ML25041A271

Issuance of Amendment Nos. 320 and 316 to Revise Technical Specification 3.4.3, Reactor Coolant System Pressure and Temperature Limits

ML25041A271
Person / Time
Site:	Catawba   (NPF-035, NPF-052)
Issue date:	03/14/2025
From:	Bryant J Plant Licensing Branch II
To:	Flippin N Duke Energy Carolinas
Stone Z
References
EPID L-2024-LLA-0082
Download: ML25041A271 (23)

Text

March 14, 2025 Nicole Flippin Site Vice President Catawba Nuclear Station Duke Energy Carolinas, LLC 4800 Concord Road York, SC 29745

SUBJECT:

CATAWBA NUCLEAR STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 320 AND 316 TO REVISE TECHNICAL SPECIFICATION 3.4.3, REACTOR COOLANT SYSTEM PRESSURE AND TEMPERATURE LIMITS (EPID L-2024-LLA-0082)

Dear Nicole Flippin:

The U.S. Nuclear Regulatory Commission has issued the following enclosed Amendment No. 320 to Renewed Facility Operating License No. NPF-35 and Amendment No. 316 to Renewed Facility Operating License No. NPF-52 for the Catawba Nuclear Station (Catawba),

Units 1 and 2, respectively. The amendments are in response to your application dated June 18, 2024.

The amendments revise Technical Specification (TS) 3.4.3, RCS [Reactor Coolant System]

Pressure and Temperature (P/T) Limits, to reflect an update to the P/T limit curves in Figures 3.4.3-1 RCS Heatup Limitations (UNIT 2 ONLY) and 3.4.3-2 RCS Cooldown Limitations (UNIT 2 ONLY) for Catawba, Unit 2. The proposed changes will also reflect that the revised Catawba, Unit 2 P/T limit curves in TS 3.4.3 are applicable until 54 effective full power years (EFPY). Although the TSs are common to both Catawba Units 1 and 2, the proposed changes are only applicable to Catawba, Unit 2. The applicable TS 3.4.3-1 RCS Heatup Limitations (UNIT 1 ONLY) and 3.4.3-2 RCS Cooldown Limitations (UNIT 1 ONLY) remain unchanged.

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions Federal Register notice.

If you have any questions, please contact me at (301) 415-0610 or via email at Jack.Minzerbryant@nrc.gov.

Sincerely,

/RA/

Jack Minzer Bryant, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-413 and 50-414

Enclosures:

1. Amendment No. 320 to NPF-35

2. Amendment No. 316 to NPF-52

3. Safety Evaluation cc: Listserv

DUKE ENERGY CAROLINAS, LLC NORTH CAROLINA ELECTRIC MEMBERSHIP CORPORATION DOCKET NO. 50-413 CATAWBA NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 320 Renewed License No. NPF-35

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility) Renewed Facility Operating License No. NPF-35 filed by the Duke Energy Carolinas, LLC (licensee), dated June 18, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as 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 set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-35 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 320, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Michael Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-35 and Technical Specifications Date of Issuance: March 14, 2025 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2025.03.14 10:49:04 -04'00'

DUKE ENERGY CAROLINAS, LLC NORTH CAROLINA MUNICIPAL POWER AGENCY NO. 1 PIEDMONT MUNICIPAL POWER AGENCY DOCKET NO. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 316 Renewed License No. NPF-52

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility) Renewed Facility Operating License No. NPF-52 filed by the Duke Energy Carolinas, LLC (licensee), dated June 18, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as 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 set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-52 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 316, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 120 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Michael Markley, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-52 and the Technical Specifications Date of Issuance: March 14, 2025 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2025.03.14 10:49:49 -04'00'

ATTACHMENT AMENDMENT NO. 320 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-35 AMENDMENT NO. 316 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-52 CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414 Renewed Facility Operating License Nos. NPF-35 and NPF-52 Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert NPF-35, page 4 NPF-35, page 4 NPF-52, page 4 NPF-52, page 4 Appendix A to Renewed Facility Operating License Nos. NPF-35 and NPF-52 Replace the following pages of the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert TS 3.4.3-4 TS 3.4.3-4 TS 3.4.3-6 TS 3.4.3-6 Renewed License No. NPF-35 Amendment No. 320 (2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 320, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.

