Text

Issuance of Amendment No. 356 Regarding Updating the Reactor Coolant System Pressure-Temperature Limits
	ML20329A001
Person / Time
Site:	Cook
Issue date:	01/12/2021
From:	Scott Wall; Plant Licensing Branch III
To:	Gebbie J; Indiana Michigan Power Co
	Wall S
References
	EPID L-2020-LLA-0081
	Download: ML20329A001 (28)
	v • d • e

January 12, 2021 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 NO. 1 - ISSUANCE OF AMENDMENT NO. 356 RE: UPDATING THE REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE LIMITS (EPID L-2020-LLA-0081)

Dear Mr. Gebbie:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 356 to Renewed Facility Operating License No. DPR-58 for Donald C. Cook Nuclear Plant, Unit No. 1 (CNP-1). The amendment consists of changes to the license and technical specifications (TSs) in response to your application dated April 7, 2020.

The amendment revises the CNP-1 TSs to replace the current pressure-temperature limits for the reactor coolant system (RCS) in TS 3.4.3, RCS Pressure and Temperature (P/T) Limits, which are applicable for a service period up to 32 effective full-power years (EFPY), with limits that extend up to 48 EFPY. In addition, the amendment revises TS 3.4.12, Low Temperature Overpressure Protection (LTOP) System, to align with an updated LTOP analysis. The proposed changes to the LTOP requirements in TS 3.4.12 result in changes to TS 3.4.6, RCS Loops - MODE 4; TS 3.4.7, RCS Loops - MODE 5, Loops Filled; and TS 3.4.10, Pressurizer Safety Valves.

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 No. 50-315

Enclosures:

1. Amendment No. 356 to DPR-58

2. 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. 356 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 7, 2020, 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.

Enclosure 1

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. 356, 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 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Nancy L. Nancy L. Salgado Date: 2021.01.12 Salgado 13:42:38 -05'00' 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: January 12, 2021

ATTACHMENT TO AMENDMENT NO. 356 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58 DONALD C. COOK NUCLEAR PLANT, UNIT NO. 1 DOCKET NO. 50-315 Renewed Facility Operating License No. DPR-58 Replace the following page of Renewed Facility Operating License No. DPR-58 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change REMOVE INSERT Technical Specifications Replace the following pages of 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 areas of change.

REMOVE INSERT 3.4.3-3 3.4.3-3 3.4.3-4 3.4.3-4 3.4.6-1 3.4.6-1 3.4.7-1 3.4.7-1 3.4.10-1 3.4.10-1 3.4.12-1 3.4.12-1 3.4.12-2 3.4.12-2 3.4.12-3 3.4.12-3 3.4.12-4 3.4.12-4 3.4.12-5 3.4.12-5

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. 356, 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, Renewed License No. DPR-58 Amendment No: 355, 356

RCS P/T Limits 3.4.3 2500 Leak Test Limit 2250 Unacceptable Operation Acceptable 2000 Operation 1750 Criticality Limit 1500 Heatup Limit 60°F/Hr Reactor Coolant System Pressure (psig) 1250 1000 750 500 Boltup Temperature 250 0

RCS Vacuum

-14.7 psig

-250 0 50 100 150 200 250 300 350 400 450 500 550 Average Reactor Coolant System Temperature (ºF)

Figure 3.4.3-1 (page 1 of 1)

Reactor Coolant System Pressure versus Temperature Limits -

Heatup Limit, Criticality Limit, and Leak Test Limit (Applicable for service period up to 48 EFPY and during vacuum fill)

Cook Nuclear Plant Unit 1 3.4.3-3 Amendment No. 287, 323, 356

RCS P/T Limits 3.4.3 2500 2250 Unacceptable Acceptable 2000 Operation Operation 1750 Reactor Coolant System Pressure (psig) 1500 1250 1000 Cooldown Rate 750

(°F/Hr) 0 20 500 40 60 100 250 Boltup Temperature 0

RCS Vacuum -14.7 psig

-250 0 50 100 150 200 250 300 350 400 450 500 550 Average Reactor Coolant System Temperature (ºF)

Figure 3.4.3-2 (page 1 of 1)

Reactor Coolant System Pressure versus Temperature Limits -

Various Cooldown Rates Limits (Applicable for service period up to 48 EFPY and during vacuum fill)

Cook Nuclear Plant Unit 1 3.4.3-4 Amendment No. 287, 323, 356

RCS Loops - MODE 4 3.4.6 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Loops - MODE 4 LCO 3.4.6 Two loops consisting of any combination of RCS loops and residual heat removal (RHR) loops shall be OPERABLE, and one loop shall be in operation.

NOTES-------------------------------------------

1. All reactor coolant pumps (RCPs) and RHR pumps may be removed from operation for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

a. No operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1, "SHUTDOWN MARGIN (SDM)"; and

b. Core outlet temperature is maintained at least 10°F below saturation temperature.

