all LOOP loads to bring the unit to a cold shutdown as a substitute for the EDG in an extended AOT and carry all SBO loads for the unit that has an SBO event without any load shedding.
In the May 16, 2024, supplement, the licensee states, in part:
The two SDGs are rated to a combined capacity of 4500kW, which is above the 3500kW rating of each EOG (CNP has 2 EDGs per Unit, one for Train A and one for Train B). This capability is verified in CNP Calculation 12-E-S-EPEDG-001, Allowable Outage Time (AOT) Diesel Generator ETAP Analysis. The two SDGs provide sufficient capacity to bring a Unit to safe shutdown in case of a LOOP event with a single failure during plant operation (Mode 1).
The NRC staff notes that the SDGs have sufficient combined capacity to take the units to cold shutdown (MODE 5) in case of a LOOP concurrent with a single failure during plant operation.
Therefore, the staff finds that the proposed change is consistent with intent of the BTP 8-8 position with respect to the supplemental power source capacity and inoperable EDG substitution.
3.1.2 Time to Make AAC Source Available The guidance in NUREG-0800, BTP 8-8, states, in part:
For plants using AAC or supplemental power sources discussed above, the time to make the AAC or supplemental power source available, including accomplishing the cross-connection, should be approximately one hour to enable restoration of battery chargers and control reactor coolant system inventory.
In the May 16, 2024, supplement, the licensee states, in part:
CNP can sustain the SBO event for the coping duration of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (CNP UFSAR 8.7.2), which exceeds the required one-hour time to make the supplemental AC power source available.
The SDGs are permanently installed plant equipment and automatically start upon a sustained loss of voltage on the 4.16 kV EP system. the SDGs automatically load to the EP Bus #1, which is then available for the operator to manually load to any of the safety related 4.16 kV Buses (CNP UFSAR 8.3.1).
CNP emergency operating procedure 1(2)-OHP-4023-ECA-0-0, Loss of All AC Power, Step 9, directs restoration of power to AC emergency buses via the SDGs in the event that reserve power is not available. This action has been demonstrated by plant operators, when performed in the CNP simulator, to be completed in less than 30 minutes.
SDG load capabilities are demonstrated in CNP Calculation 12-E-S-EPEDG-001, Allowable Outage Time (AOT) Diesel Generator ETAP Analysis. Relay settings are validated in CNP design document 12-E-S-EPEDG-002, Supplemental Diesel Generator System Relay Settings.
The NRC staff notes that the SDGs are not credited for the initial four hours of an SBO but are only necessary after that time if offsite and onsite AC power sources are not restored.
Therefore, CNP has the ability to cope with loss of all AC power (SBO) for more than one hour (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) independent of a supplemental AC power source (SDGs); thus, consistent with the BTB 8-8 guidance. Furthermore, the SDG operation and loading is procedurally controlled and when performed on the simulator took less than 30 minutes. Verification of SDG capacity and relays settings are per the referenced CNP calculations. Therefore, the staff finds that the proposed change is consistent with intent of the BTP 8-8 position with respect to time required to make the AAC source available.
3.1.3 AAC Source Operability Verification The guidance in NUREG-0800, BTP 8-8, states, in part:
The availability of AAC or supplemental power source should be verified within the last 30 days before entering extended AOT by operating or bringing the power source to its rated voltage and frequency for 5 minutes and ensuring all its auxiliary support systems are available or operational.
In the May 16, 2024, supplement, the licensee states, in part:
CNP's Technical Requirements Manual (TRM) Surveillance Requirement (SR) 8.8.3.2 requires: Exercise the SDGs by running each one unloaded for > = 5 minutes, with a Frequency of 14 days.
At a Frequency of 14 days, both of CNPs Supplemental Diesel Generators will be tested in accordance with the normal TRM surveillance frequency prior to entering the extended Completion Time period per CNP procedure 12-OHP-4030-033-001, Supplemental Diesel Generator Testing, Attachment 3, which requires checking that each SDG runs for at least 5 minutes unloaded and stabilizes at the rated voltage and frequency.
The NRC staff notes that the SDGs are tested every 14 days in accordance with SR 8.8.3.2 and CNP procedure 12-OHP-4030-033-001 which means they will be tested during the 30 days prior to entry in extended CT as described by BTP 8-8. CNP procedure 12-OHP-4030-033-001 requires verifying each SDG runs for at least 5 minutes unloaded at the rated voltage and frequency. Since the SDGs are operated unloaded for 5 minutes, their support systems are verified. Therefore, the staff finds that the proposed change is consistent with the intent of the BTP 8-8 position with respect to verification of AAC source operability.
