ML25126A076
| ML25126A076 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 05/30/2025 |
| From: | William Orders Plant Licensing Branch IV |
| To: | Heflin A Arizona Public Service Co |
| Orders, William | |
| References | |
| EPID L-2024-LLA-0116 | |
| Download: ML25126A076 (1) | |
Text
May 30, 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. 225, 225, AND 225 REGARDING REVISION TO TECHNICAL SPECIFICATIONS 3.5.1 AND 3.5.2 SAFETY INJECTION TANK PRESSURE BANDS AND USE OF GOTHIC CODE (EPID L-2024-LLA-0116)
Dear Mr. Heflin:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 225 to Renewed Facility Operating License No. NPF-41, Amendment No. 225 to Renewed Facility Operating License No. NPF-51, and Amendment No. 225 to Renewed Facility Operating License No. NPF-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 August 28, 2024,as supplemented by letter dated March 28, 2025.
The amendments revise Palo Verde TS section 3.5.1, Safety Injection Tanks (SITs) -
Operating and TS section 3.5.2, Safety Injection Tanks (SITs) - Shutdown and their bases.
Specifically, the TS changes revise Surveillance Requirements 3.5.1.3 and 3.5.2.3 to increase the upper limit of their SIT pressure bands, and to list their pressure requirements in units of pounds per square inch absolute as reflected in the Palo Verde safety analyses, with no instrument uncertainties included, instead of the SIT instrument units of pounds per square inch gauge with instrument uncertainties included.
Palo Verde currently uses the Bechtel Containment Pressure and Temperature Transient Analysis (COPATTA) code as part of the methodology for the calculation of the containment pressure and temperature response to various postulated pipe breaks. The proposed changes also include use of the industry standard Generation of Thermal Hydraulic Information for Containments (GOTHIC) code as a replacement to the COPATTA code as part of the methodology to perform calculations of the containment pressure and temperature response to various postulated pipe breaks, including determination of the response to the increase in the upper limit of the SIT pressure bands.
A. Heflin 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. 225 to NPF-41
- 2. Amendment No. 225 to NPF-51
- 3. Amendment No. 225 to NPF-74
DOCKET NO. STN 50-528 PALO VERDE NUCLEAR GENERATING STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 225 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 August 28, 2024, as supplemented by letter dated March 28, 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. 225, 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: May 30, 2025 TONY NAKANISHI Digitally signed by TONY NAKANISHI Date: 2025.05.30 15:29:31 -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. 225 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 August 28, 2024, as supplemented by letter dated March 28, 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. 225, 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: May 30, 2025 TONY NAKANISHI Digitally signed by TONY NAKANISHI Date: 2025.05.30 15:29:57 -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. 225 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 August 28, 2024, as supplemented by letter dated March 28, 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. 225, 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: May 30, 2025 TONY NAKANISHI Digitally signed by TONY NAKANISHI Date: 2025.05.30 15:30:19 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NOS. 225, 225, AND 225 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 License 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 3.5.1-2 3.5.1-2 3.5.2-2 3.5.2-2 Renewed Facility Operating License No. NPF-41 Amendment No. 225 (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. 225, 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. 225 (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. 225, 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. 225 (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. 225, 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.
SITs-Operating 3.5.1 PALO VERDE UNITS 1,2,3 3.5.1-2 AMENDMENT NO. 224, 225 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D.
Required Action and associated Completion Time of Condition A, B, or C not met.
D.1 Be in MODE 3.
AND 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> D.2 Reduce pressurizer pressure to
< 1837 psia.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each SIT isolation valve is fully open.
In accordance with the Surveillance Frequency Control Program SR 3.5.1.2 Verify borated water volume in each SIT is 1750 cubic feet and 1950 cubic feet.
In accordance with the Surveillance Frequency Control Program SR 3.5.1.3 Verify nitrogen cover pressure in each SIT is 602 psia and 675 psia.
In accordance with the Surveillance Frequency Control Program (continued)
SITs - Shutdown 3.5.2 PALO VERDE UNITS 1,2,3 3.5.2-2 AMENDMENT NO. 224, 225 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify each required SIT isolation valve is fully open when pressurizer pressure is 430 psia.
In accordance with the Surveillance Frequency Control Program SR 3.5.2.2 Verify borated water volume in each required SIT is:
a.
For four OPERABLE SITs, > 908 cubic feet and < 2000 cubic feet.
OR b.
For three OPERABLE SITs, > 1361 cubic feet and < 2000 cubic feet.
In accordance with the Surveillance Frequency Control Program SR 3.5.2.3 Verify nitrogen cover pressure in each required SIT is 250 psia and 675 psia.
In accordance with the Surveillance Frequency Control Program (continued)
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 225, 225, AND 225 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: 50-528, 50-529, AND 50-530
1.0 INTRODUCTION
By application dated August 28, 2024 (Reference 1), as supplemented by letter dated March 28, 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, Units 1, 2, and 3 (Palo Verde, PVNGS).
The proposed amendments would modify Palo Verde TS section 3.5.1, Safety Injection Tank (SITs) - Operating, and TS section 3.5.2, Safety Injection Tanks (SITs) - Shutdown.
