ML25254A064

Issuance of Amendment Nos. 226, 226, and 226 to Adopt Westinghouse Thermal Design Procedure Topical Report WCAP-18240-P-A

ML25254A064
Person / Time
Site:	Palo Verde
Issue date:	09/22/2025
From:	William Orders Plant Licensing Branch IV
To:	Heflin A Arizona Public Service Co
Orders, William
References
EPID L-2025-LLA-0015 WCAP-18240-P-A
Download: ML25254A064 (1)

Text

September 22, 2025 Mr. Adam Heflin Executive Vice President/

Chief Nuclear Officer Mail Station 7605 Arizona Public Service Company P.O. Box 52034 Phoenix, AZ 85072-2034

SUBJECT:

PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 -

ISSUANCE OF AMENDMENT NOS. 226, 226, AND 226 TO ADOPT WESTINGHOUSE THERMAL DESIGN PROCEDURE TOPICAL REPORT WCAP-18240-P-A (EPID L-2025-LLA-0015)

Dear Mr. Heflin:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued enclosed Amendment No. 226 to Renewed Facility Operating License No. NPF-41, Amendment No. 226 to Renewed Facility Operating License No. NPF-51, and Amendment No. 226 to Renewed Facility Operating License No. 74 for the Palo Verde Nuclear Generating Station, Units 1, 2, and 3 (Palo Verde), respectively. The amendments consist of changes to the technical specifications (TSs) in response to your application dated January 17, 2025, as supplemented by letter dated July 3, 2025.

The amendments revise Palo Verde TS 5.6.5.b in TS section 5.6.5, Core Operating Limits Report (COLR), by adding the NRC-approved Westinghouse Topical Report WCAP-18240-P-A, Westinghouse Thermal Design Procedure (WTDP), to the list of referenced analytical methods for the determination of reactor core operating limits.

Consistent with WCAP-18240-P-A, the amendments authorize the option for discretionary use of a single NRC-approved thermal-hydraulic subchannel code in the core protection calculator system and core operating limit supervisory system setpoints analyses, in lieu of the two NRC-approved thermal-hydraulic codes that are currently used.

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

Sincerely,

/RA/

William Orders, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-528, STN 50-529, and STN 50-530

Enclosures:

1. Amendment No. 226 to NPF-41

2. Amendment No. 226 to NPF-51

3. Amendment No. 226 to NPF-74

4. Safety Evaluation cc: Listserv

ARIZONA PUBLIC SERVICE COMPANY, ET AL.

DOCKET NO. STN 50-528 PALO VERDE NUCLEAR GENERATING STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 226 License No. NPF-41

1.

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

A.

The application for amendment by the Arizona Public Service Company (APS) on behalf of itself and the Salt River Project Agricultural Improvement and Power District, El Paso Electric Company, Southern California Edison Company, Public Service Company of New Mexico, Los Angeles Department of Water and Power, and Southern California Public Power Authority dated January 17, 2025, as supplemented by letter dated July 3, 2025,complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions regulations set forth in 10 CFR Chapter I.

B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Renewed Facility Operating License No. NPF-41 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 226, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this renewed operating license. APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

3.

This license amendment is effective as of the date of issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Tony Nakanishi, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-41 and the Technical Specifications Date of Issuance: September 22, 2025 TONY NAKANISHI Digitally signed by TONY NAKANISHI Date: 2025.09.22 11:34:55 -04'00' ARIZONA PUBLIC SERVICE COMPANY, ET AL.

DOCKET NO. STN 50-529 PALO VERDE NUCLEAR GENERATING STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 226 License No. NPF-51 1.

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

A.

The application for amendment by the Arizona Public Service Company (APS) on behalf of itself and the Salt River Project Agricultural Improvement and Power District, El Paso Electric Company, Southern California Edison Company, Public Service Company of New Mexico, Los Angeles Department of Water and Power, and Southern California Public Power Authority dated January 17, 2025, as supplemented by letter dated July 3, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions regulations set forth in 10 CFR Chapter I.

B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Renewed Facility Operating License No. NPF-51 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 226, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this renewed operating license. APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

3.

This license amendment is effective as of the date of issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Tony Nakanishi, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-51 and the Technical Specifications Date of Issuance: September 22, 2025 TONY NAKANISHI Digitally signed by TONY NAKANISHI Date: 2025.09.22 11:33:53 -04'00' ARIZONA PUBLIC SERVICE COMPANY, ET AL.

DOCKET NO. STN 50-530 PALO VERDE NUCLEAR GENERATING STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 226 License No. NPF-74 1.

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

A.

The application for amendment by the Arizona Public Service Company (APS) on behalf of itself and the Salt River Project Agricultural Improvement and Power District, El Paso Electric Company, Southern California Edison Company, Public Service Company of New Mexico, Los Angeles Department of Water and Power, and Southern California Public Power Authority dated January 17, 2025, as supplemented by letter dated July 3, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions regulations set forth in 10 CFR Chapter I.

B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Renewed Facility Operating License No. NPF-74 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.226, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this renewed operating license. APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

3.

This license amendment is effective as of the date of issuance and shall be implemented within 90 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Tony Nakanishi, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-74 and the Technical Specifications Date of Issuance: September 22, 2025 TONY NAKANISHI Digitally signed by TONY NAKANISHI Date: 2025.09.22 11:34:15 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NOS. 226, 226, AND 226 TO RENEWED FACILITY OPERATING LICENSE NOS. NPF-41, NPF-51, AND NPF-74 PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 DOCKET NOS. STN 50-528, STN 50-529, AND STN 50-530 Replace the following pages of Renewed Facility Operating Licenses Nos. NPF-41, NPF-51, and NPF-74, and the 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.

Renewed Facility Operating License No. NPF-41 REMOVE INSERT 5

5 Renewed Facility Operating License No. NPF-51 REMOVE INSERT 6

6 Renewed Facility Operating License No. NPF-74 REMOVE INSERT 4

4 Technical Specifications REMOVE INSERT 5.6-7 5.6-7

Renewed Facility Operating License No. NPF-41 Amendment No. 226 (1)

Maximum Power Level Arizona Public Service Company (APS) is authorized to operate the facility at reactor core power levels not in excess of 3990 megawatts thermal (100% power), in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 226, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this renewed operating license.

APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

(3)

Antitrust Conditions This renewed operating license is subject to the antitrust conditions delineated in Appendix C to this renewed license.

(4)

Operating Staff Experience Requirements Deleted (5)

Post-Fuel-Loading Initial Test Program (Section 14, SER and SSER 2)*

Deleted (6)

Environmental Qualification Deleted (7)

Fire Protection Program APS shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility, as supplemented and amended, and as approved in the SER through Supplement 11, subject to the following provision:

APS may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

• The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

Renewed Facility Operating License No. NPF-51 Amendment No. 226 (1)

Maximum Power Level Arizona Public Service Company (APS) is authorized to operate the facility at reactor core power levels not in excess of 3990 megawatts thermal (100% power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 226, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this renewed operating license.

APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

(3)

Antitrust Conditions This renewed operating license is subject to the antitrust conditions delineated in Appendix C to this renewed operating license.

(4)

Operating Staff Experience Requirements (Section 13.1.2, SSER 9)*

Deleted (5)

Initial Test Program (Section 14, SER and SSER 2)

Deleted (6)

Fire Protection Program APS shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility, as supplemented and amended, and as approved in the SER through Supplement 11, subject to the following provision:

APS may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

(7)

Inservice Inspection Program (Sections 5.2.4 and 6.6, SER and SSER 9)

Deleted

• The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

Renewed Facility Operating License No. NPF-74 Amendment No. 226 (4)

Pursuant to the Act and 10 CFR Part 30, 40, and 70, APS to receive, possess, and use in amounts required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Pursuant to the Act and 10 CFR Parts 30, 40, and 70, APS 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 Commissions regulations set forth in 10 CFR Chapter I 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 Arizona Public Service Company (APS) is authorized to operate the facility at reactor core power levels not in excess of 3990 megawatts thermal (100% power), in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 226, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this renewed operating license.

APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

(3)

Antitrust Conditions This renewed operating license is subject to the antitrust conditions delineated in Appendix C to this renewed operating license.

(4)

Initial Test Program (Section 14, SER and SSER 2)

Deleted (5)

Additional Conditions The Additional Conditions contained in Appendix D, as revised through Amendment No. 212, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Additional Conditions.

Reporting Requirements 5.6 5.6 Reporting Requirements PALO VERDE UNITS 1,2,3 5.6-7 (continued) 5.6.5 Core Operating Limits Report (COLR) (continued) 20.

CENPD-382-P-A, "Methodology for Core Designs Containing Erbium Burnable Absorbers." [Methodology for Specifications 3.1.1, Shutdown Margin-Reactor Trip Breakers Open; 3.1.2, Shutdown Margin-Reactor Trip Breakers Closed; and 3.1.4, Moderator Temperature Coefficient.]

21.

CEN-386-P-A, "Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/kgU for Combustion Engineering 16 x 16 PWR Fuel."

[Methodology for Specifications 3.1.1, Shutdown Margin-Reactor Trip Breakers Open; 3.1.2, Shutdown Margin-Reactor Trip Breakers Closed; and 3.1.4, Moderator Temperature Coefficient.]

22.

WCAP-16500-P-A, "CE 16x16 Next Generation Fuel Core Reference Report." [Methodology for Specifications 2.1.1, Reactor Core SLs; 3.2.4, DNBR]

23.

WCAP-14565-P-A, "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis."

[Methodology for Specifications 2.1.1, Reactor Core SLs; 3.2.4, DNBR]

24.

CENPD-387-P-A, "ABB Critical Heat Flux Correlations for PWR Fuel."

[Methodology for Specifications 2.1.1, Reactor Core SLs; 3.2.4, DNBR]

25.

WCAP-16523-P-A, "Westinghouse Correlations WSSV and WSSV-T for Predicting Critical Heat Flux in Rod Bundles with Side-Supported Mixing Vanes." [Methodology for Specifications 2.1.1, Reactor Core SLs; 3.2.4, DNBR]

26.

WCAP-16072-P-A, "Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs."

[Methodology for Specifications 2.1.1, Reactor Core SLs; 3.2.4, DNBR]

27.

EMF-2103P-A, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors." [Methodology for Specification 3.2.1, Linear Heat Rate]

28.

EMF-2328 (P)(A), "PWR Small Break LOCA Evaluation Model, S-RELAP5 Based." [Methodology for Specification 3.2.1, Linear Heat Rate]

29.

BAW-10231P-A, "COPERNIC Fuel Rod Design Computer Code."

[Methodology for Specification 3.2.1, Linear Heat Rate]

30.

BAW-10241(P)(A), "BHTP DNB Correlation Applied with LYNXT."

[Methodology for Specification 3.2.4, DNBR]

31.

EPRI-NP-2511-CCM-A, "VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores." [Methodology for Specification 3.2.4, DNBR]

32.

WCAP-18240-P-A, Westinghouse Thermal Design Procedure (WTDP).

[Methodology for Specifications 2.1.1, Reactor Core SLs; 3.2.4, DNBR; and 3.2.5, Axial Shape Index]

AMENDMENT NO. 212, 226 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 226, 226, AND 226 TO RENEWED FACILITY OPERATING LICENSE NOS. NPF-41, NPF-51, AND NPF-74 ARIZONA PUBLIC SERVICE COMPANY, ET AL.

PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 DOCKET NOS. STN 50-528, STN 50-529, AND STN 50-530

1.0 INTRODUCTION

By letter dated January 17, 2025 (Reference 1), as supplemented by letter dated July 3, 2025 (Reference 2), Arizona Public Service Company (APS, the licensee) submitted a license amendment request (LAR) to the U.S. Nuclear Regulatory Commission (NRC, the Commission) requesting changes to the Technical Specifications (TSs) for Palo Verde Nuclear Generating Station (Palo Verde or PVNGS), Units 1, 2, and 3.

The proposed amendments would revise Palo Verde TS 5.6.5.b in TS section 5.6.5, Core Operating Limits Report (COLR), by adding the NRC-approved Westinghouse Topical Report (TR) WCAP-18240-P-A, Westinghouse Thermal Design Procedure (WTDP) (Reference 3), to the list of referenced analytical methods. Consistent with TR WCAP-18240-P-A, the proposed change would authorize the option for discretionary use of a single NRC-approved thermal-hydraulic (T-H) subchannel code in the core protection calculator system (CPCS) and core operating limit supervisory system (COLSS) setpoint analyses, in lieu of the two NRC-approved T-H codes that are currently used.

The supplemental letter dated July 3, 2025 provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register (FR) on May 13, 2025 (90 FR 20326).

2.0 REGULATORY EVALUATION

2.1 Plant Description The Palo Verde Units 1, 2, and 3 are Combustion Engineering (CE) designed pressurized water reactors (PWRs). As described in the Updated Final Safety Analysis Report (UFSAR),

section 1.2.3.3 (Reference 4), the reactor coolant system is arranged as two closed loops connected in parallel to the reactor vessel. Each loop consists of one 42-inch internal diameter outlet (hot) pipe, one steam generator, two 30-inch internal diameter inlet (cold) pipes, and two

pumps. An electrically heated pressurizer is connected to one of the loops, and safety injection lines are connected to each of the four cold legs and two hot legs. As described in UFSAR, section 1.1.3, the containment for each unit is a single containment system consisting of a steel-lined prestressed concrete, cylindrical structure, with a hemispherical dome.

2.2 TS Changes The proposed TS change will revise TS 5.6.5.b by updating the references to be consistent with the analytical methods that will be used to determine the core operating limits following WTDP implementation. The proposed change will add the following NRC-approved TR to the list of referenced analytical methods:

WCAP-18240-P-A, Westinghouse Thermal Design Procedure (WTDP).

[Methodology for Specifications 2.1.1, Reactor Core SLs; 3.2.4, DNBR; and 3.2.5, Axial Shape Index]

2.3 Current and Proposed TS and Reason for Change The licensee currently uses the NRC-approved CEN-356(V)-P-A, Revision 01-P-A, Modified Statistical Combination of Uncertainties, methodology (Reference 5) for determining the departure from nucleate boiling (DNB) ratio (DNBR) specified acceptable fuel design limit (SAFDL) (i.e., a 95/95 DNBR analytical limit) and associated DNBR probability density function (PDF). This methodology is based on the modified statistical combination of uncertainties (MSCU). The CE designed nuclear steam supply system (CE-NSSS) digital setpoint reload analysis methodology within CEN-356(V)P-A, Revision 01-P-A, was augmented by TR WCAP-16500-P-A, Supplement 1, Revision 1, Application of CE Setpoint Methodology for CE 16x16 Next Generation Fuel (NGF), December 2010 (Reference 6), with the introduction of Westinghouse CE16NGF fuel as a replacement for Westinghouse CE16STD fuel.

The NRC-approved TR WCAP-18240-P-A (or WTDP) describes a new methodology proposed for determining the DNBR SAFDL and associated DNBR PDF for anticipated operational occurrences (AOOs) and subsequently using this limit and PDF for calculating the number of fuel rods that would be expected to be damaged due to DNB during postulated accidents.