(3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4),

following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

(4) Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license.

(5) Fire Protection Program Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program that complies with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated September 25, 2013; as supplemented by letters dated January 13, 2015; January 28, 2015; February 27, 2015; March 30, 2015; April 28, 2015; July 15, 2015; August 14, 2015; September 3, 2015; December 11, 2015; January 7, 2016; March 23, 2016; June 15, 2016; August 2, 2016; September 7, 2016; and, January 26, 2017, as approved in the SE dated February 8, 2017. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c),

the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

Renewed License No. NPF-52 Amendment No. 316 (2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 316, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.

(3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e)(4),

following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

(4) Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license.

(5) Fire Protection Program Duke Energy Carolinas, LLC shall implement and maintain in effect all provisions of the approved fire protection program that complies with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated September 25, 2013; as supplemented by letters dated January 13, 2015; January 28, 2015; February 27, 2015; March 30, 2015; April 28, 2015; July 15, 2015; August 14, 2015; September 3, 2015; December 11, 2015; January 7, 2016; March 23, 2016; June 15, 2016; August 2, 2016; September 7, 2016; and, January 26, 2017, as approved in the SE dated February 8, 2017. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c),

the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

RCS P/T Limits 3.4.3 Catawba Units 1 and 2 3.4.3-4 Amendment Nos. 320/316

RCS P/T Limits 3.4.3 Catawba Units 1 and 2 3.4.3-6 Amendment Nos. 320/316

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 320 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-35 AND AMENDMENT NO. 316 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-52 DUKE ENERGY CAROLINAS, LLC CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414

1.0 INTRODUCTION

By application dated June 18, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24170A696), Duke Energy Carolinas, LLC (Duke Energy or the licensee) requested changes to the Technical Specifications (TSs) for Catawba Nuclear Station, Units 1 and 2 (Catawba, Units 1 and 2). Although the TSs are common to both Catawba, Units 1 and 2, the proposed change is only applicable to Catawba, Unit 2. The applicable TS 3.4.3-1 Reactor Coolant System (RCS) Heatup Limitations (UNIT 1 ONLY) and 3.4.3-2 RCS Cooldown Limitations (UNIT 1 ONLY) remain unchanged. If approved, the proposed changes would permit the licensee to revise the pressure-temperature (P/T) limit curves for Catawba Unit 2 to be effective for licensed power operations up to and including 54 effective full power years (EFPY).

The license amendment request (LAR) included two Westinghouse Electric Company (WEC) non-proprietary WCAP reports as supporting information for the request: (1) WCAP-18843-NP, Revision 0, Verification of the Catawba Unit 2 Heatup and Cooldown Limit Curves for Normal Operation (Attachment 2 to the LAR), and (2) WCAP-15285, Revision 0, Catawba Unit 2 Heatup and Cooldown Limit Curves for Normal Operation using Code Case N-640 (Attachment 3 to the LAR). The licensee also included a Low Temperature Overpressure Protection (LTOP) setpoint analysis on page 5 of the Enclosure of the LAR.

2.0 REGULATORY EVALUATION

2.1

System Description

In Section 2.1, System Design and Operation, of the LAR, the licensee provided the following description:

All components of the CNS [Catawba Nuclear Station] Reactor Coolant System (RCS) are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients and reactor trips. CNS is required to limit the pressure and temperature changes during RCS heatup and cooldown within the design assumptions and the stress limits for cyclic operation.