3.1.4 Verification of Availability of AAC Source The guidance in NUREG-0800, BTP 8-8, states, in part:
The availability of AAC or supplemental power source shall be checked every 8-12 hours (once per shift). If the AAC or supplemental power source becomes unavailable any time during extended AOT, the unit shall enter the LCO and start shutting down within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This 24-hour period will be allowed only once within any given extended EDG AOT.
The proposed note to TS 3.8.1, states, in part:
During the 288-hour extended Completion Time, the SDGs shall be verified as available once per shift. If the SDGs becomes unavailable after the initial 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> while in the extended 288-hour Completion Time period, it shall be made available within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or the unit shall 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 be in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This 24-hour period will be allowed only once within the single extended Completion Time.
The NRC finds that the proposed change is consistent with intent of the BTP 8-8 position with respect to the verification of the AAC source availability, and actions needed to be taken when the AAC source becomes unavailable.
3.1.5 Regulatory Commitments The guidance in NUREG-0800, BTP 8-8, lists six Regulatory Commitments that the staff has found acceptable to maintain the availability of the equipment. Specifically, BTP 8-8, states:
The extended AOT will be used no more than once in a 24-month period (or refueling interval) on a per diesel basis to perform EDG maintenance activities, or any major maintenance on offsite power transformer and bus.
The preplanned maintenance will not be scheduled if severe weather conditions are anticipated.
The system load dispatcher will be contacted once per day to ensure no significant grid perturbations (high grid loading unable to withstand a single contingency of line or generation outage) are expected during the extended AOT.
Component testing or maintenance of safety systems and important non safety equipment in the offsite power systems that can increase the likelihood of a plant transient (unit trip) or LOOP will be avoided. In addition, no discretionary switchyard maintenance will be performed.
TS required systems, subsystems, trains, components, and devices that depend on the remaining power sources will be verified to be operable and positive measures will be provided to preclude subsequent testing or maintenance activities on these systems, subsystems, trains, components, and devices.
Steam-driven emergency feed water pump(s) in case of PWR [Pressurized Water Reactors] units will be controlled as protected equipment.
In Enclosure 5 of the May 16, 2024, supplement, the licensee committed to take these Regulatory Commitment actions. Additionally, the licensee codified these actions in the proposed note added to 3.8.1. Therefore, the NRC staff finds that the proposed change is consistent with the BTP 8-8 position with respect to the extended CT limitation, severe weather conditions restriction, grid perturbation monitoring, testing and maintenance restriction, and control of protected equipment.
3.1.6 Justification of 12-Day Completion Time The guidance in NUREG-0800, BTP 8-8 states, in part:
The EDG or offsite power AOT should be limited to 14 days to perform maintenance activities.... The licensee must provide justification for the duration of the requested AOT (actual hours plus margin based on plant-specific past operating experience).
The modification referred to in the LAR is replacing original plant equipment. Thus, the licensee has no existing plant operating experience that would provide a precedent for this specific modification. Although this redesign and overhaul is a first-time evolution, the licensee stated that there are portions of the work scope that have been performed previously and operating experience from those occurrences were taken into consideration. In the July 15, 2024, supplement, the licensee estimated that installation of the new equipment would take 8 days.
The licensee further stated that the worst-case scenario would involve needing to reinstall plant equipment back to the original configuration if the new equipment does not perform as intended. The licensee estimated this would result in an additional 4 days to be required to complete the project. The duration of 8 days plus the potential worst-case margin of 4 days resulted in the requested 12-day (288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br />) time period for the modification.
As described in the LAR, the licensee will implement compensatory measures to minimize risk to the station. The compensatory measures include protecting equipment, verifying the AAC power source availability, briefing the operators, and performing fire risk management actions.
The NRC staff finds the licensee's estimate for the time required to complete the modifications is reasonable, and the proposed CT is within the BTP-8-8 14-day limit. The modification activities described in the licensees proposed schedule provide a reasonable set of actions and timeline for completion of the replacement of the 12AB (Train B) LFE and associated bus while also reasonably accounting for unanticipated challenges during the replacement. Therefore, the staff finds that the proposed change is consistent with the BTP 8-8 position with respect to the 14-day limit.