The licensee stated that the reason for the TS change is that the adoption of the WTDP methodology will allow non-loss-of-coolant accident (non-LOCA) DNB fuel failure analyses for both Westinghouse CE16NGF fuel and Framatome CE16HTP fuel. Additionally, the adoption of WTDP includes the option for discretionary use of a single NRC-approved T-H subchannel code in lieu of two NRC-approved T-H codes in uncertainty evaluations and DNBR calculations that are part of transient and setpoint analyses.

2.4 Regulations and Guidance 2.4.1 Regulations The NRC staff considered the following General Design Criteria (GDC) in Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and

Utilization Facilities, Appendix A, General Design Criteria for Nuclear Power Plants, and in 10 CFR Part 100, Reactor Site Criteria, in its review of the proposed change:

GDC 10, Reactor design, which states that, [t]he reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

GDC 19, Control room, insofar as adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

10 CFR Part 100 insofar as the specified radiation dose limits to individuals during a postulated fission product release is concerned.

2.4.2 Regulatory Guidance The NRC staff considered the following sections of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]

Editions in its review of the proposed change.

Section 4.2, Revision 3, :Fuel System Design, (Reference 7)

Section 4.4, Revision 2, Thermal and Hydraulic Design, (Reference 8) Acceptance Criterion II.1 which is based on meeting the relevant requirements of GDC 10 2.5 Precedents Palo Verde is the first CE plant to implement the WTDP methodology.

3.0 TECHNICAL EVALUATION

WTDP is a T-H analysis methodology for determining the DNBR limit for AOOs. The WTDP TR describes a method for calculating a statistical limit on DNBR, below which fuel failure may occur. The TR also describes a method for using the statistical DNBR limit to determine the number of rods that would be expected to be damaged due to DNB during an accident. The methodology is intended to be applicable to PWRs, including those with CE and Westinghouse-designed NSSSs. The WTDP methodology utilizes the NRC-approved WCAP-14565-P-A (Reference 9) methodology based on VIPRE-W or VIPRE-01 subchannel computer codes to evaluate the potential for DNB. and consequential fuel rod failure due to overheating of fuel rod cladding. The VIPRE-W and VIPRE-01 codes model conditions exterior to and at the cladding surface including, but not limited to, reactor coolant system pressure and temperature, reactor coolant subchannel flow rate, and heat flux at the cladding surface.

3.1 Current Methods The current methods used for DNBR SAFDL analysis are as follows:

CEN-283(S)-P, Part I, Statistical Combination of Uncertainties (Reference 10)

CEN-283(S)-P, Part II, Statistical Combination of Uncertainties (Reference 11)

CEN-356(V)-P-A, Modified Statistical Combination of Uncertainties (Reference 5)

WCAP-16500-P-A, Supplement 1, Revision 1, Application of CE Setpoint Methodology for CE 16x16 Next Generation Fuel (NGF) (Reference 6)

The current method CEN-283(S)-P, Part I, which is a part of the NRC-approved WTDP methodology, describes the statistical combination of system parameter uncertainties in thermal margin analyses for San Onofre Nuclear Generating Station (SONGS) Units 2 and 3. A detailed description of the uncertainty probability distributions and response surface methodologies for the DNBR SAFDL determination is provided in the report.

The current method CEN-283(S)-P, Part II, also a part of the NRC-approved WTDP methodology, describes the methodology used for statistically combining uncertainties involved in the determination of the linear heat rate and DNBR limiting safety system setting (LSSS) for SONGS, Units 2 and 3, and for CE-NSSS 80 plants, such as Palo Verde, Units 1, 2, and 3. It describes the statistical combination of state parameter and modeling uncertainties for the determination of the LSSS overall uncertainty factors related to the CETOP-D code applications.

The current method CEN-356(V)-P-A describes a methodology change to statistically combine uncertainty components from two groups of system parameters related to fuel type (e.g., fuel rod pitch, fuel rod outside diameter, and applicable critical heat flux (CHF) correlations) and state parameters related to plant operating conditions (e.g., core power distribution and reactor coolant pressure, temperature, and flow rate) to obtain overall uncertainty factors in determining the LSSS and limiting condition for operation (LCO) for the Palo Verde COLSS and CPCS. This methodology has been referenced and used for existing CE-NSSS safety analyses and reload evaluations.

The current setpoint methodology, which uses two NRC-approved T-H codes (e.g., CETOP-D (Reference 12) and either TORC (Reference 13) or VIPRE [WCAP-14565-P-A (Reference 9)])

is based on NRC-approved CEN-356(V)-P-A, Revision 01-P-A (MCSU methodology)

(Reference 5) for combining uncertainty terms to calculate COLSS DNB power operating limit and CPCS DNBR addressable uncertainty constants. This is done using the CETOP-D code along with a set of correction factors which allows it to calculate DNBR values conservative to either the TORC or VIPRE code. This methodology allows for implementation of a rod bow penalty in the overall uncertainty factors. The MSCU methodology was augmented by the NRC-approved TR WCAP-16500-P-A, Supplement 1, Revision 1 (Reference 6), This TR describes application of the CE-NSSS setpoint methodology including the MSCU process to the CE16NGF reload evaluations. TR WCAP-16500-P-A, and its supplement applies to core reloads for which the CHF correlation (CE-1), is different than the CHF correlations ABB-NV and WSSV for Westinghouse CE16NGF fuel used in fuel design. The NRC staff subsequently approved the use of TR WCAP-16500-P-A, Supplement 1, methodology for Palo Verde, Units 1, 2, and 3 core designs that utilize Framatome CE16HTP fuel and the BHTP CHF correlation in Amendment Nos. 212, 212, and 212 dated March 4, 2020 (Reference 14).

3.2 Proposed Methods in WTDP The WTDP TR describes a method for determining a DNBR SAFDL (i.e., a 95/95 DNBR analytical limit) below which fuel failure may occur. It also describes a method for using a DNBR SAFDL and associated DNBR PDF to calculate the number of fuel rods damaged due to the occurrence of DNB during a postulated accident.

The WTDP TR also describes a method for discretionary use of a single NRC-approved T-H subchannel code (e.g., VIPRE (VIPRE-W or VIPRE-01)) in lieu of the use of two NRC-approved T-H codes (e.g., CETOP-D and either TORC or VIPRE) in uncertainty evaluations and DNBR calculations that are part of transient and setpoint analyses.

Section 2.0 of WTDP TR describes the methodology for determining DNBR SAFDL, which uses the statistical Monte Carlo approach. The DNBR calculations are performed using a T-H subchannel code and a DNB correlation already approved for safety analysis and licensing applications. More details are provided in response to RAI-WTDP-01, in section D of the WTDP TR.

Section 3 of WTDP TR describes the statistical methodology for determining number of fuel rods in DNB in non-LOCA infrequent (American National Standards Institute (ANSI) Condition III and IV) events. The method for using a DNBR SAFDL and associated DNBR PDF to determine the number of fuel rods in DNB is based on NRC-approved TR CENPD-183-A (Reference 15) that is currently approved for use by the licensee for Palo Verde Units 1, 2, and 3.