Specifically, the proposed TS changes would revise Surveillance Requirements (SRs) 3.5.1.3 and 3.5.2.3 by increasing the upper limit of SIT pressure bands, and to list their pressure requirements in units of pounds per square inch absolute (psia) as reflected in the Palo Verde safety analyses, with no instrument uncertainties included, instead of the SIT instrument units of pounds per square inch gauge (psig) with instrument uncertainties included.
The proposed changes also include the use of Generation of Thermal Hydraulic Information for Containments (GOTHIC) code as part of the methodology to perform calculations of the containment pressure and temperature response to the postulated pipe breaks. The currently used code for containment response analysis is the Bechtel Containment Pressure and Temperature Transient Analysis (COPATTA) code.
The supplemental letter dated March 28, 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 January 14, 2025 (90 FR 3256).
2.0 REGULATORY EVALUATION
2.1
System Description
The Palo Verde nuclear steam supply system design features four SITs partially filled with borated water pressurized by nitrogen gas. Each SIT is connected to its associated reactor coolant cold leg by a separate line containing check valves that isolate the SIT from the reactor coolant system (RCS) during normal operation. They are passive components, since no operator or control action is required for them to perform their function. Internal SIT pressure is sufficient to discharge the contents to the RCS, if RCS pressure decreases below the SIT pressure. Each SIT is capable of being isolated from the RCS by a motor operated isolation valve and two check valves in series. The motor operated isolation valves are normally open, with power removed from the valve motor to prevent inadvertent closure prior to or during an accident.
The functions of the four SITs are to supply water to the reactor vessel during the blowdown phase of a loss-of-coolant accident (LOCA), to provide inventory to help accomplish the refill phase that follows thereafter, and to provide RCS makeup for a small break LOCA. The minimum volume requirement for the SITs ensures that three SITs can provide adequate inventory to reflood the core and downcomer following a LOCA. The downcomer then remains flooded until the high pressure safety injection (SI) and the low pressure SI systems start to deliver flow. The maximum volume limit is based on maintaining an adequate gas volume to ensure proper injection and the ability of the SITs to fully discharge, as well as limiting the maximum amount of boron inventory in the SITs.
2.2 TS Changes 2.2.1. TS 3.5.1 - Safety Injection Tanks (SITs) - Operating TS 3.5.1 addresses a limiting condition for operation (LCO) and related SRs for the SITs during operating conditions. TS 3.5.1 is applicable in MODES 1 and 2 and MODES 3 and 4 with pressurizer pressure equal to or greater than () 1837 psia.
The current SR 3.5.1.3 requires verification that nitrogen cover pressure in each SIT is:
600 psig and [equal to or less than] 625 psig.
The proposed SR 3.5.1.3 requires verification that nitrogen cover pressure in each SIT is:
602 psia and 675 psia.
The proposed change to SR 3.5.1.3 is an increase in the upper limit of its SIT pressure band.
The revised upper pressure limit of 675 psia remains within the SIT design pressure of 700 psig (714.2 psia). This change also lists the pressure requirements in units of psia as reflected in the Palo Verde safety analyses, with no instrument uncertainties included, instead of the SIT instrument units of psig with instrument uncertainties included. The licensee states that accounting for instrument uncertainties will continue as part of surveillance procedures.
2.2.2 TS 3.5.2 - Safety Injection Tanks (SITs) - Shutdown TS 3.5.2 addresses an LCO and related SRs for the SITs during shutdown conditions. TS 3.5.2 is applicable in MODES 3 and 4 with pressurizer pressure less than 1837 psia.
The current SR 3.5.2.3 requires verification that nitrogen cover pressure in each SIT is:
260 psig and 625 psig.
The proposed SR 3.5.2.3 requires verification that nitrogen cover pressure in each SIT is:
250 psia and 675 psia.
The proposed change to SR 3.5.2.3 is an increase in the upper limit of its SIT pressure band to provide consistency with the change to SR 3.5.1.3. The revised upper pressure limit of 675 psia remains within the SIT design pressure of 700 psig (714.2 psia). This change also lists the pressure requirements in units of psia as reflected in the Palo Verde safety analyses, with no instrument uncertainties included, instead of the SIT instrument units of psig with instrument uncertainties included. The licensee states that accounting for instrument uncertainties will continue as part of surveillance procedures.
2.3 Reason for Proposed Change The licensee states that at the beginning of an operating cycle, fluctuations in the containment pressure and temperature may increase SIT pressure leading to cycling of the SIT vent valves as they open and close to maintain the SIT pressure within its operating band. If an SIT vent valve does not fully close during this cycling, then the plant is challenged with restoring operability of the SIT vent valve within the completion times required by TS 3.5.1 and TS 3.5.2.
To reduce the challenges to the SIT vent valves, the proposed TS change increases the upper limit of the SIT pressure band specified in SR 3.5.1.3 and increases the upper limit of the SIT pressure band specified in SR 3.5.2.3.
2.4 Regulations The NRC staff considered the following regulations in its review of the proposed changes:
The regulation in Title 10 of the Code of Federal Regulations (10 CFR) 50.36(c)(3),
Surveillance Requirements, states:
Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.