2. Reactor coolant pumps shall not be started with one or more RCS cold leg temperatures 297°F unless the secondary water temperature of each steam generator is < 50°F above each of the RCS cold leg temperatures.

APPLICABILITY: MODE 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required loop A.1 Initiate action to restore a Immediately inoperable. second loop to OPERABLE status.

AND A.2 --------------NOTE--------------

Only required if RHR loop is OPERABLE.

Be in MODE 5. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Cook Nuclear Plant Unit 1 3.4.6-1 Amendment No. 287, 356

RCS Loops - MODE 5, Loops Filled 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Loops - MODE 5, Loops Filled LCO 3.4.7 One residual heat removal (RHR) loop shall be OPERABLE and in operation, and either:

a. One additional RHR loop shall be OPERABLE; or

b. The secondary side water level of at least two steam generators (SGs) shall be above the lower tap of the SG wide range level instrumentation by 420 inches.

NOTES-------------------------------------------

1. The RHR pump of the loop in operation may be removed from operation for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided:

a. No operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1, "SHUTDOWN MARGIN (SDM)"; and

b. Core outlet temperature is maintained at least 10°F below saturation temperature.

2. One required RHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided that the other RHR loop is OPERABLE and in operation.

3. Reactor coolant pumps shall not be started with one or more RCS cold leg temperatures 297°F unless the secondary water temperature of each steam generator is < 50°F above each of the RCS cold leg temperatures.

4. All RHR loops may be removed from operation during planned heatup to MODE 4 when at least one RCS loop is in operation.

APPLICABILITY: MODE 5 with RCS Loops Filled.

Cook Nuclear Plant Unit 1 3.4.7-1 Amendment No. 287, 356

Pressurizer Safety Valves 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Pressurizer Safety Valves LCO 3.4.10 Three pressurizer safety valves shall be OPERABLE with lift settings 2411 psig and 2559 psig.

APPLICABILITY: MODES 1, 2, and 3, MODE 4 with all RCS cold leg temperatures > 297°F.

NOTE--------------------------------------------

The lift settings are not required to be within the LCO limits during MODES 3 and 4 for the purpose of setting the pressurizer safety valves under ambient (hot) conditions. This exception is allowed for 54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months following entry into MODE 3 provided a preliminary cold setting was made prior to heatup.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pressurizer safety A.1 Restore valve to 15 minutes valve inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND OR B.2 Be in MODE 4 with any 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months RCS cold leg temperatures Two or more pressurizer 297°F.

safety valves inoperable.

Cook Nuclear Plant Unit 1 3.4.10-1 Amendment No. 287, 356

LTOP System 3.4.12 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.12 Low Temperature Overpressure Protection (LTOP) System LCO 3.4.12 An LTOP System shall be OPERABLE with the following:

A. No safety injection (SI) pump capable of injecting into the RCS and:

1. The accumulators isolated, except an accumulator may be unisolated when the accumulator pressure is less than the maximum RCS pressure for the existing RCS cold leg temperature allowed by the P/T limit curves provided in TS 3.4.3;

2. One of the following pressure relief capabilities:

a. The residual heat removal (RHR) suction relief valve with a setpoint 450 psig and RCS cold leg temperature 150°F;

b. The residual heat removal (RHR) suction relief valve with a setpoint 450 psig and at least one RCP running;

c. Two PORVs with lift settings 435 psig and the residual heat removal (RHR) suction relief valve with a setpoint 450 psig;

d. Two PORVs with lift settings 435 psig and RCS cold leg temperature 210°F; or

e. The RCS depressurized and an RCS vent of 2.0 square inches or any single PORV blocked open.

NOTE--------------------------------------------

Reactor coolant pumps shall not be started with one or more RCS cold leg temperatures 297°F unless the secondary water temperature of each steam generator is < 50°F above each of the RCS cold leg temperatures.

Cook Nuclear Plant Unit 1 3.4.12-1 Amendment No. 287, 356

LTOP System 3.4.12 APPLICABILITY: MODE 4 when any RCS cold leg temperature is 297°F, MODE 5, MODE 6 when the reactor vessel head is on.

ACTIONS

NOTE----------------------------------------------------------

LCO 3.0.4.b is not applicable when entering MODE 4.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SI pumps A.1 Initiate action to verify all SI Immediately capable of injecting into pumps are not capable of the RCS. injecting into the RCS.

B. An accumulator not B.1 Isolate affected 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months isolated when the accumulator.

accumulator pressure is greater than or equal to the maximum RCS pressure for the existing cold leg temperature allowed by TS 3.4.3.

C. Required Action and C.1 Increase RCS cold leg 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion temperature to > 297°F.