The CNS TS contain P/T limit curves for heatup and cooldown, inservice leak and hydrostatic (ISLH) testing and data for the maximum rate of change of reactor coolant temperature. Each P/T limit curve defines an acceptable region for normal operation. The typical use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

Operating limits are established that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB).

CNS Updated Final Safety Analysis Report (UFSAR) Section 5.3.2 provides additional details regarding the methodology to develop the P/T limit curves that are contained in the CNS TS [ADAMS Package No. ML24296A046].

2.2 Licensees Proposed Changes Catawba TS 3.4.3 addresses RCS Pressure and Temperature (P/T) Limits. Limiting condition for operation (LCO) 3.4.3 states, in part, that:

RCS pressure and RCS temperature shall be limited during RCS heatup and cooldown, criticality, and inservice leak and hydrostatic testing in accordance with:

a. A maximum heatup rate as specified in Figure 3.4.3-1;

b. A maximum cooldown rate as specified in Figure 3.4.3-2; Figures 3.4.3-1 and 3.4.3-2 are two unit-specific figures. In its LAR, the licensee stated:

The proposed change is necessary because the existing CNS Unit 2 P/T limit curves in TS 3.4.3 are only applicable up to 34 EFPY. CNS Unit 2 is expected to reach 34 EFPY in August of 2025. A new set of P/T limit curves with a longer term of applicability is required.

TS 3.4.3, Figure 3.4.3-1, (UNIT 2 ONLY) RCS Heatup Limitations, proposed change is as follows:

The existing RCS heatup limitation curve is superseded entirely by a new curve applicable up to 54 EFPY.

The words Limiting ART at 34 EFPY are replaced with Limiting ART at 54 EFPY.

The 1/4-T value of 121 °F is revised to state 126 °F.

The 3/4-T value of 106 °F is revised to state 112 °F.

TS 3.4.3, Figure 3.4.3-2, (UNIT 2 ONLY) RCS Cooldown Limitations, proposed change is as follows:

The existing RCS cooldown limitation curve is superseded entirely by a new curve applicable up to 54 EFPY.

The words Limiting ART at 34 EFPY are replaced with Limiting ART at 54 EFPY.

The 1/4-T value of 121 °F is revised to state 126 °F.

The 3/4-T value of 106 °F is revised to state 112 °F.

2.3 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.

The regulation in 10 CFR 50.36(c)(2)(ii)(B) states that a TS 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 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 50.60, Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation, requires that all operating light-water nuclear power reactors (unless an exemption is granted, which is not applicable here) must meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, and 10 CFR Part 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements.

Appendix G requires, in part, that the P/T limits for an operating light-water nuclear power reactor be at least as conservative as the limits obtained by following the methods of analysis and margins of safety of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Appendix G, Fracture Toughness Criteria for Protection Against Failure, Division 1. Table 1 in 10 CFR Part 50, Appendix G, establishes the specific minimum temperature requirements that must be incorporated into the calculations of P/T limit curves. The regulation in 10 CFR Part 50, Appendix G,Section IV.A.2.a also states that the P/T limits and RCS minimum temperature requirement criteria for operating reactors are defined by the operating condition (i.e. hydrostatic pressure and leak tests, or normal operation including anticipated operational occurrences), the vessel pressure, whether or not fuel is in the vessel, and whether the core is critical. For components located in the beltline of the Reactor Pressure Vessel (RPV), 10 CFR Part 50, Appendix G,Section IV.A requires the values of adjusted reference temperature (ART) used in 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 by 10 CFR Part 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements. 10 CFR Part 50 Appendix G Section I.A states that the methods in ASME Section XI, Appendix G define one acceptable method for performing calculation of site-specific P/T limit curves.