The regulations in 10 CFR 50.46(a)(1)(i) require an acceptable emergency core cooling system (ECCS) evaluation model that realistically describes the behavior of the reactor during LOCAs.
The regulations in 10 CFR 50.46(b) require that during a LOCA event, the following criteria are satisfied:
(1) Peak cladding temperature. The calculated maximum fuel element cladding temperature shall not exceed 2200 °F [degrees Fahrenheit].
(2) Maximum cladding oxidation. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.
(3) Maximum hydrogen generation. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
(4) Coolable geometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling.
(5) Long-term cooling. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.
The regulations in 10 CFR Part 50, Appendix K, ECCS Evaluation Models,Section II, Required Documentation, specify documentation requirements for the emergency core cooling performance evaluation models specified in 10 CFR 50.46(a)(1)(i).
The following General Design Criteria (GDC) in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, are applicable:
GDC 16, Containment design, as it relates to providing a reactor containment and associated systems to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.
GDC 35, Emergency core cooling, as it relates to demonstrating that the ECCS would provide abundant emergency core cooling to satisfy the ECCS safety function of transferring heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling would be prevented, and (2) clad metal-water reaction would be limited to negligible amounts.
GDC 50, Containment design basis, as it relates to designing the reactor containment structure, including access openings, penetrations, and the containment heat removal system so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any LOCA.
2.5 Precedents Multiple precedents have been established for including use of the GOTHIC code as part of the methodology to perform calculations of the containment pressure and temperature response to various postulated pipe breaks. These precedents include submittals by Entergy Operations, Inc. for Waterford Steam Electric Station, Unit 3; Southern Nuclear Operating Company for Joseph M. Farley Nuclear Plants, Units 1 and 2; and Luminant Generation Company, LLC for Comanche Peak Steam Electric Station, Unit Nos. 1 and 2.
3.0 TECHNICAL EVALUATION
As described in the Palo Verde Updated Final Safety Analysis Report (UFSAR) (Reference 3),
the currently used methodology for the containment response analysis for various postulated pipe breaks is the Bechtel COPATTA code. The proposed changes to the Palo Verde licensing basis include use of the GOTHIC code as part of the methodology to perform calculations of the containment pressure and temperature response, which includes the determination of the response due to increase in the upper limit of the SR 3.5.1.3 SIT pressure band.
3.1 COPATTA Code The COPATTA code was developed by Bechtel Power Corporation (Bechtel) for LOCA containment response analysis. It was derived from the NRCs original CONTEMPT computer code used by NRC to perform independent containment analysis. As stated in the code topical report BN-TOP-3, Performance and Sizing of Dry Pressure Containments, Revision 1 (Reference 4), the COPATTA model predicts both the pressure and temperature within the containment regions and the temperatures in the containment structures. The licensee currently uses the COPATTA code as part of the methodology for the calculation of the containment pressure and temperature response analysis to various postulated pipe breaks. Since Bechtel decided to discontinue the development, maintenance, and licensing of COPATTA, the licensee decided to use the GOTHIC code for this analysis.
3.2 GOTHIC Code The GOTHIC code is a general-purpose thermal-hydraulics code used for the design, licensing, safety, and operating analysis of nuclear power plant containments, confinement buildings, system components, and piping. Applications of GOTHIC include evaluation of containment response and containment subcompartment response to the full spectrum of high energy line breaks within the design basis envelope, and a wide variety of systems evaluations involving multiphase flow and heat transfer, gas mixing, and other thermal hydraulic behavior. The GOTHIC Technical Manual, Thermal Hydraulic Analysis Package Technical Manual, Version 8.4 (QA) (Reference 5), describes the equations and models and the numerical methods used to solve them. The GOTHIC Technical Manual also provides the necessary information for understanding GOTHIC, including assumptions and limitations. The GOTHIC User Manual, Thermal Hydraulic Analysis Package User Manual, Version 8.4 (QA)
(Reference 6), provides detailed description of inputs.
The NRC safety evaluations for Amendment No. 169 dated September 29, 2003, for the Kewaunee Nuclear Power Plant (Kewaunee) (Reference 7), and Amendment Nos. 171 and 161 dated August 12, 2005, for Prairie Island Nuclear Generating Plant, Units 1 and 2 (Prairie Island), respectively (Reference 8), state a condition for using the GOTHIC code. The condition is that the heat transfer mist diffusion layer model (MDLM) shall not be used for licensing calculations. In the licensees GOTHIC benchmark model documented in enclosure attachment 5 to the LAR dated August 28, 2024, and in the technical analysis of changes to the SIT pressure bands documented in enclosure attachment 6 to the LAR, the licensee addressed this condition by not using MDLM post-LOCA heat transfer model for containment pressure and temperature analysis. The GOTHIC model for containment response analysis uses the Tagami and Uchida condensing heat transfer correlations as described in the Palo Verde UFSAR, section 6.2.1.1.3.1, Containment Peak Pressure and Temperature Analysis (Reference 3).
3.3 Benchmarking of GOTHIC Code Results with COPATTA Code Results For benchmarking of the COPATTA code containment response model with the similar GOTHIC model, the licensee selected a double-ended discharge leg slot (DEDLS) break LOCA with maximum ECCS (i.e., SI) flow. The licensee selected this LOCA model for benchmarking because the licensee determined this pipe break resulted in highest peak containment pressure.