Section 3.1.2 of WTDP TR discusses the option of using a single NRC-approved T-H subchannel code in both uncertainty evaluations and DNBR calculations that are part of transient and setpoint analyses using the MSCU methodology. The WTDP allows for discretionary use of a single NRC-approved T-H subchannel code in uncertainty evaluations that are part of CPCS and COLSS setpoint analyses. The licensee stated that if it chooses the single code option, then the setpoint analysis will be performed using the NRC-approved VIPRE code with its models depending on the fuel type that utilizes the NRC-approved CE-1, ABB-NV, WSSV, and BHTP CHF correlations. Therefore, with WTDP methodology, the use of the CETOP-D code with its correction factors to VIPRE/TORC in an overall uncertainty analysis (OUA) can be replaced with direct use of an NRC-approved T-H subchannel code (e.g., VIPRE) in the OUA. The licensee stated that since the VIPRE code is a detailed T-H model, there may not be a need for the correction factors that were used as input to OUAs with the CETOP-D code. However, the capability to have correction factors will be maintained to allow the flexibility to use a cycle-independent VIPRE model instead of a cycle-specific VIPRE model. As such, the use of correction factors allows the cycle independent VIPRE model to be conservative to the cycle-specific VIPRE model. The option for discretionary use of a single NRC-approved T-H subchannel code (e.g., VIPRE (VIPRE-W or VIPRE-01)) in lieu of two codes result in a simplification of the MSCU interface and process improvement by eliminating use of the CETOP-D code. The licensee stated that such simplification would not change the current MCSU setpoint methodology as augmented by TR WCAP-16500-P-A, Supplement 1, Revision 1 (Reference 6). For consistency with the current methodology, the fuel rod bow methods are not changed.

3.3 Technical Specifications Impact 3.3.1 TS 2.1.1.1 - DNBR Safety Limit The margin to DNB is quantified through DNBR, which is defined as the ratio of the CHF at a specific location to the operating local heat flux at that location. GDC 10 of Appendix A to 10 CFR Part 50, requires that SAFDLs are not exceeded during normal operation and AOOs. In Modes 1 and 2, TS Safety Limit (SL) 2.1.1.1 and TS 3.3.1, table 3.3.1-1 requires DNBR be maintained greater than or equal to () 1.34. The licensee stated that for the fuel types used at Palo Verde, Units 1, 2, and 3, the NRC has approved the CHF correlation limits given below.

These correlation limits are design basis limits for fission product barriers (DBLFPBs) related to the fuel rod cladding heat flux at which DNB may occur.

1.19 using the CE-1 CHF correlation for Westinghouse CE16STD fuel.

1.13 using the ABB-NV CHF correlation for Westinghouse CE16STD and CE16NGF fuel.

1.12 using the WSSV CHF correlation for Westinghouse CE16NGF fuel.

1.110 using the BHTP CHF correlation for Framatome CE16HTP fuel.

The DBLFPBs provide a 95 percent probability at a 95 percent confidence level of not experiencing DNB on the hot rod during normal operation and AOOs. The CHF correlation limit DBLFPBs are statistically combined with system parameters and state parameters to generate cycle-specific DNBR SAFDLs and associated DNBR PDFs for candidate limiting fuel assemblies. The DNBR SAFDLs prevent overheating of any fuel rod in the reactor core due to reaching DNB. Adoption of the WTDP methodology does not require changes to TS SL 2.1.1.1.

In the plant CPCS, the licensee has installed an LSSS DNBR value of 1.34 that uses the CE-1 CHF correlation. In the cycle-specific setpoints analyses, the licensee establishes overall uncertainty factors and installs them in the CPCS so that the LSSS DNBR value ensures that the cycle-specific DNBR SAFDLs are protected. The cycle-specific DNBR SAFDLs are dependent on the fuel types present in the reactor core:

1. For a core where all candidate limiting fuel assemblies are CE16STD fuel, the cycle-specific DNBR SAFDL is determined using the CE-1 or ABB-NV CHF correlation with MSCU methodology.

2. For a core where all candidate limiting fuel assemblies are CE16NGF fuel, the cycle-specific DNBR SAFDL is determined using the WSSV and ABB-NV CHF correlations with either MSCU or WTDP methodology.

3. For a core where all candidate limiting fuel assemblies are CE16HTP fuel, the cycle-specific DNBR SAFDL is determined using the BHTP CHF correlation with either MSCU or WTDP methodology.

4. For a transition or mixed core where candidate limiting fuel assemblies are of different fuel types, cycle-specific DNBR SAFDLs are determined for each fuel type using the appropriate CHF correlations with either MSCU or WTDP methodology.

The licensee will retain the CE-1 correlation and the DNBR-low trip setpoint and Allowable Value of 1.34 in the CPCS algorithm, even though the WSSV, ABB-NV, and BHTP CHF correlations will be used in the cycle-specific safety and setpoint analyses. Based on the cycle-

specific setpoints analyses, the licensee establishes the overall uncertainty factors and installs them in the CPCS so that the LSSS DNBR value ensures that the cycle-specific DNBR SAFDLs are protected.

The NRC staff finds it acceptable that the adoption of the WTDP methodology does not require changes to TS SL 2.1.1.1, because in the plant CPCS, the licensee has installed an LSSS DNBR TS value of 1.34 that uses the CE-1 CHF correlation.

3.3.2 TS 3.2.4, Departure from Nucleate Boiling Ratio (DNBR)

TS 3.2.4, Departure From Nucleate Boiling Ratio (DNBR), addresses maintaining DNBR when COLSS is either in-service or out-of-service. The purpose of TS LCO 3.2.4 is to limit the core power distribution to the initial value assumed in the accident analyses. This LCO limits the amount of damage to the fuel cladding during an accident by ensuring that the plant is operating within acceptable conditions at the onset of a transient. The DNB correlations (ABB-NV, WSSV, BHTP, and CE-1) used and their range of applicability are not being changed. Therefore, adoption of the WTDP methodology does not require changes to TS 3.2.4.

The NRC staff finds that the adoption of the WTDP methodology does not affect TS 3.2.4 because the licensee did not add a new or change any of the DNB correlations already in the COLSS 3.3.3 TS 3.3.1, Reactor Protective System (RPS) Instrumentation - Operating TS 3.3.1, Reactor Protective System (RPS) Instrumentation - Operating, addresses the RPS during normal power operations. The RPS initiates a reactor trip to protect against violating the core specified acceptable fuel design limits and breaching the reactor coolant pressure boundary during selected AOOs. During AOOs, the acceptable limits include the DNBR SL of 1.34.

The NRC staff finds that adoption of the WTDP methodology does not affect TS 3.3.1 because there is no change in the fuel design limits.

3.3.4 TS 4.2.1, Fuel Assemblies TS 4.2.1, Fuel Assemblies, addresses the reactor core fuel assemblies. TS 4.2.1.a states, in part, that [e]ach unit-specific Core Operating Limits Report (COLR) shall contain an identification of the fuel types and cladding material in the reactor, and the associated COLR methodologies (e.g., inclusion of WTDP Topical Report WCAP-18240-P-A). Adoption of the WTDP methodology does not require changes to TS 4.2.1.

The NRC staff finds that adoption of WTDP methodology does not affect TS 4.2.1 because none of its contents require any change.

3.3.5 TS 5.6.5, Core Operating Limits Report (COLR)

TS 5.6.5.b lists the documents previously reviewed and approved by the NRC that describe the COLR analytical methods used to determine the core operating limits presented in each Palo Verde unit-specific COLR. The proposed change will add NRC-approved WTDP TR WCAP-18240-P-A to the list of referenced analytical methods, which states: :

WCAP-18240-P-A, Westinghouse Thermal Design Procedure (WTDP).

[Methodology for Specifications 2.1.1, Reactor Core SLs; 3.2.4, DNBR; and 3.2.5, Axial Shape Index.]

The proposed change will not delete currently listed Reactor Core SLs, DNBR, and Axial Shape Index related topical reports from the list of referenced analytical methods, since these TRs have elements of methodology that are not addressed in WTDP TR WCAP-18240-P-A (e.g., CENPD-183-A addresses the modeling of reactor core flow coastdown in transient analyses).

The NRC finds that the proposed addition of the TR WCAP-18240-P-A to TS 5.6.5.b is acceptable because its adoption for developing cycle-specific COLR is justified.