The regulation in 10 CFR Part 50, Appendix H, requires light water reactor licensees to have a surveillance monitoring program for ferritic components located in the beltline of the RPV, if the RPVs are anticipated to operate with cumulative neutron fluence exposures in excess of 1 x 1017 neutrons per square centimeter (n/cm2) (E > 1.0 MeV). The regulation in 10 CFR Part 50, Appendix H, requires, in part, that for each surveillance capsule withdrawal; for reactor vessels purchased after 1982, the test procedure for the testing of capsule specimens must meet the requirements specified in ASTM Standard Practice E185-82 to the extent practicable for the configuration of the specimens in the capsules. Additionally, 10 CFR Part 50, Appendix H,Section IV.A 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 eighteen months of the capsule withdrawal, unless an extension is granted by the Director of the Office of Nuclear Reactor Regulation.

The NRC staff review was performed in consideration of the requirements contained in Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, that were determined to be applicable to the LAR; specifically, General Design Criteria (GDC):

Criterion 14, Reactor coolant pressure boundary, states, The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

Criterion 30, Quality of reactor coolant pressure boundary, states, Components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.

Criterion 31, Fracture prevention of reactor coolant pressure boundary, 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.

Catawba conformance with these GDC is reflected in their Updated Final Safety Analysis Report (UFSAR), Revision 24, Section 3.1, Conformance with General Design Criteria (ML24296A046).

2.4 Regulatory Guidance and NRC Staff-Approved Industry Guidelines That May Be Applied to P/T Limit or LTOP [low temperature overpressure protection] System Setpoint Calculations NRC Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, (ML003740284), describes a method acceptable to the NRC staff for calculating ART values of ferritic base metal or weld components located in the beltline of the RPV. This RG also provides a method that may be used to calculate attenuated neutron fluence values across the RPV wall thickness and a methodology for calculating the component-specific fluence factors that are used in the ART calculations.

NRC RG 1.190, Revision 0, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, (ML010890301) provides guidance on methods for determining RPV neutron fluence that are acceptable to the NRC staff, based on Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50.

NUREG-0800, Standard Review Plan, Section 5.2.2, Revision 3, Overpressure Protection, (ML070540076) provides, in part, guidance to the NRC staff for reviewing calculations of Pressurized Water Reactor (PWR) LTOP system pressure lift and system enable temperature (arming temperature) setpoints. Additional staff guidance for reviewing LTOP system setpoints is provided in NUREG-0800, Branch Technical Position (BTP) 5-2, Overpressurization Protection of Pressurized-Water Reactors While Operating at Low Temperatures, Revision 3 (ML ML070850008) NUREG-0800, Standard Review Plan, Section 5.3.2, Revision 2, Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock, (ML070380185), provides guidance to staff on how to review determinations of the P/T limits based on the methodology of the ASME Code,Section XI, Appendix G.

NRC Regulatory Issue Summary (RIS) 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components (ML14149A165), provides guidance that may be used to ensure that the scope of P/T limit evaluations includes all components in the RCPB that are made from ferritic steel materials, including assessment of RPV nozzle appurtenances that are made from ferritic steel materials and their ferritic nozzle-to-vessel pressure retaining welds. Licensing topical report PWROG-15109-NP-A, PWR Pressure Vessel Nozzle Appendix G Evaluation, Revision 0, (ML20024E573) provides an NRC-approved methodology licensees may reference in addressing the guidance provided in RIS 2014-11.

3.0 TECHNICAL EVALUATION

3.1 Evaluation of the Proposed Changes to the P/T Limit Curves The licensee describes its basis for meeting the requirements specified in 10 CFR Part 50, Appendix G, for P/T limit calculations in Section 5.3.1.5 of the Catawba UFSAR (ML24296A048). The licensee describes its unit-specific RPV material surveillance programs for Catawba, Unit 2, and its basis for meeting the RPV surveillance program requirements of 10 CFR Part 50, Appendix H, in UFSAR Section 5.3.1.6.