In the benchmarking process, differences between the GOTHIC and COPATTA results for containment pressure profile, containment vapor temperature profile, and sump temperature profile were noted. Based on its knowledge of the two codes, as described in subsections 3.3.1 and 3.3.2 below, the licensee performed sensitivity runs by changing the GOTHIC model to gain an understanding of the cause of differences. The licensees performance of the sensitivity runs was not intended to imply that the official GOTHIC run used for the analysis should make these modeling changes or additions.
The licensee used the following key elements of the GOTHIC code for the DEDLS benchmark model: (a) control volumes, (b) flow paths, (c) boundary conditions, (d) initial conditions, (e) thermal conductors, (f) control variables, (g) table functions, (h) heaters, (i) trips, (j) valves, and (k) heat exchangers. A detailed description of these elements is given in the GOTHIC users manual (Reference 6).
The licensee used GOTHIC version 8.4 for conversion of the COPATTA model of the DEDLS break LOCA with maximum SI flow. The licensee states that the analysis models and methods described in this benchmark evaluation are not intended to be restricted to a specific GOTHIC code version. GOTHIC is not an NRC-approved code but has been accepted for containment response analysis, and the future possible changes in GOTHIC are not controlled by APS. In the LAR supplement dated March 28, 2025 (Reference 2), the licensee states that according to APS internal procedures, a new or revised version of a code is required to go through a verification and validation process, including acceptance testing prior to being approved.
Acceptance testing includes ensuring the software adequately and correctly performs all intended functions; properly handles abnormal conditions and events, as well as credible failures; does not perform adverse unintended functions; and does not degrade the system by the computer program.
The licensee further states that future revisions to GOTHIC would enter the 10 CFR 50.59, Changes, tests, and experiments, evaluation process. Methodology departures from a method of evaluation described in the Palo Verde UFSAR used in establishing the design bases or in the safety analyses would require NRC approval prior to implementation. APS procedure AG 93DP-0LC07-01, section 6.2.3.6c states:
Changes to methods of evaluation included in the UFSAR are considered adverse and require evaluation under 10 CFR 50.59 or 72.48 if the changes are outside the constraints and limitations associated with use of the method. If the changes are within the constraints and limitations associated with the use of the method, then the change is not considered adverse and may be screened out.
To demonstrate the acceptable performance of the GOTHIC code, the licensee performed benchmarking of the GOTHIC code results with the COPATTA code results for the following parameters:
Containment Pressure Containment Vapor Temperature Containment Sump Temperature Containment Liner Temperature Transient Energy Profiles Transient Mass Profiles The licensee divided the COPATTA to GOTHIC results comparison in the following three categories:
Graphically identical graph results from GOTHIC that are visually indistinguishable from the COPATTA graphs when overlaid on the same plot.
Nearly identical graph results from GOTHIC that have slight variations for brief periods of time when compared to the equivalent COPATTA graphs. These variations can be considered inconsequential based on qualitative reasoning.
Valid graph results acknowledge that the conversion from COPATTA to GOTHIC is deemed appropriate for the specific set of results evaluated.
3.3.1 Containment Pressure Benchmarking Figures 3.0-1 and 3.0-2 in enclosure attachment 5 to the LAR, show the containment pressure profile benchmarking results without and with the recirculation fan modeled, respectively. The plot shows that GOTHIC predicts a lower peak containment pressure than COPATTA by approximately 0.83 pounds per square inch. The COPATTA results also show a pressure increase up to 95 seconds (the time of containment spray (CS) initiation) and then a momentary sharp reduction in the pressure before reaching the peak pressure. This is similar in trend to the temperature profile (figure 3.0-3 in enclosure attachment 5) where there is a significant increase in containment vapor space temperature compared to the GOTHIC results. In section 3.2.1, Containment Pressure Benchmarking, of enclosure attachment 5 to the LAR, the licensee states, in part:
results in GOTHIC predicting a lower peak containment pressure compared to COPATTA is due to more realistic heat and mass transfer dynamics. In contrast, COPATTA immediately sends introduced liquid to the sump unless the containment atmosphere becomes superheated, in which case small droplets are reintroduced for instant vaporization. This conservative approach leads to spikes in temperature above saturation in COPATTAs results, contrasting with GOTHICs maintenance of saturation conditions. The GOTHIC modeling is more realistic and removes some of the excessive conservatism in the COPATTA modeling.
The licensee performed a sensitivity study by adding a high flow volumetric recirculation fan in the GOTHIC model. Adding a new fan flow path is intended to quickly remove the suspended droplets using the flow rate from the fan. This change modified the droplet removal process, enhancing thermal equilibrium as the drops enter the containment space and nearly replicated the COPATTA model pressure profile.
Figure 3.0-2 shows much closer agreement between COPATTA and GOTHIC containment atmosphere pressure profiles when using the recirculation fan modeling and reduced droplet size in GOTHIC model.
The NRC staff finds the benchmarking of the GOTHIC model with the COPATTA model acceptable because the sensitivity study showed the containment pressure profiles from the two models are nearly identical.