Time of Condition B not met. OR C.2 Depressurize affected 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months accumulator to less than the maximum RCS pressure for existing cold leg temperature allowed in TS 3.4.3.

D. One required RCS relief D.1 Restore required RCS relief 7 days valve inoperable in valve to OPERABLE status.

MODE 4 while complying with LCO A.2.c or A.2.d.

(continued)

Cook Nuclear Plant Unit 1 3.4.12-2 Amendment No. 287, 356

LTOP System 3.4.12 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. One required RCS relief E.1 Restore required RCS relief 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months valve inoperable in valve to OPERABLE status.

MODE 5 or 6 while complying with LCO A.2.c or A.2.d.

F. Required RCP not F.1 Do not start a RCP. Immediately running.

AND F.2 Enter Condition G.

G. Two or more required G.1 Depressurize RCS and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months RCS relief valves establish RCS vent of 2.0 inoperable. square inches or block open a single PORV.

OR Required Action and associated Completion Time of Condition A, C, D, E, or F not met.

OR LTOP System inoperable for any reason other than Condition A, B, C, D, E, or F.

Cook Nuclear Plant Unit 1 3.4.12-3 Amendment No. 287, 334, 356

LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 Verify no SI pumps are capable of injecting into the In accordance RCS. with the Surveillance Frequency Control Program SR 3.4.12.2 Verify the required RCP is running. In accordance with the Surveillance Frequency Control Program SR 3.4.12.3 -------------------------------NOTE------------------------------

Valve position may be verified by use of administrative means.

Verify each accumulator that is required to be In accordance isolated is isolated. with the Surveillance Frequency Control Program SR 3.4.12.4 Verify RHR suction isolation valves are open for the In accordance required RHR suction relief valve. with the Surveillance Frequency Control Program SR 3.4.12.5 Verify required RCS vent 2.0 square inches open In accordance or a single PORV blocked open. with the Surveillance Frequency Control Program SR 3.4.12.6 Verify PORV block valve is open for each required In accordance PORV. with the Surveillance Frequency Control Program (continued)

Cook Nuclear Plant Unit 1 3.4.12-4 Amendment No. 287, 334, 356

LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS (continued)

SR 3.4.12.7 Verify pressure in each required emergency air tank In accordance bank is 900 psig. with the Surveillance Frequency Control Program SR 3.4.12.8 -------------------------------NOTE------------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after decreasing RCS cold leg temperature to 297°F.

Perform a COT on each required PORV, excluding actuation. In accordance with the Surveillance Frequency Control Program SR 3.4.12.9 Perform CHANNEL CALIBRATION for each required In accordance with PORV actuation channel. the Surveillance Frequency Control Program Cook Nuclear Plant Unit 1 3.4.12-5 Amendment No. 334, 356

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 356 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-58 INDIANA MICHIGAN POWER COMPANY DONALD C. COOK NUCLEAR PLANT, UNIT NO. 1 DOCKET NO. 50-315

1.0 INTRODUCTION

By application dated April 7, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20108F011), Indiana Michigan Power Company (I&M, the licensee) requested changes to the technical specifications (TSs) for the Donald C. Cook Nuclear Plant, Unit No. 1 (CNP-1).

The amendment would revise the CNP-1 TSs to replace the current pressure-temperature limits for the reactor coolant system (RCS) in TS 3.4.3, RCS Pressure and Temperature (P/T)

Limits, which are applicable for a service period up to 32 effective full-power years (EFPY), with limits that extend up to 48 EFPY. In addition, the amendment would revise TS 3.4.12, Low Temperature Overpressure Protection (LTOP) System, to align with an updated LTOP analysis.

The proposed changes to the LTOP requirements in TS 3.4.12 would result in changes to TS 3.4.6, RCS Loops - MODE 4; TS 3.4.7, RCS Loops - MODE 5, Loops Filled; and TS 3.4.10, Pressurizer Safety Valves.

2.0 REGULATORY EVALUATION

2.1 Applicable Regulatory Requirements Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(c)(2), Limiting conditions for operation, states that [l]imiting conditions for operation 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.

Enclosure 2

The regulation in 10 CFR 50.36(c)(2)(ii)(B) states that a TS limiting condition for operation (LCO) must be established for:

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.

The regulation in 10 CFR 50.36(c)(2)(ii)(C) states that a TS LCO must also be established for:

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.

10 CFR Part 50, Appendix G, Fracture Toughness Requirements (Appendix G). The criteria in Section IV.A of Appendix G require owners of U.S. light-water reactors to calculate P/T limits during normal operations and during anticipated operational transients that may impact operations of the reactor, including heatups and cooldowns of the reactor and pressure testing.