The NRC staff reviewed the proposed P/T limit curves for 54 EFPY in the referenced TS figures to ensure that the P/T limit curves meet the following regulatory objectives:

P/T limit curves given in LAR figures 3.4.3-1 and 3.4.3-2 (as based on 54 EFPY fluences) are at least as conservative and incorporate the minimum safety margin requirements specified in Appendix G of the ASME Boiler and Pressure Vessel Code,Section XI; ART values used in the development of the P/T limit curves have accounted for: (a) the increases in neutron irradiation (i.e., neutron fluence exposures) of the RPV beltline weld and base metal components needing assessment of ART, and (b) the results of the licensees 10 CFR Part 50, Appendix H RPV materials surveillance program; and proposed P/T limit curves have appropriately accounted for the minimum temperature requirements for P/T limit curves assessments in Section IV.A.2.b of the 10 CFR Part 50, Appendix G and that are specified in Table 1 of 10 CFR Part 50, Appendix G.

3.1.1 P/T Limit Curve Review - Neutron Fluence Assessment Determination of P/T limit curves involves three basic steps: (1) calculation of neutron fluence projection for a particular EFPY value, (2) determination of ARTs based on these fluence projections, and (3) determination of the P/T limit curves based on the updated ART values.

Then, the effect on the updated P/T limit curves on the LTOP settings is evaluated.

WCAP-15285 (Attachment 3 to the LAR) was developed in 1999 and contains P/T limit curves that WCAP-15286 describe as being applicable through 51 EFPY. After the development of the 51 EFPY P/T limit curves, an end-of-life extension term of 54 EFPY was established for Catawba Units 1 and 2 as part of the Catawba license renewal (ADAMS Package ML030850251). Specifically, in the NRCs safety evaluation report (SER) for the Catawba license renewal, the period of extended operation was assumed to go to 54 EFPY and was used for the time-limited aging analyses in Chapter 4 of the SER. In the SER for the license renewal, the NRC staff did not review P/T limit curves for application to the projected end of life, as stated in Section 4.2.3.2. Thus, in the license renewal, the NRC staff only approved the curves up to 51 EFPY. The NRC staff reviewed WCAP-15285 and WCAP-188843-NP only for the update of the Catawba Unit 2 P/T limit curve to 54 EFPY. Application of WCAP-15285 and WCAP-188843-NP for other purposes would require additional NRC review. In this application, the licensee provided WCAP-18843-NP (Attachment 2 to the LAR) to demonstrate that the 51 EFPY P/T limit curves are applicable to 54 EFPY through comparison of the 54 EFPY ART values, as calculated with the 54 EFPY fluence projections, and the 51 EFPY ART values.

Section 2 of WCAP-18843-NP describes neutron fluence projections to 54 EFPY and beyond.

The licensee used the WCAP-18843-NP fluence projections to 54 EFPY to calculate 54 EFPY ART values in later sections of WCAP-18843-NP and compared them against the 51 EFPY ART values in WCAP-15285.

The fluence calculations in WCAP-18843-NP were performed using the NRC-approved WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET (ML18204A010) and WCAP-18124-NP-A, Revision 0, Supplement 1-P/NP-A, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials (ADAMS Package ML22153A136). WCAP-18124-NP-A and its supplement were reviewed and approved for estimating the fluence to reactor vessel beltline and extended beltline materials based on their adherence to RG 1.190.

The NRC staff reviewed the information contained in Section 2 of WCAP-18843-NP. Based on its review, the NRC staff determined that the plant-specific fluence calculations were performed in a manner consistent with the NRC-approved methodology contained in WCAP-18124-NP-A and its supplement. This includes the 3D plant-specific geometric modeling, the use of fuel-cycle specific neutronic data for past operating cycles, and the use of 3D discreet ordinates transport methods to model neutron transport. Although benchmarking is addressed generically in WCAP-18124-NP-A and its supplement, the licensee also provided comparisons to Catawba, Unit 2 capsule dosimetry to confirm that the transport calculations agree with measured data within 20 percent, as recommended by RG 1.190.