3.3.2 Containment Vapor and Sump Temperature Benchmarking Enclosure attachment 5 to the LAR, figure 3.0-3 shows the containment vapor and sump temperature benchmarking results. The trend of the COPATTA vapor temperature profile is similar to the pressure profile where there is a significant increase to the containment pressure compared to GOTHIC profile. The licensees explanation for this trend is as follows:
This spike in temperature (or superheating) is due to how COPATTA treats suspended liquid in the vapor. At the end of each time step, COPATTA removes the liquid phase from the containment atmosphere and adds it to the sump (Reference [4]). In GOTHIC, the liquid enters as droplets that remain suspended in the atmosphere at saturation, with no superheat (Reference [5]). The GOTHIC modeling is more realistic and removes some of the excessive conservatism in the COPATTA modeling.
Enclosure attachment 5 to the LAR, figure 3.0-4 shows the temperature profiles from the recirculation fan sensitivity case described in section 3.2.1 of enclosure attachment 5. These profiles show much close agreement between COPATTA and GOTHIC containment atmosphere temperature profiles. The reason for the difference noted in figure 3.0-3 is that COPATTA removes the suspended liquid from the containment atmosphere and directly adds it to the sump whereas the GOTHIC model is more realistic in which the droplets remain suspended in the containment atmosphere.
Regarding the sump temperature profiles, the licensee states in section 3.2.2, Containment Vapor and Sump Temperature Benchmarking, of enclosure attachment 5 to the LAR :
Temperature differences in the short term are due to the way the codes treat the liquid phase introduced to the containment atmosphere from the break. It is also important to note for the short-term comparison that there is no initial inventory in the sump. In GOTHIC, the liquid phase emerging from the break is modeled as droplets that remain suspended in the containment atmosphere before either vaporizing or eventually settling to the sump (Reference [5]). In COPATTA, any portion of the blowdown that emerges in the liquid phase is added to the sump at the saturation temperature at the end of each time step (Reference [4])
Since both the COPATTA and GOTHIC models do not credit heat transfer between the sump and the containment atmosphere, the sump temperature is unimportant in the model until recirculation initiates at 1403 seconds.
It is noted that the COPATTA results bound the GOTHIC results before the recirculation phase.
During the recirculation phase, the calculated sump temperatures from GOTHIC and COPATTA show good agreement as illustrated in figure 3.0-3 of enclosure attachment 5 to the LAR.
3.3.3 Containment Liner Temperature Benchmarking Enclosure attachment 5 to the LAR, figure 3.0-5 shows the containment liner temperature profile benchmarking results. The COPATTA and GOTHIC graphs are nearly identical. In section 3.2.3, Containment Liner Temperature Benchmarking, of enclosure attachment 5, the licensee states that, The slight variation in temperature is attributed to the variation in the way the Uchida condensation heat transfer coefficient is modeled, in that GOTHIC provides a more precise Uchida correlation curve and yields a slightly lower condensing heat transfer coefficient for the steam-to-gas mass ratios utilized in the analysis.
3.3.4 Transient Energy Profiles Benchmarking Enclosure attachment 5 to the LAR, figures 3.0-6 through 3.0-11 show the transient energy profiles for the following:
Integrated Heat Sink Energy Spray Heat Exchanger Energy Spray Energy Reactor Vessel Energy Steam Energy Sump (Liquid) Energy Figure 3.0-6 shows the comparison of the COPATTA and GOTHIC results for the integrated heat sink energy. The heat sink energy in COPATTA is the integral of the heat transfer rate over time. The COPATTA heat sink energy profile starts near zero at the beginning of the transient run. The GOTHIC heat sink energy profile has a significant amount of reported energy after the first printed time interval. This indicates there is a discrepancy between the reference point used in determining the integrated heat sink energy that COPATTA uses compared to GOTHIC code.
Therefore, an adjustment is required to compare the COPATTA output to the GOTHIC output.
The licensee removed the discrepancy between the COPATTA and GOTHIC results by making an adjustment to the initial energy term in GOTHIC by setting the energy at the initial time interval to zero, which results in nearly identical heat sink energy profiles for the COPATTA and GOTHIC models. The NRC staff finds the adjustment acceptable and necessary for comparing the output of the two codes because of their inherent difference in determining the integrated energy at the beginning of the transient.
Figure 3.0-7 shows nearly identical results from the COPATTA and GOTHIC models for the containment spray heat exchanger energy (British thermal unit (Btu)).
Figure 3.0-8 shows nearly identical results from the COPATTA and GOTHIC models for the total integrated containment spray energy (Btu).
Figure 3.0-9 shows nearly identical results from the COPATTA and GOTHIC models for the reactor vessel energy (Btu).
Figure 3.0-10 shows the steam energy (Btu) from the COPATTA and GOTHIC models. Steam energy is the internal energy of the steam relative to energy at 32°F contained in the containment atmosphere. A discrepancy was noted among the results. In order to compare the COPATTA output to the GOTHIC output, the licensee made an adjustment equal to the difference between the initial energy calculated in GOTHIC at 0.1 seconds and the initial energy calculated in COPATTA at 0.1 seconds. This adjustment results in a vapor energy profile corrected to the initial energy. The adjustment resulted in nearly identical profiles. The NRC staff finds the adjustment acceptable and necessary for comparing the output of the two codes because of their inherent difference in determining the energy at the beginning of the transient.