For materials located in the beltline of the reactor pressure vessel (RPV),Section IV.A of Appendix G requires the values of RTNDT (the nil ductility reference temperature) used in the development of the P/T limits to account for the effects of neutron irradiation, including incorporation of the results of the RPV materials surveillance program that is required and implemented in accordance with the requirements in 10 CFR Part 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements.Section IV.A.2 of Appendix G requires the P/T limits to be at least as conservative as limits obtained by following the methods of analysis and the margins of safety specified in Appendix G of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code),Section XI, Division 1.Section IV.A.2 of Appendix G also establishes specific minimum temperature requirements that must be incorporated into the calculations of P/T limit curves. The methods in ASME Code,Section XI, Appendix G define ASMEs methods for performing calculation of P/T limit curves.

10 CFR Part 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements (Appendix H). The regulation in Appendix H requires a licensed owner of a U.S. light-water reactor to have a surveillance monitoring program for ferritic materials located in the beltline of the RPV if cumulative neutron fluence exposures are projected to be in excess 1.0 X 1017 n/cm2 (E > 1.0 MeV). For plant-specific programs, like that currently in place for CNP-1, Appendix H requires the program to be designed in accordance with the edition of the American Society for Testing and Materials (ASTM) Standard E 185, Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels, that was current on the issue date of the ASME Code to which the RPV was purchased.

Appendix H requires each capsule withdrawal and the test results for test specimens in the capsules to be the subject of a summary technical report that is required to be submitted within 12 months of the capsule withdrawal, unless an extension is granted by the Director of the Office of Nuclear Reactor Regulation.

2.2 Applicable Regulatory Guidance Regulatory Guide (RG) 1.99, Revision 2. RG 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, dated May 1988 (ADAMS Accession No. ML003740284),

establishes an acceptable methodology that may be used by licensees for calculating the adjusted reference temperature values of ferritic base metals or welds. For RTNDT calculations, the methodology provides two positions and methods for calculating chemistry factors that are part of the RTNDT calculations: (a) use of the copper and nickel alloying chemistries and the chemistry factor tables provided in the RG or (b) for RPV base metal and weld materials that are included in the RPV surveillance program, use of the applicable surveillance data for the materials. For the RTNDT calculations that rely on applicable RPV surveillance data, the RG methods include methods for checking the credibility of the surveillance data and establishing the impact that the degree of credibility will have on the margin term value used in the RTNDT calculations. The RG also includes methods for calculating the fluence factors that are used in the RTNDT calculations and for attenuating (decreasing) the neutron fluence values as a function of RPV wall thickness.

RG 1.190. RG 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, dated March 2001 (ADAMS Accession No. ML010890301), describes methods and assumptions acceptable to the NRC staff for determining the RPV neutron fluence with respect to the General Design Criteria (GDC) contained in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants. In consideration of the guidance set forth in RG 1.190, GDCs 14, 30, and 31 are applicable.

GDC 14, Reactor coolant pressure boundary, requires the design, fabrication, erection, and testing of the reactor coolant pressure boundary so as to have an extremely low probability of abnormal leakage, or rapidly propagating failure, and of gross rupture.

GDC 30, Quality of reactor coolant pressure boundary, requires, in part, that components comprising the reactor coolant pressure boundary be designed, fabricated, erected, and tested to the highest quality standards practical.

GDC 31, Fracture prevention of reactor coolant pressure boundary, pertains to the design of the reactor coolant pressure boundary and states:

The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws.

The guidance in RG 1.190 indicates that the following elements comprise an acceptable fluence calculation:

1. Determination of the geometrical and material input data,

2. Determination of the core neutron source,

3. Propagation of the neutron fluence from the core to the vessel and into the cavity, and

4. Qualification of the calculational procedure.

The NRCs review of the fluence calculation was performed to establish that elements 1 through 4, above, of the calculational method adhere to the regulatory positions set forth in RG 1.190.

WCAP-14040-A, Revision 4. The methods in Westinghouse Electric Company Non-Proprietary Class 3 Report No. WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, dated May 2004 (ADAMS Accession No. ML050120209), are an NRC staff-approved methodology for determining pressurized-water reactor plant-specific P/T limit values for RPV beltline materials and LTOP system enable temperature and pressure lift setpoint values.

Regulatory Issue Summary (RIS) 2014-11. RIS 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, dated October 14, 2014 (ADAMS Accession No. ML14149A165), describes P/T limit requirements in 10 CFR Part 50, Appendix G that apply to all materials in the reactor coolant pressure boundary that are made from ferritic steel, including assessment of RPV discontinuities.