3.4 TS License Conditions 3.4.1 Radial Fall-Off Penalty The licensee analyzed the fuel performance for Westinghouse CE16NGF fuel using the FATES3B code to generate predictions of fuel performance parameters. The analysis applies bounding power histories, axial power shapes and peaking factor reduction with burnup with the intention of producing conservative results. However, the effects of high burnup thermal conductivity degradation (TCD) are not explicitly accounted for in the FATES3B code.

To address the non-conservative neglect of TCD in current licensing basis analysis methods, the LAR supporting implementation of Westinghouse NGF (Reference 16) established a license condition imposing a radial power fall off (RFO) curve penalty to accommodate the anticipated impacts of TCD on the predictions of FATES3B at high burnup for Westinghouse CE16NGF fuel. The licensee stated that the TCD-related RFO penalty specified in the license condition shall continue to be applied to accommodate the effects of TCD on FATES3B predictions for Westinghouse CE16NGF fuel. The method of applying the RFO curve penalty and its amount to conservatively accommodate the effects of TCD on FATES3B code predictions is described in detail in section 3.5.4, Fuel Performance, of the NRC safety evaluation associated with Palo Verde, Units 1, 2, and 3, Amendment Nos. 205, 205, and 205, respectively (Reference 16), for implementation of Westinghouse CE16NGF fuel.

Adoption of the WTDP methodology does not require changes to the RFO penalty license condition because WTDP involves subchannel T-H analyses with the VIPRE-W or VIPRE-01 computer codes, not fuel performance analyses with the FATES3B code. WTDP involves different, burnup-dependent design criteria specified in NUREG-0800, section 4.2 (Reference 7), that are analyzed with different computer codes. WTDP does not involve use of the FATES3B computer code, and as such, use or non-use of WTDP does not impact application of the license condition RFO penalty.

3.4.2 Reload Batch The NRC staff review of the LAR supporting implementation of Framatome CE16HTP fuel identified a question concerning whether approval of the LAR would permit operation with mixed batches of fresh fuel. The question related to whether the analytical methods for reactor core design and safety analysis that are listed in TS 5.6.5.b would be applied to core configurations involving mixed batches of fresh fuel and the potential for increased uncertainty in the predicted results.

To address this question, the LAR that supported implementation of Framatome High Thermal Performance Fuel (Reference 14) established a license condition stating that prior to use of fresh fuel from multiple fuel vendors in a single reload batch, the licensee will obtain NRC approval of the methodology used to perform the associated reload safety analyses. Adoption of the WTDP methodology does not require changes to the mixed core license condition and therefore will remain as is.

3.5 Thermal-Hydraulic Analyses The licensee performed the following T-H analysis to demonstrate the acceptability of the WTDP methodology.

Demonstration Analysis for Westinghouse CE16NGF Fuel Demonstration Analysis for Framatome CE16HTP Fuel Demonstration Analysis for Loss-of-Flow from SAFDL Transient Setpoint Analyses 3.5.1 Demonstration Analysis for Westinghouse CE16NGF Fuel To implement the WTDP methodology in a Palo Verde fuel cycle containing a full core of CE16NGF fuel, the licensee performed analysis to demonstrate the generation of a cycle-specific CE16NGF DNBR SAFDL and associated DNBR PDF. This analysis is based on representative Palo Verde, Unit 1, Cycle 24 (U1C24) core parameters and the corresponding limiting assembly definition. In enclosure attachment 6 of the LAR, a table in section 4.1 provides the random sampling range for inlet temperature, system pressure, vessel flow and axial shape index.

By using the NRC-approved VIPRE-W code, the licensee made numerous runs and generated a set of DNBR data and statistically determined the DNBR SAFDL and its associated DNBR PDF. Consistent with the response to RAI-WTDP-04 reported in TR WCAP-18240-P-A, the licensee excluded selected VIPRE-W cases from the DNBR SAFDL calculation. Even with the exclusion of the VIPRE-W cases, the sample size to determine DNBR distribution is greater than the minimum number of samples from the operating space specified in limitation and condition 3 of WTDP safety evaluation.

In compliance with Limitation and Condition (L&C) 2, and consistent with the LAR on transition to the Westinghouse CE16NGF fuel (Reference 17; attachment 8, section 5.2.3), the licensee addressed the parameters and code uncertainties. The parameter uncertainties are basically the same as in table 1-1 of the currently approved T-H CEN-356(V) -P-A Revision 1 (Reference 5) methodology.

In compliance with WTDP TR L&C 1, the licensee used the ABB-NV and WSSV CHF correlations in the analysis with the NRC-approved VIPRE-W T-H subchannel code. The DNBR distribution is generated by combining system parameter and code uncertainties with WSSV or ABB-NV CHF correlation statistics. The resultant DNBR SAFDL, which may be further adjusted to account for rod bow DNBR penalty, is taken to be the DNBR SAFDL for the entire CE16NGF bundle.

The licensee stated that the Westinghouse fuel rod bow methodology described in CENPD-225-P-A, Technical Evaluation of the Combustion Engineering Fuel and Poison Rod Bowing (Reference 18), is not changed by the WTDP methodology. Palo Verde methodology accounts for the NGF rod bow penalty using a methodology which was approved for use with CE16NGF fuel at Palo Verde, Units 1, 2, and 3 in the LAR supporting implementation of Westinghouse NGF (Reference 16 and Reference 17, attachment 8, section 5.6).

In enclosure attachment 6 to the LAR, section 4.1, the table titled Components Combined in the DNBR PDF for PVNGS Westinghouse CE16NGF Fuel, the licensee listed the mean values of the components combined in the DNBR PDF calculation and their corresponding 95 percent confidence standard deviation.

The NRC staff finds the T-H demonstration analysis for the Westinghouse CE16NGF fuel acceptable because using the NRC-approved T-H VIPRE-W methodology and satisfying the required L&Cs on the use of this methodology, the licensee calculated DNBR results based on ABB-NV and WSSV CHF correlations predicted/measured (P/M) and preserved the NRC-approved ABB-NV CHF correlation maximum limit of 1.13, as required by NRC Information Notice (IN)-2014-01 (Reference 19).

Since CE16NGF fuel models both WSSV and ABB-NV CHFs, two DNBR SAFDLs and their associated DNBR PDFs are generated. The licensee stated that it may use an NRC-accepted statistical approach to determine a single DNBR SAFDL and associated DNBR PDF to bound both DNBR SAFDLs and PDFs from the two CHFs. The single DNBR SAFDL and associated DNBR PDF would bound both WSSV and ABB-NV CHF correlations.

The licensee stated that it may choose to validate the results of this demonstration analysis for a given future cycle of operation or to generate a new cycle-specific DNBR SAFDL and associated DNBR PDF.

3.5.2 Demonstration Analysis for Framatome CE16HTP Fuel The licensee performed a demonstration analysis to implement the WTDP method and process in a Palo Verde fuel cycle containing a full core of Framatome CE16HTP fuel. The purpose of the analysis is to generate CE16HTP fuel cycle-specific DNBR SAFDL and associated DNBR PDF.

The methodology and process used in the CE16HTP demonstration analysis are the same as those used in the CE16NGF demonstration analysis presented in section 3.3.1 above. The licensee took the following exceptions associated with input data:

The CE16HTP analysis is based on a representative core physics and corresponding limiting assembly definition established in demonstration cycle N2 in Unit 2 Cycle 22 (U2C22) LAR supporting implementation of Framatome CE16HTP fuel (Reference 20).

In compliance with the WTDP TR safety evaluation L&C 2, the parameters and code uncertainties are as addressed in the LAR dated July 6, 2018, supporting implementation of Framatome CE16HTP Fuel. (Reference 20; attachment 10, section 5.6).

The following are the key conservative features of the licensees analysis:

The DNBR distribution used to determine the DNBR SAFDLs and associated DNBR PDF is based on a sample size that is greater than the minimum number of samples from the operating space specified in L&C 3.