Figure 3.0-11 shows the sump energy (Btu) from COPATTA and GOTHIC models. Sump energy is the internal energy of the water in the sump relative to energy at 32°F. Regarding the discrepancy between the results, the licensee states, in section 3.2.4.6, Sump Energy Benchmarking, of enclosure attachment 5 to the LAR:
The discrepancy early in time is attributed to the difference in the way the codes treat the liquid phase introduced to the containment atmosphere from the break as discussed in Section 3.2.2 [of Enclosure Attachment 6]. In GOTHIC, the liquid phase emerging from the break is assumed to enter in droplet form and remain suspended in the containment atmosphere for some time before reaching the sump. Note that when combining the GOTHIC sump and droplet energy, the resulting profile is nearly identical to the COPATTA sump energy profile, confirming that the variation is due to the difference in treatment of the liquid phase emerging from the break.
The NRC staff finds the transient energy benchmarking of the GOTHIC code with the COPATTA code is acceptable because the containment spray heat exchanger energy, integrated containment spray energy, and reactor vessel energy show identical profiles. For the integrated heat sink energy and steam energy, the licensee appropriately removed the discrepancy by adjusting the GOTHIC code and obtained identical profiles. Regarding the sump energy, the NRC staff finds the GOTHIC and COPATTA code difference reasonable because the profiles become nearly identical considering the difference in treatment of the liquid phase emerging from the break by the two codes.
3.3.5 Transient Mass Profiles Benchmarking Enclosure attachment 5 to the LAR, figure 3.0-12 shows the water mass inside the RCS, the containment sump, the steam, air, and water droplet mass in the containment atmosphere throughout the duration of the transient. These graphs provide the following observations:
The total mass, steam mass, RCS mass, and air mass are all nearly identical between the GOTHIC and COPATTA codes.
The RCS water mass remains constant, showing nearly identical results between the GOTHIC and COPATTA codes.
The air mass shows nearly identical results between the GOTHIC and COPATTA codes.
The sump water mass calculated in the COPATTA code is equal to the mass of sump water plus droplet water output by the GOTHIC code, which are nearly identical.
The transient steam mass profile nearly follows the containment pressure profile, indicating a containment pressure response due to a variation in containment atmosphere steam mass.
The NRC staff finds the transient mass profiles benchmarking results between the COPATTA and GOTHIC codes acceptable because the total mass, RCS water mass, air mass, and sump water mass profiles are nearly identical.
3.3.6 NRC Staff Evaluation of Benchmarking In sections 3.3.1 through 3.3.5 above, the NRC staff reviewed the licensees benchmarking of the GOTHIC code results with the COPATTA code results. The NRC staff finds the use of the GOTHIC code for LOCA containment response analysis is acceptable because its results for the containment pressure, vapor temperature, sump water temperature, containment liner temperature, transient energy and mass profiles are nearly identical with the results of the currently used COPATTA code.
3.4 Technical Analysis of Proposed Changes to the SIT Pressure Bands The licensee evaluated the impact of the proposed changes to the SIT pressure bands on the ECCS performance analysis of record (AOR) for both Westinghouse fuel (CE16STD and CE16NGF) and Framatome fuel (CE16HTP). Each evaluation considered the effects to the AOR for a large break LOCA, a small break LOCA, and during long-term cooling (LTC) post-accident.
In each evaluation, the licensee states that the proposed change to the lower limit of the SIT pressure band involved changing the units from psig to psia and removing instrument uncertainties from the TS values, with no impact on the ECCS AOR. The NRC staff reviewed the summary of the AOR as described in the Palo Verde UFSAR (Reference 3) and concluded that the lower limit setpoint is not part of the accident analyses; therefore, these proposed changes do not affect the AOR.
3.4.1 Westinghouse CE16STD and CE16NGF Fuel - Impact on ECCS AOR From Changes to the SR 3.5.1.3 SIT Pressure Band
- a. CE16STD Fuel Large Break LOCA In the LAR, the licensee states that the CE16STD large break LOCA AOR models maximum SIT pressure in the limiting peak cladding temperature (PCT) and limiting peak local oxidation cases. As a result, the licensee reanalyzed both cases with a maximum SIT pressure of 675 psia. The result of the reanalysis shows a 2°F increase in PCT (to 2108°F) and unchanged peak local oxidation (value remains at 11.9 percent after rounding). The NRC staff reviewed the licensees engineering evaluations described in the LAR supplement dated March 28, 2025 (Reference 2), that were performed to determine the effects of the proposed SIT pressure increase on the AOR and confirmed there was no change to the methodology, inputs, or assumptions used, apart from the SIT pressure. The NRC staff concluded these changes remain within the acceptance criteria of 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, and are therefore acceptable for the CE16STD fuel for large break LOCA.