NUREG-0800. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, Section 5.2.2, Revision 3, Overpressure Protection, dated March 2007 (ADAMS Accession No. ML070540076), and Branch Technical Position (BTP) 5-2, Revision 3, Overpressurization Protection of Pressurized-Water Reactors while Operating at Low Temperatures, dated March 2007 (ADAMS Accession No. ML070850008), provide guidance to the NRC staff in reviewing overpressurization protection of pressurized-water reactors while operating at low temperatures. Paragraph B.1 of BTP 5-2 specifies that the LTOP system should be capable of relieving pressure during all anticipated overpressurization events at a rate sufficient to satisfy the TS limits while operating at low temperatures.

The April 7, 2020, application included Westinghouse Electric Company Non-Proprietary Class 3 Report No. WCAP-18455-NP, Revision 1, D.C. Cook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, dated February 2020 (ADAMS Accession No. ML20108F000), and proprietary and non-proprietary versions of the LTOP analysis for 48 EFPY.

By letter dated October 1, 2014 (ADAMS Accession No. ML14259A549), the NRC approved the CNP-1 P/T limit curves approved for 32 EFPY and established License Condition 2.C.(19),

Operation with Vacuum Fill, that required the licensee to submit an analysis of the P/T curves that demonstrates consideration of all ferritic reactor vessel materials as defined in Appendix G, including non-beltline ferric reactor vessel materials. By letter dated September 25, 2015 (ADAMS Accession No. ML15272A491), the licensee fulfilled this condition. The information in the September 25, 2015, letter was used in the staffs evaluation of the April 7, 2020, application.

2.3 Proposed TS Changes

The licensee proposes the following revisions of the TS for CNP-1:

Replace the existing TS Figure 3.4.3-1, Reactor Coolant System Pressure versus Temperature Limits - Heatup Limit, Criticality Limit, and Leak Test Limit (Applicable for service period up to 32 EFPY and during vacuum fill), and TS Figure 3.4.3-2, Reactor Coolant System Pressure versus Temperature Limits - Various Cooldown Rates Limits (Applicable for service period up to 32 EFPY and during vacuum fill), with proposed TS Figure 3.4.3-1 and TS Figure 3.4.3-2, as shown in Enclosure 3 to the April 7, 2020, application.

Change TS 3.4.12 to ensure that the new LTOP analysis requirements are reflected in the LCO. The previous LTOP analysis and TS reflect the requirement to limit RCS mass injection capability to either one or two centrifugal charging pumps (CCPs), dependent on RCS temperature and available relief capacity. The new LTOP analysis demonstrates that RCS overpressure protection is provided when the limiting mass injection transient is from two operating charging pumps for the full range of LTOP applicability. Therefore, the licensee proposes that the restriction on CCPs that may be in operation be deleted.

Structure the proposed TS 3.4.12 as a series of five LCO conditions based on relief capabilities, RCS temperature limitations, and reactor coolant pump (RCP) status as applicable, that must be met to ensure RCS overpressure protection. Only one of the five proposed LCO conditions must be met to meet the requirements of the LCO.

Change the TS 3.4.12 mode of applicability to MODE 4 when any RCS cold leg temperature is less than or equal to 297 degrees Fahrenheit (°F).

Change the TS 3.4.12 note for RCP start to add the new LTOP enable temperature (297 °F) and to delete the allowance to start RCPs if pressurizer level is less than 62 percent.

Delete TS 3.4.12 Condition B. This condition provided actions if two CCPs were capable of injecting into the RCS when only one was allowed.

Relabel TS 3.4.12 Condition C as Condition B and reword it as follows:

An accumulator not isolated when the accumulator pressure is greater than or equal to the maximum RCS pressure for the existing cold leg temperature allowed by TS 3.4.3.

Relabel TS 3.4.12 Condition D as Condition C and reword Action C.1 to reflect the new LTOP enable temperature (297 °F) and Action C.2 to reflect the new wording of relabeled Condition B.

Relabel TS 3.4.12 Condition E as Condition D and reword it as follows:

One required RCS relief valve inoperable in MODE 4 while complying with LCO A.2.c or A.2.d.

Relabel TS 3.4.12 Condition F as Condition E and reword it as follows:

One required RCS relief valve inoperable in MODE 5 or 6 while complying with LCO A.2.c or A.2.d.

Add a new TS 3.4.12 Condition F to provide actions if the required RCP is not running.

The prescribed actions are to not start a RCP and to enter Condition G immediately.

Modify the second OR statement in TS 3.4.12 Condition G to reflect the relabeled Condition B.

Replace Surveillance Requirement (SR) 3.4.12.2 with a requirement to verify that the required RCP is running.

Add clarification wording to SR 3.4.12.3.

Change the LTOP enable temperature to 297 °F in SR 3.4.12.8.

Update TS 3.4.6, TS 3.4.7, and TS 3.4.10 to reflect the proposed changes to the LTOP requirements in TS 3.4.12.