Since the plant-specific calculation was performed in a manner consistent with an NRC-approved methodology in both the beltline and extended beltline that adheres to RG 1.190, the NRC staff determined that the plant-specific calculations are also consistent with the guidance contained in RG 1.190. Based on the above, the NRC staff determined that the fluence calculations are acceptable.

RIS 2014-11 (ML14149A165) addresses concerns related to radiation embrittlement of the nozzles during temperature changes. The Pressurized Water Reactor Owners Group (PWROG) provides generic bases to address the irradiation embrittlement of nozzles in topical report PWROG-15109-NP-A (ML20024E573), which the NRC staff has previously approved. The NRC staff safety evaluation approving PWROG-15109-NP-A (ADAMS Package ML19301D191) notes that nozzles do not need to be considered for a temperature shift in the ART if a fluence screening threshold of 4.28 x 1017 n/cm2 is not exceeded. The licensee compared the projected reactor pressure nozzle fluences to the 4.28 x 1017 n/cm2 screening criterion in order to determine whether the nozzles should be considered as potentially limiting in the development of the P/T limit curves. The Catawba, Unit 2 peak nozzle fluence at 54 EFPY was estimated to be 4.14 x 1016 n/cm2 in Table 2-6 of WCAP-18843-NP. This is a realistic (i.e., nominal) fluence estimate. If a 40% adder were conservatively applied, double the recommended uncertainty per RG 1.190 (i.e., 20% uncertainty), and the nozzle fluence would be 5.80 x 1016 n/cm2, which is still well below the screening criterion. The nozzle fluence could be underpredicted by over 600% and still not meet the 4.28 x 1017 n/cm2 threshold for explicit treatment as a potentially limiting component with regard to P/T limits. The NRC staff finds that the Catawba, Unit 2 nozzle welds do not require explicit treatment as reactor pressure vessel beltline materials at 54 EFPY exposure because the fluence was calculated with NRC-approved fluence methods, as previously described, and the peak nozzle fluence at 54 EFPY is well below the PWROG-15109-NP-A screening criterion. This is discussed further in section 3.1.3 below.

3.1.2 P/T Limit Curve Review - P/T Limit Conservatism Assessment The P/T limit curves for 54 EPFY proposed in TS Figure 3.4.3-1, (UNIT 2 ONLY) RCS Heatup Limitations, and TS Figure 3.4.3-2, (UNIT 2 ONLY) RCS Cooldown Limitations, of the LAR are based on the analysis documented in WCAP-18843-NP.

The NRC staff verified the accuracy and validity of these P/T limit calculations through the performance of: (a) a review of the information included in the LAR to address regulatory matters raised in RIS 2014-11, (b) review of the licensees analysis of ART, and (c) an independent P/T limit calculation for plant heatup transients.

3.1.3 P/T Limit Curve Review - RIS 2014-11 Review The licensee evaluated the impacts of RPV nozzles in Section 7 of WCAP-18843-NP. The licensee noted that if the plant-specific Catawba, Unit 2 nozzle fluence is less than the screening criterion of 4.28x1017 n/cm2, then the conclusions of PWROG-15109-NP-A are valid.

The licensee stated that it demonstrated that the screening criterion was met. The NRC reviewed the licensees fluence evaluation in Section 3.1.1 above and determined that it did meet the criterion. Since the licensee demonstrated that the conclusions of NRC-approved PWROG-15109-NP-A are valid for Catawba, Unit 2, the staff finds that the licensee appropriately accounted for potential nozzle impacts on the P/T limits.

3.1.4 P/T Limit Curve Review - ART Review The licensee determined Catawba, Unit 2 ART values in section 6 of WCAP-18843-NP, Tables 6-2 and 6-3. The NRC staff performed independent calculations of the 1/4-T and 3/4-T ART values to verify that the following limiting 1/4-T and 3/4-T ART values listed in Table 6-4 of WCAP-18843-NP were correctly calculated:

A limiting 1/4-T ART value of 125ºF for the intermediate shell plate B8605-2 as calculated according to Regulatory Position C.1.1 in RG 1.99, Revision 2.