- b. CE16NGF Fuel Large Break LOCA For CE16NGF fuel, the licensee states that the large break LOCA AOR only models maximum SIT pressure in the limiting PCT case. When reevaluated with the proposed peak SIT pressure of 675 psia, the licensee states there is an increase in PCT of 2°F (to 2132°F). The NRC staff reviewed the licensees engineering evaluations described in the LAR supplement dated March 28, 2025, that were performed to determine the effects of the proposed SIT pressure increase on the AOR and confirmed there was no change to the methodology, inputs, or assumptions used, apart from the SIT pressure. The NRC staff concluded these changes remain within the acceptance criteria of 10 CFR 50.46, and are therefore acceptable for the CE16NGF fuel for large break LOCA.
- c. CE16STD Fuel Small Break LOCA In the LAR, the licensee states that for the small break LOCA AOR, SIT pressure is set at an intentionally low value (200 psia) to prevent SIT injection. This approach allows the model to be conservative and not take any credit for SIT injection to mitigate the small break LOCA. The NRC staff concludes that the 200 psia value used in the AOR is well below the actual TS values for SIT pressure, and therefore, will not be affected by the proposed change.
- d. CE16NGF Fuel Small Break LOCA Similar to the analysis for the CE16STD fuel, the licensee states in the LAR that the AOR for CE16NGF fuel is a conservative model, which prevents SIT injection because it is not credited in the small break LOCA accident analysis. The existing AOR for the CE16NGF fuel small break LOCA uses 602 psia as an input for the SIT pressure. The NRC staff confirms that the proposed SIT pressure band of 602 psia and 675 psia does not affect the existing AOR, and its results remain valid.
- e. Long-Term Cooling For LTC after the core is quenched but before the plant is secured from the accident, the licensee evaluated two primary concerns in the LAR: maintaining the core at a safe temperature and avoiding boric acid precipitation. The proposed changes to the allowable SIT pressure band will have an impact on LTC due to the increased density of the water in the SIT with the proposed increase in allowed maximum pressure. This increased density results in an increased mass of SIT water. An increase in mass of SIT water has the potential to affect both decay heat removal and boric acid precipitation because those calculations use mass of available water as an input.
In enclosure attachment 6 to the LAR, section 3.3.1, CE16STD and CE16NGF LTC Boric Acid Precipitation Control, the licensee states:
Increasing the maximum SIT pressure from 651 psia to 675 psia (at 50°F) increases the SIT water density by 0.016%. Since the change in liquid mass is directly proportional to liquid density, the resulting SIT liquid mass for the increased maximum SIT pressure is calculated to be 487,890 lbm [(pound mass)]
(= 487,812 lbm x 1.00016).
For the boric acid precipitation control AOR, the licensee states in the LAR that it uses a SIT liquid mass of 487,812 lbm for a maximum SIT pressure of 651 psia, which is rounded up to 488,000 lbm. The 487,890 lbm of water calculated from the density change due to the proposed SIT pressure change remains within the rounded value modeled in the boric acid precipitation control AOR.
For decay heat removal, the licensee states in the LAR that the AOR for LTC decay heat removal uses the same rounded value (488,000 lbm). As a result, the effects of the mass increase due to the increased density of the increased SIT pressure are the same as in the boric acid precipitation control AOR.
The NRC staff reviewed the conclusions of these AORs and determined that the existing analyses for LTC remain valid with the proposed SIT pressure increase to a maximum of 675 psia for CE16STD and CE16NGF fuel.
3.4.2 Framatome CE16HTP Fuel - Impact on ECCS AOR From Changes to the SR 3.5.1.3 SIT Pressure Band
- a. Framatome CE16HTP Fuel Large Break LOCA The large break LOCA AOR for Framatome CE16HTP fuel uses SIT pressure as an input. The licensee states in the LAR that the AOR was redone with the new proposed SIT pressure band to ensure the results remain within the acceptance criteria of 10 CFR 50.46. This revised analysis resulted in changes similar to those seen with Westinghouse fuel, namely changes to PCT, total hydrogen production from the cladding oxidation, and maximum local oxidization of cladding. The licensee states the results of this revised analysis in the LAR, with a 39°F increase in PCT (to 1791°F) and a -0.40 percent increase in maximum local oxidization (to 1.97 percent). The licensee did not directly calculate total hydrogen production from cladding oxidation but instead used the calculated core wide oxidation to conservatively bound the value.
The core wide oxidation changed by -0.006 percent to 0.014 percent. The NRC staff concluded these changes remain within the acceptance criteria of 10 CFR 50.46, and are therefore acceptable for the CE16HTP fuel for large break LOCA.
In addition, the licensee states in the LAR that there is some difference from the generally approved large break LOCA methodology and the licensees site-specific AOR, as noted in the NRC safety evaluation for the Palo Verde Operating Licenses in Amendment 212 regarding implementation of Framatome CE16HTP fuel (Reference 9). The NRC staff reviewed the licensees LAR and the previously performed safety evaluation and determined there is no impact on the previously accepted differences with the proposed changes listed in the LAR.
- b. Framatome CE16HTP Fuel Small Break LOCA Although SIT pressure is an input into the Framatome CE16HTP small break LOCA AOR, the licensee states in the LAR that the assumed minimum SIT pressure is 602 psia, which will not be affected by the proposed TS changes. The NRC staff confirms that the proposed SIT pressure band of 602 psia and 675 psia does not affect the existing AOR and its results remain valid.