3.0 TECHNICAL EVALUATION

3.1 Evaluation of the Proposed Changes to the P/T Limit Curves The NRC staff reviewed the proposed P/T limit curves for 48 EFPY in the submittal in order to assess whether:

the proposed P/T limit curves are at least as conservative and incorporate the minimum safety margin requirements specified in Appendix G of the ASME Code.

the adjusted reference temperature (RTNDT) values used in the development of the proposed P/T limit curves have accounted for: (a) the increases in the neutron fluence exposures for RPV beltline welds and base metals that require RTNDT calculations, (b) the results of the licensees Appendix H RPV material surveillance program, and (c) structural discontinuities as described in NRC RIS 2014-11.

the proposed P/T limit curves have appropriately accounted for the minimum temperature requirements that are mandated for P/T limit curve assessments in Section IV.A.2 of Appendix G and are specified in Table 1 of Appendix G.

3.1.1 Neutron Fluence Assessment Accurate and reliable neutron fluence values are required in order to satisfy the provisions of RG 1.190. The neutron fluence analysis provided in WCAP-18455-NP, Revision 1, utilizes RAPTOR-M3G and FERRET codes, which is consistent with the NRC-approved methodology described in WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, dated July 2018 (ADAMS Accession No. ML18204A010). This methodology is used for both the beltline and extended beltline regions. The neutron fluence values that are used in the calculations of the adjusted reference temperature values for the P/T limit calculations are

projected to 48 EFPY of operation. The NRC staff review has determined that the calculated neutron fluence values are appropriately consistent with the guidelines in RG 1.190 and, therefore, are acceptable.

3.1.2 P/T Limit Curves Review The NRC staff noted that the proposed P/T limit curves are based on the methods of analysis in WCAP-18455-NP, Revision 1. The staff performed an independent assessment to confirm that the P/T limit curves for 48 EFPY would meet the requirements for P/T limit calculations specified in Appendix G, including the use of applicable surveillance data from the licensees RPV material surveillance program for CNP-1 or from applicable sister plant RPV material surveillance programs.

P/T Limit Curves Review - RIS 2014-11 Assessment In Section 7 of WCAP-18455-NP, Revision 1, the licensee made the following statement in relation to the P/T limits assessment of RPV discontinuity locations:

Per Table 2-7, both the inlet and outlet nozzle forgings and welds for [CNP-1]

have projected fluence values at the lowest extent of the nozzle welds that do not exceed the 1 x 1017 n/cm2 fluence threshold at 48 EFPY. Consistent with NRC RIS 2014-11 . . ., neutron radiation embrittlement need not be considered herein for either the inlets or outlet nozzle materials.

The NRC staff noted that the licensees letter dated September 25, 2015, indicates that the limiting RPV geometric discontinuity locations for stress and neutron fluence are the discontinuities created by the presence of the RPV inlet and outlet nozzles, which are ferritic nozzles that are welded to the upper shell of the RPV. Specifically, in the September 25, 2015, letter, the licensee concluded that the P/T limit curves for the beltline of the vessel are bounding for the P/T limits through 32 EFPY, including consideration of the applied stress loads and fluence exposures for applicable RPV nozzles.

The NRC staff confirmed that the submittal and the information in WCAP-18455-NP, Revision 1, provides sufficient demonstration that the projected 1/4 thickness (1/4T) and 3/4 thickness (3/4T)

RTNDT values for the limiting RPV beltline materials at 48 EFPY are bounding relative to those for the assessments of the RPV inlet and outlet nozzles at 48 EFPY. Since the stress level ratio for the RPV beltline components will stay the same relative to the stress loads that are imparted to the RPV inlet and outlet nozzles, the staff concludes that the assessment of RPV inlet and outlet nozzles in WCAP-18455-NP, Revision 1, when assessed in conjunction with the information provided in the September 25, 2015, letter, provides sufficient demonstration that the P/T limit curves for the beltline region of the RPV will remain bounding for the P/T limits assessment at 48 EFPY.

P/T Limit Curves Review - RPV Surveillance Data Assessment and RTNDT Assessment The NRC staff verified that the licensees 1/4T and 3/4T RTNDT values used in the P/T limit calculations were established in accordance with the methods of analysis and criteria defined in WCAP-14040-A, Revision 4, which is consistent with the analytical methods in RG 1.99, Revision 2, for calculation of RTNDT values. This includes the WCAP methodologys use of Position C.2.1 in RG 1.99, Revision 2, for the calculation of chemistry factor values and RTNDT values of those RPV beltline materials that are represented with applicable CNP-1 RPV

surveillance data or sister plant surveillance data. The staff confirmed that the following specified RPV materials in the CNP-1 RPV are represented by RPV surveillance program materials:1 RPV intermediate shell plates B4406-2 and B4406-3 made from Heat No. C3506:

transverse and longitudinal plate specimens made from Heat C3506 in CNP-1 RPV surveillance capsules T, X, Y, and U, with the applicable surveillance results reported in the following documents:

o Southwest Research Institute Project No. 02-4770, Final Report, Reactor Vessel Material Surveillance Program for Donald C. Cook Unit No. 1, Analysis of Capsule T, December 8, 1977 (ADAMS Accession No. ML12236A067).