In compliance with WTDP L&C 1, the demonstration analysis uses the BHTP CHF correlation with the VIPRE-W T-H subchannel code as approved for use with CE16HTP fuel at Palo Verde, Units 1, 2, and 3 in the LAR dated July 6, 2018 (Reference 20) supporting implementation of CE16HTP fuel.

The DNBR distribution is generated by combining system parameter and code uncertainties with BHTP CHF correlation statistics.

The DNBR SAFDL, after accounting the rod bow DNBR penalty, is taken to be the DNBR SAFDL for the entire CE16HTP bundle.

Consistent with section 3.3.3 in the NRCs safety evaluation for the Palo Verde license amendments implementing CE16HTP fuel (Reference 20), the VIPRE-W model conservatively applies the CE-1 CHF correlation for DNB analysis in the region below the first HTP' spacer grid.

In enclosure attachment 6 to the LAR, section 4.2, the table titled Components Combined in the DNBR PDF for PVNGS Framatome CE16HTP Fuel, lists the mean and 1 uncertainties (standard deviations) of system parameters, BHTP CHF correlation, and VIPRE-W T-H code uncertainties at the 95 percent confidence level.

The NRC staff finds the T-H demonstration analysis for the Framatome CE16HTP fuel acceptable because using the NRC-approved T-H VIPRE-W methodology and satisfying the required L&Cs on the use of this methodology, the licensee calculated DNBR results based on BHTP CHF correlation (P/M) and preserved the NRC-approved maximum limit of 1.110, as required by NRC IN-2014-01 (Reference 19).

In enclosure attachment 6 to the LAR, the last table in section 4.2 presents the DNBR PDF for CE16HTP fuel based on BHTP CHF correlation for the operating conditions..

The licensee stated that it may choose to validate the results of this demonstration analysis for a given future cycle of operation or to generate a new cycle-specific DNBR SAFDL and associated DNBR PDF.

3.6 Transient Analyses (Loss-of-Flow from SAFDL) - Demonstration Analysis The licensee stated that the loss-of-flow (LOF) from SAFDL is a composite event that has been evaluated to bound all other infrequent events, including AOOs in combination with a single active failure, with respect to the degradation in the DNBR. For the demonstration analysis, the licensee considered normal coastdown of one or more reactor coolant pumps and the LOF transient event resulting from shaft seizure of one pump. The licensee used VIPRE-W code for both CE16NGF and CE16HTP fuels.

3.6.1 DNBR Versus Fr Calculation The LOF from SAFDL event demonstration analysis results provide CE16NGF and CE16HTP DNBR versus time profiles, including their minimum DNBRs (mDNBRs) obtained during the event. The operating conditions at the time of mDNBR are entered into VIPRE to calculate mDNBR versus Fr (i.e., radial peaking factor) pairs to be used in calculating fuel failure.

Section 5.2 of the LAR enclosure attachment 6 shows two graph results obtained from the demonstration analysis using VIPRE. The graphs show mDNBR approached a value of 0.78 at approximately 1.85 seconds for Westinghouse CE16NGF fuel, and 0.84 at approximately 2.10 seconds for Framatome CE16HTP fuel, following the loss of power, which is well below the DNBR SAFDL of 1.20 modeled in the demonstration analysis. These figures depict a representative hot channel DNBR transient for this limiting event. The licensee stated that DNB propagation is not predicted to occur because within approximately 3.95 seconds, due to control element assembly insertion, local and average core heat flux has decreased enough such that no pins experiencing DNB would remain.

3.6.2 Fuel Failure Analysis For the LOF from SAFDL fuel failure analysis (i.e., calculating percent fuel failure), the licensee used the statistical method described in the NRC-approved TR CENPD-183-A (Reference 15).

In the supplemental letter dated July 3, 2025 (Reference 2), the licensee stated that prior to 2003, conservative assumptions listed in table 1 of TR CENPD-183-A were used, and starting in 2003, the licensee modified the licensing basis analysis based on assumptions generally consistent with, but more conservative than those listed in table 1 of TR CENPD-183-A. The modified methodology supported license amendments related to steam generator replacements and power uprate for all three Palo Verde units, as documented in NRC safety evaluations for Palo Verde, Unit 2, Amendment No. 149 dated September 29, 2003 (Reference 21), and Amendment Nos. 157, 157, and 157 for Units 1, 2, and 3, respectively dated November 16, 2005 (Reference 22). Both safety evaluations are referenced in TS 5.6.5.b clarifying how CENPD-183-A is utilized in the Palo Verde licensing basis LOF event analyses.

This demonstration analysis combines the mDNBR versus Fr data pairs with cycle-specific fuel data from Unit 1 Cycle 24 (U1C24) for CE16NGF fuel and from Unit 2 Cycle 25 (U2C25) for CE16HTP fuel. In the LAR enclosure attachment 6, the tables in section 5.3 provide the results of these calculations, which are compared with the results for the actual cycle-specific (non-WTDP) calculations. Using the WTDP methodology, the calculated product of FF and Fr (i.e.,

FF x Fr) shown in these tables is 15.23 percent for CE16NGF fuel and 5.71 percent for the CE16HTP fuel. As stated in the Palo Verde UFSAR, appendix 15E, the offsite radiological dose consequences for the LOF event are limited to a small fraction, or 10 percent, of 10 CFR Part 100 guideline values, which is met if FF x Fr is less than or equal to () 18.5 percent.

The NRC staff finds that the LOF demonstration analysis for CE16NGF and CE16HTP fuel failure using the WTDP methodology is acceptable because the product FF x Fr for both fuels is bounded by the UFSAR value of 18.5 percent.

3.7 Setpoint Analyses Section 3.1 of this safety evaluation provides an overview of the setpoint analysis methodology.

The currently used general application of the COLSS and CPCS DNBR OUA process based on the MSCU methodology was provided to the NRC in response to SNPB RAI-3 associated with

the LAR supporting implementation of Framatome CE16HTP fuel (Reference 23). In enclosure attachment 6, section 6.2, a figure shows the WTDP methodology impacts on the current process because of the choice of the T-H model shown by the boxes with bold and italicized text in the figure.

TR WCAP-18240-P-A safety evaluation, L&C 4 requires the use of approved VIPRE code in lieu of CETOP-D to be consistent with the CE-NSSS setpoint methodology as defined in TR WCAP-16500-P-A, Supplement 1, Revision 1 (Reference 6). The licensee stated that this requirement is satisfied because the VIPRE code is only an alternate means of calculating DNBR and does not alter the methodology or processes set forth in TR WCAP-16500-P-A, Supplement 1, Revision 1.

The NRC staff finds that the licensees discretionary use of the current CETOP-D or VIPRE codes in the setpoint process acceptable because the use of VIPRE code in the setpoint analysis is technically justified.

3.8 Limitations and Conditions in safety evaluation for WCAP-18240-P-A 3.8.1 Limitation and Condition 1 - Sub-Channel Code and CHF Correlation Combination In application to a given plant, WTDP shall be used with a subchannel code and CHF correlation combination that has been approved for the plant type and the fuel type in use at the plant.

Licensees Compliance Statement and NRC Staff Evaluation The licensee stated:

The CE16NGF fuel evaluation documented in this WTDP technical analysis demonstrates a successful application of the WTDP methodology by using the ABB-NV and WSSV CHF correlations with the VIPRE T-H subchannel code as approved for use with CE16NGF fuel at PVNGS Units 1, 2, and 3 in the LAR supporting implementation of Westinghouse Next Generation Fuel (Reference 7.15 [Reference 16]).