3.2 Risk Insights Evaluation In Section 3.3, Evaluation of Risk Impact, of the LAR, the licensee states that the CNP probabilistic risk assessment (PRA) models, including Full Power Internal Events (FPIE),
Internal Flooding (IF), Fire PRA (FPRA) and Seismic PRA (SPRA), have been developed, peer-reviewed and undergone Fact & Observation (F&O) Closure Reviews. The NRC staff evaluated the technical quality of the CNP PRA models in accordance with RG 1.200, and found that, consistent with the guidance in RG 1.200, the FPIE, IF and FPRA models were developed per American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS)
RA-Sa-2009, Standard for Level I/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addendum A to RA-S-2008. The NRC staff also determined that the FPIE, IF, and FPRA underwent peer and F&O closure views in accordance with Nuclear Energy Institute (NEI) 05-04, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard, November 2008, Revision 2 (ML083430462), and NEI 07-12, Fire Probalistic Risk Assessment (FPRA) Peer Review Process Guidelines, June 2010, Revision 1 (ML102230070), as appropriate. By letter dated May 3, 2017 (ML17079A427), the NRC found NEI 05-04, and NEI 07-12, as acceptable guidance for licensees to use to close F&Os that were generated during a peer review process The NRC staff reviewed the technical acceptability of the licensees SPRA to determine if the SPRA is sufficient to provide confidence in the results such that it can be used to support the license amendment request. The licensee described an independent peer review of the SPRA, which was performed against the ASME/ANS RA-S Case 1 Case for ASME/ANS Ra-Sb-2013 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications (code case standard). This code case standard was endorsed, with staff exceptions and clarifications, by the NRC in Revision 3 of Regulatory Guide 1.200. In the May 16, 2024, supplement, the licensee summarized the use of the code case standard and peer review process. The licensee performed a closeout independent assessment of resolutions for finding level F&Os from the full-scope peer review following a process outlined by NEI 12-13, External Hazards PRA Peer Review Process Guidelines, August 2012 (ML12240A027), which was also accepted as guidance by the NRC in a May 3, 2017, letter. The closeout independent assessment included a concurrent focused-scope peer review for upgrades to the SPRA which also followed the guidance provided by NEI 12-13.
Based on the above, the NRC staff finds the technical quality of the CNP PRA models acceptable for use to support the proposed risk-informed license amendment request because they are consistent with the guidance in RG 1.200. The staff also verified that the risk impact evaluation performed by the licensee was consistent with RG 1.200 and, therefore, determined that it appropriately included internal events, internal flood, fire, seismic, and addressed other hazards.
3.2.1 Evaluation of PRA Open F&Os RG 1.200 states, in part:
F&Os that are not closed using an NRC-endorsed process should be evaluated by the licensee or applicant for their impact on a risk-informed application and addressed with documented justification with necessary changes made to the PRA prior to the use of PRA in the risk-informed application.
The NRC staff identified several open or partially open F&Os associated with the CNP FPIE, FPRA and SPRA models. For the open F&Os in the FPIE model, the licensee stated that the affected event failures were modeled conservatively, which result in overestimation of risk. The licensee also stated that although there is a lack of detailed interview and plant walkdown documentation, some walkdowns and interviews were performed, and no model changes were required as the result. The remaining system walkdowns and interview is also expected by the licensee to yield the same result.
For the open F&Os in the FPRA, the licensee also stated that conservatism in the FPRA model and the disconnect in the documentation does not impact the numerical model results.
However, the NRC staff determined that model uncertainties due to open F&Os with respect to hot short probabilities that have not been incorporated into the FPRA model and the inconsistencies between the values presented in documentation and those used in the FPRA model needed to be addressed in further detail.
In the July 2, 2024, supplement, the licensee stated that the inconsistency is primarily related to cable type identification. Some components that had mixed or unknown cable compositions were assigned thermoset cable uncertainty parameters, when they should have used more conservative thermoplastic cable uncertainty parameters. The licensee also stated that further review of the issue has concluded that only the cable uncertainty parameters are impacted, and that the hot short probabilities used to calculate core damage frequency (CDF) and large early release frequency (LERF) have been applied correctly. Since the licensee stated that the LAR application uses the model's point estimate values, which do not consider parametric uncertainty, the NRC staff determined that the subject partially open F&O does not impact the CDF and LERF values, and the risk insights presented in the LAR application.
In Enclosure 2 of the LAR, the licensee summarized the SPRA finding level F&Os that remained open after the closeout process, along with the licensees dispositions and the impact the open F&Os may have on the SPRA results. The NRC staff reviewed these elements in the context of the regulatory decisionmaking associated with this one-time TS CT extension. Since the SPRA had F&Os that remained open after the closeout process, the NRC staff evaluated the open F&Os for impacts to this specific application. On May 30, 2024 (ML24156A007), the NRC staff issued two requests for additional information (RAIs) to support the review of the SPRA. The first RAI was associated with the potential impacts of cracking in the auxiliary building structure on the LAR CDF and LERF values and the risk insights presented in the LAR application. The second RAI was associated with the impact of structural variability on component fragilities in the SPRA.