o Southwest Research Institute Project No. 02-6159, Final Report, Reactor Vessel Material Surveillance Program for Donald C. Cook Unit No. 1, Analysis of Capsule X, June 22, 1981 (ADAMS Accession No. ML20330A194).

o Southwest Research Institute Project No. 06-7244-001, Final Report, Reactor Vessel Material Surveillance Program for Donald C. Cook Unit No. 1, Analysis of Capsule Y, January 1984 (ADAMS Accession No. ML12236A079).

o Westinghouse Electric Company Non-Proprietary Class 3 Report No. WCAP-12483, Revision 1, Analysis of Capsule U from the American Electric Power Company D. C. Cook Unit 1 Reactor Vessel Radiation Surveillance Program, December 2002 (ADAMS Accession No. ML023460493).

The NRC staff confirmed that the data are credible using the credibility criteria defined in RG 1.99, Revision 2.

RPV intermediate shell-to-lower shell circumferential weld No. 9-442, Heat 1P3571:

sister plant weld specimens made from Heat 1P3571 as exposed in Capsules V, R, P, S, and T of the Kewaunee RPV surveillance program and the W-263º, W-253º, A-25º, and A-35º Capsules in the Maine Yankee RPV surveillance program, with the applicable surveillance results reported in the following documents:

o Westinghouse Electric Company Non-Proprietary Class 3 Report No.

WCAP-15074, Evaluation of the 1P3571 Weld Metal from the Surveillance Programs for Kewaunee and Maine Yankee, November 18, 1998 (ADAMS Accession No. ML111861650) o Westinghouse Electric Company Non-Proprietary Class 3 Report No.

WCAP-16641-NP, Revision 0, Analysis of Capsule T from Dominion Energy Kewaunee Power Station Reactor Vessel Radiation Surveillance Program, October 2006 (ADAMS Accession No. ML063250415).

1 The information in WCAP-18455-NP, Revision 1, reports that the RPV axial welds were fabricated from tandem weld Heat 13253/12008(T). The CNP-1 surveillance capsule weld specimens were made from single weld heat (i.e.,

Heat No. 13253). Since the heat of material for the surveillance weld specimens does not match the heat of material of intermediate shell, lower shell, and nozzle shell axial welds, the licensee performed the chemistry factor calculations of the axial welds using the nonsurveillance-based chemistry factor methods and chemistry factor tables defined in Position C.1.1 of RG 1.99, Revision 2.

The NRC staff confirmed that the data are credible using the credibility criteria defined in RG 1.99, Revision 2.

The NRC staff performed independent calculations of the RTNDT values using the regulatory positions and methods in RG 1.99, Revision 2, and verified that the 1/4T RTNDT values reported in Table 7-2 of WCAP-18455-NP, Revision 1, and the 3/4T RTNDT values reported in Table 7-3 of WCAP-18455-NP, Revision 1, are valid for the P/T limits calculations. Based on its review, the staff verified that the RTNDT assessments for 48 EFPY are limited by the RTNDT calculations of the following materials:

A 1/4T RTNDT value of 229 ºF and a 3/4T RTNDT value of 167 ºF for the most limiting RPV girth weld, as based on the calculations for RPV lower shell-to intermediate shell circumferential weld 9-442 (Material Heat 1P3571) using Regulatory Position C.2.1 in RG 1.99, Revision 2, and use of the referenced, credible sister plant data from the Kewaunee and Maine Yankee surveillance programs.

A 1/4T RTNDT value of 221 ºF and a 3/4T RTNDT value of 159 ºF for the most limiting RPV base metal or axial weld, as based on the calculations for RPV lower shell axial welds 3-442 A, B, and C (Tandem Material Heat 13253/12008(T)) using Position C.1.1 in RG 1.99, Revision 2. The 1/4T and 3/4T RTNDT calculations for these axial welds include an additional licensee-imposed safety margin of 3 ºF on RTNDT even after the margin term value of 65.5 ºF is incorporated into the RTNDT calculations for the welds.

P/T Limit Curves Review - P/T Limit Curves Conservatism Assessment and Determination The NRC staff confirmed that the P/T limit curves were based on the most limiting calculated P/T limit data points using the methods and applicable safety margins in ASME Code,Section XI, Appendix G and the applicable minimum temperature/lowest service temperature requirements mandated in Table 1 of Appendix G.