The CE16HTP fuel evaluation documented in this WTDP technical analysis demonstrates a successful application of the WTDP methodology by using the BHTP CHF correlation with the VIPRE T-H subchannel code as approved for use with CE16HTP fuel at PVNGS Units 1, 2, and 3 in the LAR supporting implementation of Framatome High Thermal Performance Fuel (Reference 7.16

[Reference 14]).

Therefore, WTDP WCAP-18240-P-A Safety Evaluation limitation and condition 1 is met.

The NRC staff finds that L&C 1 is satisfied because the licensee used WTDP methodology which in turn uses NRC approved CHF correlations.

3.8.2 L&C 2 - Parameter Uncertainties used in DNBR Limit Calculation Parameter uncertainties used in the 95/95 DNBR limit calculation must be justified on a plant specific basis.

Licensees Compliance Statement and NRC Staff Evaluation The licensee stated:

The CE16NGF fuel evaluation documented in this WTDP technical analysis demonstrates a successful application of the WTDP methodology using the same parameter uncertainties as those utilized in the currently approved T-H MSCU methodology. These parameter uncertainties were approved for use with CE16NGF fuel at PVNGS Units 1, 2, and 3 in the LAR supporting implementation of Westinghouse Next Generation Fuel (Reference 7.15

[Reference 16]).

The CE16HTP fuel evaluation documented in this WTDP technical analysis demonstrates a successful application of the WTDP methodology using the same parameter uncertainties as those utilized in the currently approved T-H MSCU methodology. These parameter uncertainties were approved for use with CE16HTP fuel at PVNGS Units 1, 2, and 3 in the LAR supporting implementation of Framatome High Thermal Performance Fuel (Reference 7.16

[Reference 14]).

Therefore, WTDP WCAP-18240-P-A Safety Evaluation limitation and condition 2 is met.

The NRC staff finds that L&C 2 is satisfied because the licensee demonstrated an acceptable application of the WTDP methodology using NRC approved parameter uncertainties for use with CE16HTP fuel at Palo Verde, Units 1, 2, and 3.

3.8.3 L&C 3 - Quantity of Samples in the DNBR Distribution The DNBR distribution used to determine the statistical DNBR limit shall be based on a selected minimum [number of] samples from the operating space.

Licensees Compliance Statement and NRC Staff Evaluation The licensee stated:

The CE16NGF fuel evaluation documented in this WTDP technical analysis is based on a sample size that is greater than the minimum number of samples from the operating space specified in WTDP WCAP-18240-P-A Safety Evaluation limitation and condition 3.

The CE16HTP fuel evaluation documented in this WTDP technical analysis is based on a sample size that is greater than the minimum number of samples from the operating space specified in WTDP WCAP-18240-P-A Safety Evaluation limitation and condition 3.

Therefore, WTDP WCAP-18240-P-A Safety Evaluation limitation and condition 3 is met.

The NRC staff finds that L&C 3 is satisfied because in the licensees demonstration analyses for CE16NGF and CE16HTP fuel, the sample size of DNBR distribution used to determine the DNBR SAFDL and associated DNBR PDF is greater than the licensees selected minimum number of samples.

3.8.4 L&C 4 - Use of an Approved Subchannel Code in Lieu of CETOP-D The use of an approved subchannel code (e.g., VIPRE-W) in lieu of CETOP-D must be consistent with the CE-NSSS setpoint methodology as defined in WCAP-16500-P-A, Supplement 1, Application of CE Setpoint Methodology for CE 16x16 Next Generation Fuel, Revision 1 (Ref. 15 [Reference 6]).

Licensees Compliance Statement and NRC Staff Evaluation The licensee stated:

The use of approved T-H subchannel code VIPRE (e.g., VIPRE-W or VIPRE-01) in lieu of CETOP-D is consistent with the CE-NSSS setpoint methodology as defined in WCAP-16500-P-A, Supplement 1, Revision 1 (Reference 7.2

[Reference 6]). This requirement is satisfied since the VIPRE code is only an alternate means of calculating DNBR and does not otherwise alter the methodology or processes set forth in WCAP-16500-P-A, Supplement 1, Revision 1.

Therefore, WTDP WCAP-18240-P-A Safety Evaluation limitation and condition 4 is met.

The NRC staff finds that L&C 4 is satisfied because in the demonstration analyses for CE16NGF and CE16HTP fuel, the licensee used VIPRE code in lieu of the CETOP-D code.

3.9 Technical Conclusions The NRC staff finds the addition of the WTDP methodology in Palo Verde TS 5.6.5.b acceptable because its adoption of developing a cycle-specific COLR is justified based on the reasons listed below.

The licensees T-H demonstration analysis based on Westinghouse CE16NGF and Framatome CE16HTP fuels provided acceptable DNBR results.

The licensees DNBR and the fuel failure demonstration analyses results for the most limiting transient (i.e., LOF event meets the acceptance criteria).

The licensees demonstration analysis satisfied all L&Cs listed in the safety evaluation on the use of WTDP.

The license condition required by Amendment Nos. 205, 205, and 205 documented in Palo Verde Units 1, 2, and 3, TS, appendix D, is not impacted by the proposed change.

The licensee justified the use of WTDP VIPRE code as an option, in lieu of the currently used two NRC-approved T-H codes, in the setpoint analysis process in the plant CPCS and COLSS.

The impact on the TSs on adoption of WTDP methodology is as follows:

TS SL 2.1.1.1 is not affected because the LSSS DNBR value 1.34 currently installed in plant CPCS remains as is.

TS 3.2.4 is not affected because DNB correlations already in the COLSS are not changed.

TS 4.2.1 is not affected as none of its contents requires any change.

TS 3.3.1 is not affected because there is no change in the fuel design limits.

3.10 Regulatory Conclusions GDC 10 of Appendix A to 10 CFR Part 50 is satisfied because the WTDP describes a method for determining a DNBR SAFDL below which fuel failure may occur. Compliance with GDC 10 provides assurance that the integrity of the fuel and cladding will be maintained, thus preventing the potential for release of fission products during normal operation or AOOs.

WTDP requires control room and offsite dose due to damaged fuel rods be evaluated which are subject to GDC 19 and 10 CFR Part 100 requirements. Since the use of WTDP is acceptable for the most limiting non-LOCA, the control room and offsite dose requirements will be satisfied using the WTDP.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Arizona State official was notified of the proposed issuance of the amendment on September 10, 2025. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to 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 May 13, 2025 (90 FR 20326), 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.

7.0 REFERENCES

1.

Spina, J. L., Arizona Public Service Company, letter to U.S. Nuclear Regulatory Commission, Palo Verde Nuclear Generating Station Units 1, 2, and 3 Docket Nos.

STN 50-528, 50-529, and 50-530 Renewed Operating License Nos. NPF-41, NPF-51, and NPF-74 License Amendment Request to Adopt Westinghouse Thermal Design Procedure Topical Report WCAP18240-P-A, dated January 17, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25017A380).

2.

Spina, J. L., Arizona Public Service Company, letter to U.S. Nuclear Regulatory Commission, Palo Verde Nuclear Generating Station Units 1, 2, and 3 Docket Nos.

STN 50-528, 50-529, and 50-530 Renewed Operating License Nos. NPF-41, NPF-51, NPF-74 Response to Request for Additional Information Regarding License Amendment Request to Adopt Westinghouse Thermal Design Procedure Topical Report WCAP18240-P-A, dated July 3, 2025 (ML25184A397).

3.

Westinghouse Electric Company LLC, Westinghouse Thermal Design Procedure (WTDP), WCAP-18240-P-A, April 2020 (ML20104C042; not publicly available, proprietary information).

4.

Arizona Public Service Company, Enclosure 2-Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Revision 22 to Updated Final Safety Analysis Report, June 2023 (ML23181A166).

5.