In the July 2, 2024, supplement, the licensee explained that cracking in the auxiliary building structure would not impact the risk results presented in the LAR. The licensee also provided a detailed list of fragilities that contributed significantly to the model results and a disposition of the impact of the fragilities on the risk results presented in the LAR. The NRC staff determined that the licensees responses confirmed statements in the LAR that the open F&Os are not expected to impact the FPIE, the FPRA, and SPRA results associated with the LAR. Based on that, the staff determined that the subject open F&Os do not impact the CDF and LERF values and the risk insights presented in the LAR application. Therefore, based on the above, the NRC staff finds the technical quality of the PRAs used in this application is acceptable.
3.2.2 Evaluation of PRA Results Regulatory Guide 1.177 provides quantitative acceptance guidelines for risk impact related to risk-informed changes to CTs of plant TS. One-time CT changes are considered acceptable if incremental conditional core damage probability (ICCDP) is less than 1.0E-6 and incremental conditional large early release probability (ICLERP) is less than 1.0E-7. The NRC staff reviewed the licensees evaluation methodology and the reported total increase in ICCDP of 3.25E-07 for Unit 1 and 3.09E-07 for Unit 2, and ICLERP of 3.09E-08 for Unit 1 and 1.28E-08 for Unit 2 due to the proposed change and determine it is acceptable because the increase is within the acceptance guidelines.
3.2.3 Evaluation of Defense-in-Depth and Safety Margin The licensee stated that during the proposed one-time extension, defense-in-depth measures will be applied to account for unknown and unforeseen failure mechanisms or other phenomena to assure the safety function is maintained. These measures include compensatory actions such as implementing risk management actions (RMAs) to prevent high-risk configurations; providing guarded and readily available alternate power sources; implementing fire watches and prohibiting hot works in high fire risk areas; removing flooding sources; and refraining from elective maintenance on components credited for accident mitigation. To maintain safety margin, the licensee stated that the proposed one-time amendment will not alter the design or capabilities of the emergency safeguards systems, will not result in plant operation in a configuration outside the design basis, and sufficient equipment redundancy will be provided due to the availability of emergency diesel generators and auxiliaries to ensure continuous service.
Based on the above, the NRC staff concluded that the licensee has provided adequate measures to ensure defense-in-depth and safety margin is maintained during the CT extension of TS 3.8.1.
3.3 Technical Evaluation Conclusion
Based on the results of the deterministic evaluation described above in section 3.1 and the PRA evaluation described above in section 3.2, the NRC staff concludes that the proposed one-time change to TS 3.8.1, AC Sources -Operating, to extend the allowable CT for Required Action A.3 from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br /> (12 days) to complete the proposed plant modification is acceptable and will not impact the licensees continuous compliance with the requirements of 10 CFR 50.36, 50.63, and 10 CFR 50 Appendix A, GDC 17. The existing CT of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> will be restored after the modification is complete.
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 RELATED TO AMENDMENT NOS. 363 AND 344 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 (i.e., initial and supplements)
Safety Evaluation Date April 3, 2024, ADAMS Accession No. ML24094A288 Supplements May 16, 2024 (ML24137A221)
July 2, 2024 (ML24184A140)
July 15, 2024 (ML24197A126)
August 28, 2024 (ML24241A161)
September 3, 2024
1.0 INTRODUCTION
Indiana Michigan Power Company (I&M, the licensee) requested changes to the technical specifications (TSs) for Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2 (CNP), by license amendment request (application). The amendments would revise TS 3.8.1, AC [Alternating Current] Sources - Operating, by adding a footnote for TS 3.8.1, Required Action A.3 to allow a one-time completion time extension from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br /> to support the replacement of the 12AB (Train B) Loop Feed Enclosure and associated bus for the Train B reserve feed preferred power source.
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 August 6, 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 June 11, 2024 (89 FR 49241), 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.
ML24225A002 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DEX/EEEB/BC NAME SWall SRohrer WMorton DATE 08/09/2024 08/12/2024 08/07/2024 OFFICE NRR/DRA/APLB/BC NRR/DRA/APLC/BC NRR/DSS/STSB/BC NAME EDavidson SVasavada SMehta (Kwest for)
DATE 08/02/2024 08/02/2024 08/12/2024 OFFICE OGC - NLO NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME NMertz JWhited (BBallad for)
SWall DATE 08/22/2024 08/30/2024 09/03/2024