A limiting 3/4-T ART value of 110ºF, for the intermediate shell plate B8605-2 as calculated according to Regulatory Position C.1.1 in RG 1.99, Revision 2.

The NRC staff independently calculated ART using the method in Regulatory Position C.1.1 in RG 1.99, Revision 2, which describes an acceptable means to the NRC staff for calculating ART values of ferritic base metal or weld components located in the beltline of the RPV. The NRC staff verified the licensees values are consistent with its independent calculation. The NRC staff also verified that the 1/4-T ART and 3/4-T values of the intermediate shell plate B8605-2 are the most limiting ART values. Based on its independent calculations, the NRC staff concludes that the ART values referenced above are valid, limiting 1/4-T and 3/4-T ART values for the P/T limit calculations that were performed by the licensee for 54 EFPY and that the ART values are acceptable with respect to implementation of the proposed P/T limit curves.

3.1.5 P/T Limit Curve Review - Analysis The licensee described its calculation of 51 EFPY P/T limit curves in Attachment 3 of the submittal, WCAP-15285. In Attachment 2 of the submittal, WCAP-18843, the licensee confirmed that the 51 EFPY P/T limit curves remain applicable for 54 EFPY. In Section 3.2 of WCAP-15285, the licensee referenced NRC-approved topical report WCAP-14040, Revision 3, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, (ML040620297) as the methodology for calculating P/T limit curves. The NRC staff performed an independent confirmatory calculation of the P/T limit curve for 100°F/hr heatup and noted that the licensees results were as least as conservative as the Section XI, Appendix G, G-2000, method assuming steady state conditions for an inside axial surface flaw. Given the licensees use of an NRC-approved methodology and the results of the staffs confirmatory analysis, the staff finds that the licensees proposed 54 EFPY P/T limit curves are acceptable and meet the requirements of 10 CFR Part 50, Appendix G and H.

3.2 Low Temperature Overpressure Protection (LTOP) Analysis Review The licensee is proposing to replace the current P/T Limits in TS 3.4.3, applicable to 34 EFPY, with new P/T limits applicable to 54 EFPY. As part of this LAR, the licensee also provided a discussion of the LTOP setpoint considerations to support the proposed P/T limits.

The NRC staff reviewed the LTOP setpoint considerations and determined that they were acceptable for supporting the proposed new P/T limits in TS 3.4.3, as described below.

3.2.1 LTOP Design Criteria Per Acceptance Criterion 4 on page 5.2.2-7 in Section 5.2.2 of NUREG-0800, Revision 3, the LTOP system should be designed in accordance with Branch Technical Position 5-2, which states that the LTOP system should be capable of relieving pressure during all anticipated overpressure events at a rate sufficient to satisfy the TS P/T limits while operating at low temperatures.

3.2.2 LTOP Analysis The LCO in TS 3.4.12 requires that for Catawba, Unit 2, the LTOP system be OPERABLE with a maximum of two pumps (charging pumps (CPs), safety injection (SI) pumps, or a combination of one CP and one SI pump) capable of injecting into the RCS with the accumulators isolated and reactor coolant pump (RCP) operation limited as specified in TS Table 3.4.12-1 and either (1), (2), or (3) as follows: (1) two Power Operated Relief Valves (PORVs) with nominal lift setting equal to 400 psig (as left calibrated), allowable value less than or equal to 425 psig (as found),

and with RCS cold leg temperature greater than or equal to 700F; or (2) two Residual Heat Removal (RHR) suction relief valves with lift settings greater than or equal to 417 psig and less than or equal to 509 psig with an indicated RCS cold leg temperature greater than or equal to 700F; or (3) a combination of any one PORV and one RHR suction relief valve, each with lift settings as described above.