Combustion Engineering, Modified Statistical Combination of Uncertainties, CEN356(V)-P-A, Revision 1, May 1988 (ML17304A447 and ML20153H401).

6.

Gresham, J. A., Westinghouse Electric Company LLC, letter to U.S. Nuclear Regulatory Commission, Submittal of the Approved Versions of WCAP-16500-P-A Supplement 1 Revision 1, Application of CE Setpoint Methodology for CE 16x16 Next Generation Fuel (NGF) (TAC No. ME3583) (Proprietary/Non-Proprietary), dated December 6, 2010 (Package ML103510160).

7.

U.S. Nuclear Regulatory Commission, Fuel System Design, Section 4.2, NUREG-0800, Revision 3, March 2007 (ML070740002).

8.

U.S. Nuclear Regulatory Commission, Thermal and Hydraulic Design, Section 4.4, NUREG-0800, Revision 2 (ML070550060).

9.

Sepp, H. A., Westinghouse Electric Company LLC, letter to U.S. Nuclear Regulatory Commission, Transmittal of Approved Versions of Topical Reports, WCAP-14565-P-A (Proprietary) VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis, dated October 25, 1999 (Package ML993160158).

10.

ABB Combustion Engineering Nuclear Fuel, Statistical Combination of Uncertainties Part 1; Combination of System Parameter Uncertainties in Thermal Margin Analyses for San Onofre Nuclear Units 2 and 3, CEN-283(S)-NP, Part 1, Revision 0, June 1984 (Package ML19214A142).

11.

ABB Combustion Engineering Nuclear Fuel, Statistical Combination of Uncertainties Part 2; Uncertainty Analysis of Limiting Safety System Settings San Onofre Nuclear Generating Station Units 2 and 3, CEN-283(S)-P, Part 2, Revision 0, October 1984 (Package ML19214A142).

12.

Combustion Engineering, Inc., CETOP-D Code Structure and Modeling Methods for San Onofre Nuclear Generating Station Units 2 and 3, CEN-160(S)-P, Revision 1-P, September 1981 (ML15356A49; not publicly available, proprietary information).

13.

Combustion Engineering, Inc., TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core, CEN-161-P-A, April 1986 (ML090490589; not publicly available, proprietary information).

14.

Lingam, S. P., U.S. Nuclear Regulatory Commission, letter to M. Lacal, Arizona Public Service Company, Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Issuance of Amendment Nos. 212, 212, and 212 to Revise Technical Specifications to Support the Implementation of Framatome High Thermal Performance Fuel (EPID L-2018-LLA-0194), dated March 4, 2020 (ML20031C947).

15.

Combustion Engineering, Inc., C-E Methods for Loss of Flow Analysis, CENPD-183-A, June 1984 (ML20097K029).

16.

Lingam, S. P., U.S. Nuclear Regulatory Commission, letter to R. Bement, Arizona Public Service Company, Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Issuance of Amendments to Revise Technical Specifications to Support the Implementation of Next Generation Fuel (CAC Nos. MF8076, MF8077, and MF8078; EPID L-2016-LLA-0005), dated January 23, 2018 (ML17319A107).

17.

Lacal, M., Arizona Public Service Company, letter to U.S. Nuclear Regulatory Commission, License Amendment Request and Exemption Request to Support the implementation of Next Generation Fuel, dated July 1, 2016 (Package ML16188A336).

18.

Combustion Engineering Corporation, Technical Evaluation of the Combustion Engineering Fuel and Poison Rod Bowing, CENPD-225-P-A, Revision 0 and Supplements 1, 2-P and 3-P, June 1983 (ML16161A093; not publicly available, proprietary information).

19.

U.S. Nuclear Regulatory Commission, Fuel Safety Limit Calculation Inputs were Inconsistent with U.S. Nuclear Regulatory Commission-Approved Correlation Limit Values, Information Notice 2014-01, dated February 21, 2014 (ML13325A966).

20.

Lacal, M., Arizona Public Service Company, letter to U.S. Nuclear Regulatory Commission, Palo Verde Nuclear Generating Station Units 1, 2, and 3 Docket Nos.

STN 50-528, [50]-529, and 50-530 License Amendment Request and Exemption Request to Support the Implementation of Framatome High Thermal Performance Fuel, dated July 6, 2018 (ML18187A417).

21.

Pham, Bo M., U.S. Nuclear Regulatory Commission, letter to G. R. Overbeck, Arizona Public Service Company, Palo Verde Nuclear Generating Station, Unit 2 (PVNGS-2) -

Issuance of Amendment on Replacement of Steam Generators and Uprated Power Operations (TAC No. MB3696), dated September 29, 2003 (ML032720538).

22.

Fields, M. B., U.S. Nuclear Regulatory Commission, letter to J. M. Levine, Arizona Public Service Company, Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Issuance of Amendments Re: Replacement of Steam Generators and Uprated Power Operations and Associated Administrative Changes (TAC Nos. MC3777, MC3778, and MC3779),

dated November 16, 2005 (Package ML053130286).

23.

Lacal, M., Arizona Public Service Company, letter to U.S. Nuclear Regulatory Commission, Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3 Docket Nos. STN 50-528, 50-529, and 50-530 Renewed Operating License Nos. NPF-41, NPF-51, and NPF-74, Response to NRC Request for Additional Information Regarding License Amendment and Exemptions Requests Related to the Implementation of Framatome CE16HTP Fuel, dated October 4, 2019 (ML19277J457).

8.0 ABBREVIATIONS AND ACRONYMS ABBREVIATIONS AND ACRONYMS DESCRIPTION ABB-NV Westinghouse (ABB) Non-Vane Critical Heat Flux Correlation ADAMS Agencywide Documents Access and Management System ANSI American National Standards Institute AOO anticipated operational occurrence APS Arizona Public Service Company BHTP Designation for a Framatome Critical Heat Flux Correlation CE Combustion Engineering CETOP Combustion Engineering Thermal On-Line Program (code)

CFR Code of Federal Regulations CHF critical heat flux COLR core operating limits report CPC core protection calculator CPCS core protection calculator system COLSS core operating limit supervisory system DBLFPB design basis limits for fission product barriers DNB departure from nucleate boiling DNBR departure from nucleate boiling ratio FF failed fuel

ABBREVIATIONS AND ACRONYMS DESCRIPTION Fr radial peaking factor GDC General Design Criterion L&C limitation and condition LAR license amendment request LCO limiting condition for operation LOCA loss-of-coolant accident LOF Loss-of-flow LSSS limiting safety system setting mDNBR minimum departure from nucleate boiling ratio MSCU modified statistical combination of uncertainties NGF next generation fuel NRC U.S. Nuclear Regulatory Commission NSSS nuclear steam supply system OUA overall uncertainty analysis PDF probability density function P/M predicted/measured PVNGS Palo Verde Nuclear Generating Station PWR pressurized water reactor RAI request for additional information RFO radial power fall off RPS reactor protective system SONGS San Onofre Nuclear Generating Station SAFDL specified acceptable fuel design limit SL safety limit TCD thermal conductive degradation T-H thermal-hydraulic TR topical report TS technical specification UFSAR Updated Final Safety Analysis Report WSSV Westinghouse side supported Vane Critical Heat Flux Correlation WTDP Westinghouse Thermal Design Procedure Principal Contributor: A. Sallman, NRR Date: September 22, 2025

ML25254A064

• via eConcurrence OFFICE NRR/DORL/LPL4/PM*

NRR/DORL/LPL4/LA*

NRR/DSS/SNSB/BC*

NRR/DORL/LPL4/BC*

NAME WOrders PBlechman NDiFrancesco TNakanishi DATE 9/11/2025 9/18/2025 9/19/2025 9/22/2025 OFFICE NRR/DORL/LPL4/PM NAME WOrders DATE 9/22/2025