The NRC staff also confirmed that the licensees proposed heatup and cooldown curves in the April 7, 2020, application appropriately include a lowest service temperature/minimum temperature requirement of 60 ºF for RCS operating pressures less than or equal to 621.5 pounds per square inch (psi), and a minimum temperature requirement of 148 ºF for RCS operating pressures greater than 621.5 psi. Additionally, for the P/T limit curve that applies to plant operations when the reactor is in a critical fission process condition, the staff noted that the licensees curve conservatively sets a minimum temperature requirement (Tmin) of 281 ºF (i.e.,

Tmin = minimum temperature for the preservice hydrostatic pressure test > limiting RTNDT-Closure Flange of 188 ºF [28 ºF + 160 ºF]) for the assessed RCS operating pressures up to and inclusive of 1170 pounds per square inch gauge.

Based on this review, the NRC staff verified that the proposed P/T limit curves for 48 EFPY comply with: (1) the requirements for conservatisms and safety margins in P/T limit calculations specified in Section IV.A.2 of Appendix G and (2) the requirements for inclusion of appropriate minimum temperature requirements in P/T limit curves, as specified in Section IV.A.2 and Table 1 of Appendix G. Based on this review, the staff finds that the proposed P/T limit heatup, pressure test, and critical power operating curves provided for 48 EFPY in Figure 3.4.3-1 of the April 7, 2020, application and the proposed P/T limit cooldown curve for 48 EFPY in Figure 3.4.3-2 of the April 7, 2020, application are acceptable for implementation because the staff confirmed that the P/T limit curves comply with: (1) the provision in Section IV.A.2 of Appendix G that requires P/T limits to be at least as conservative as limits obtained by following

the methods of analysis and the margins of safety of Appendix G of Section XI of the ASME Code, and (2) the provision in Section IV.A.2 of Appendix G that requires P/T limit curves to incorporate the applicable minimum temperature requirements for P/T limit curves established in Table 1 of Appendix G.

3.2 LTOP Assessment The reactor vessel and connected components of the reactor coolant pressure boundary are designed to withstand the effects of neutron embrittlement from system pressure and temperature variations introduced by controlled heatup/cooldown operations and operational transients. Of the components exposed to neutron fluence, the reactor vessel is considered the most critical component susceptible to non-ductile failure because of the neutron fluence experienced over the vessel lifetime. ASME Code,Section XI, Appendix G defines the requirements for P/T limits and derives the LTOP P/T limits and the LTOP operational setpoint.

The NRC staff reviewed the Cold Overpressure Mitigating System (COMS) pressurizer power operated relief valve setpoints proposed changes to the TS requirements. The COMS setpoints were determined using the Westinghouse methodology described in WCAP-14040-A, Revision 4. The staff reviewed the limiting mass and energy analysis for the LTOP setpoints provided by the licensee. The staff determined that the LTOP setpoints ensure that the Appendix G requirements will be met, and that the limiting mass and energy analysis for the LTOP setpoints were determined with a level of detail consistent with NUREG-0800, Section 5.2.2. Based on the evaluation described above, the NRC staff has determined that the licensees LTOP System meets the criteria of RG 1.190 and GDCs 14, 30, and 31. Therefore, the NRC staff concludes that the results of the LTOP analysis for 48 EFPY are acceptable for incorporation into the TS.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Michigan State official was notified of the proposed issuance of the amendment on November 18, 2020. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes the requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. 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 amendment involves no significant hazards consideration in the Federal Register on June 2, 2020 (85 FR 33749), and there has been no public comment on such finding.

Accordingly, the amendment meets 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 amendment.

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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: J. Medoff, NRR J. Budynski, NRR Date: January 12, 2021

ML20329A001 *via e-mail OFFICE NRR/DORL/LPL3/PM* NRR/DORL/LPL3/LA* NRR/DNRL/NVIB/BC*

NAME SWall SRohrer (JBurkhardt for) HGonzalez (GCheruvenki for)

DATE 11/25/2020 11/25/2020 11/20/2020 OFFICE NRR/DSS/SNRB/BC* NRR/DSS/STSB/BC* OGC - NLO*

NAME RPatton VCusumano JWachutka DATE 11/24/2020 11/30/2020 12/22/2020 OFFICE NRR/DORL/LPL3/BC* NRR/DORL/LPL3/PM*

NAME NSalgado SWall DATE 01/12/2021 01/12/2021

v t e Letter Sequence Approval
EPID:L-2020-LLA-0081, Enclosure 7 - LTR-SCS-19-50, Revision 0, D.C. Cook Unit 1 Low Temperature Overpressure Protection System (Ltops) Analysis for 48 EFPY, Dated March 5, 2020, Attachment 2 Only (Non-Proprietary) (Approved, Closed)
Initiation Request Acceptance Administration Withholding Request Acceptance Questions Answers Results Approval Other:
MONTHYEAR2020-05-11 [Table View]

ML20329A001