In support of the proposed new P/T limits in TS 3.4.3, the licensee provided the results of the LTOP setpoint analysis for Catawba, Unit 2, on page 5 of the Enclosure to the LAR. The licensee analyzed the effects of the newly revised P/T limit curves for Catawba, Unit 2 to verify the existing LCOs in TS 3.4.12 remain valid with no changes required. The licensees comparison of the existing Unit 2 P/T Curves in TS 3.4.3 to the proposed curves showed that the allowable pressures across the analyzed temperature range have been reduced through 54 EFPY. Also, the licensee performed a comparison of the previously analyzed peak pressures for various LTOP events to the new P/T limit curves. The results confirmed that the existing setpoints for the PORVs and RHR pump suction relief valves and RCP operating restrictions in TS 3.4.12 remain adequate to ensure RCS pressure will not exceed limiting pressures illustrated in the new P/T limit curves through 54 EFPY. Based on the above, the NRC staff concludes TS 3.4.12 for the LTOP system adequately supports implementation of the new P/T limit curves and does not require revision.

3.2.3 LTOP Conclusion The NRC staff has reviewed the information provided in the licensees LAR submittal. Based on NRC staffs evaluation of the licensees analysis of the effects of the newly revised P/T curves for Catawba, Unit 2, on the LTOP setpoints in TS 3.4.12 as discussed in Section 3.2.2 of this safety evaluation, the NRC staff has concluded that: (1) the existing LTOP analysis adequately supports the limiting pressure for the heatup and cooldown curves at 54 EFPY; and (2) the proposed limiting pressures for the heatup and cooldown curves at 54 EFPY continue to meet 10 CFR 50.36(c)(2), which requires, in part, that TSs include items in the category of LCOs, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Therefore, the NRC staff determined that the limiting pressures for the heatup and cooldown curves were acceptable for plant operation to 54 EFPY for Catawba, Unit

2.

3.3 Technical Evaluation Conclusion

Based on the above, the NRC staff determined that (1) the proposed P/T curves in the Catawba, Unit 2 TS Figures 3.4.3-1 and 3.4.3-2 have considered all ferritic RPV materials consistent with RIS 2014-11, (2) the proposed P/T curves are developed based on the methodology in NRC-approved topical report WCAP-14040; and the ASME Code,Section XI, Appendix G, and (3) the proposed P/T curves satisfy applicable requirements in 10 CFR Part 50, Appendices G and H, as discussed in Section 3.1 above, and as a result meet the requirements of 10 CFR 50.60. Therefore, the NRC staff concludes that the proposed P/T curves are acceptable to be incorporated into the Catawba, Unit 2 TS Figures 3.4.3-1 and 3.4.3-

2. Further, as discussed above, the NRC staff concludes that the proposed changes continue to meet the requirements of 10 CFR 50.36.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the South Carolina State official was notified of the proposed issuance of the amendment on February 10, 2025. On March 4, 2025, the State official confirmed that the State of South Carolina had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change the requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration published in the Federal Register on September 3, 2024 (89 FR 71433), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: SSun, NRR JMessina, NRR JVande Polder, NRR MBenson, NRR JTsao, NRR RGrover, NRR Date: March 14, 2025

ML25041A271 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DNRL/NVIB/BC NAME JMinzerBryant KZeleznock ABuford DATE 02/10/2025 02/12/2025 02/10/2025 OFFICE NRR/DSS/SNSB/BC NRR/DSS/SFNB/BC NRR/DSS/STSB/BC NAME DMurdock SKrepel SMehta DATE 01/27/2025 02/03/2025 02/21/2025 OFFICE OGC - NLO NRR/DORL/LPL2-1/BC NRR/DORL/LPL2-1/PM NAME CRyan MMarkley JMinzerBryant DATE 03/06/2025 03/14/2025 03/14/2025