- c. Framatome CE16HTP Fuel Long-Term Cooling The licensee states that the attributes of the CE16HTP Framatome fuel applicable in an LTC analysis are similar to those in Westinghouse CE16NGF fuel, as evaluated and discussed in the NRC safety evaluation for the Palo Verde Operating Licenses Amendment 212 regarding implementation of Framatome CE16HTP fuel (Reference 9). The NRC staff reviewed the licensees LAR and the previously performed safety evaluation and determined the results for Framatome fuel will be comparable to those of the Westinghouse fuel as discussed in section 3.4.1.e of this safety evaluation.
As a result, the NRC staff determined that the existing analyses for LTC remain valid with the proposed SIT pressure increase to a maximum of 675 psia for CE16HTP fuel.
3.4.3 Post-LOCA Containment Pressure, Vapor Temperature, Sump Temperature Responses, and Net Positive Suction Head Analysis.
The change in the upper limit of the SIT pressure band potentially impacts the LOCA containment pressure, vapor temperature, and sump fluid temperature analyses since the nitrogen used to pressurize the four SITs is released to the containment atmosphere during a LOCA event. The licensee used the GOTHIC code as a replacement to COPATTA as part of the methodology to perform calculations of the containment pressure and temperature response to various postulated pipe breaks. The DEDLS break LOCA with maximum ECCS (i.e., SI) flow was identified as the pipe break with the highest peak pressure.
Pressure and Temperature Response The licensee performed containment analysis to calculate the impact of an increase in mass and energy release due to the proposed increase in the upper limit of the SIT pressure band to 675 psia for the DEDLS break LOCA with maximum SI flow. The licensee used the GOTHIC benchmark model for the containment pressure and temperature analysis with the changes described below.
Changed SIT cover gas to nitrogen at 675 psia and with nitrogen release from all four SITs. The COPATTA and GOTHIC benchmark analyses both model air injection at 652 psia from one SIT.
Changed the shutdown cooling heat exchanger to use the more intricate GOTHIC water-to-water heat exchanger model to account for the overall heat transfer coefficient (U-factor) as a function of tube-side flow and tube-side inlet temperature. The COPATTA and GOTHIC benchmark analyses both model a simple shutdown cooling heat exchanger using a constant U-factor with shell-and tube heat exchanger geometry.
Changed the CS mass flow rate model to a volumetric flow rate model. The COPATTA and GOTHIC benchmark analyses both model the CS recirculation phase volumetric flow rate converted to a mass flow rate based on a single containment sump water temperature.
Included the SI and CS pump energy to the fluid to account for the `local temperature rise across the pump due to pump inefficiency and also the energy imparted to the fluid via an increase in pressure. The COPATTA and GOTHIC benchmark model did not include this energy.
Consistent with the recommendation in the GOTHIC Users Manual (Reference 6),
section 27.5 and in NRC-approved containment analyses performed using GOTHIC for Kewaunee (Reference 7) and Prairie Island (Reference 8), the licensee modeled the liquid phase of the break effluent as a uniform drop field with 100 microns (0.00394 inches) drop diameter.
The licensees GOTHIC analysis calculated results are as follows:
Peak containment pressure is 71.46 psia, which is reported with inclusion of discretionary margin as 72.05 psia (equivalent to 57.85 psig based on a local atmospheric pressure of 14.2 psia). This is same as in Palo Verde UFSAR table 6.2.1-9, Summary of Calculated Containment Pressure and Temperatures Analyzed at 102% of 3990 MWt [Megawatt Thermal]. In TS 5.5.16b, the design basis LOCA peak containment pressure (Pa) for integrated leak rate test conservatively rounded to 58.0 psig remains unchanged, which is bounded by the containment design pressure of 60 psig.
Containment pressure drops to 34.2 psia at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, which is reported with inclusion of discretionary margin as 35.59 psia (equivalent to 21.39 psig based on a local atmospheric pressure of 14.2 psia). In compliance with the NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR
[Light-Water Reactor] Edition, Section 6.2.1.1.A, PWR [Pressurized-Water Reactor]
Dry Containments, Including Subatmospheric Containments, Revision 3, March 2007 (Reference 10) requirement, the containment pressure of 21.39 psig at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is less than 50 percent of the peak pressure of 57.85 psig.
Peak containment vapor temperature is 280.06°F, which is reported with inclusion of discretionary margin as 308.41°F. This is the same as in Palo Verde UFSAR table 6.2.1-9.
The NRC staff finds the licensees LOCA containment response analysis acceptable because the proposed change in the SIT pressure band does not impact the AOR values of the peak pressure and temperature documented in Palo Verde UFSAR table 6.2.1-9.
Sump Temperature Response The licensee states that due to the proposed change in the SIT pressure band, the post-LOCA sump maximum temperature for the limiting case that bound all licensed fuel types in the current analysis reported in the Palo Verde UFSAR has not changed. The licensee further states that GOTHIC code analysis results in lower post-LOCA sump temperature for the most limiting case than the COPATTA code. However, the licensee conservatively elected to report the previously calculated COPATTA maximum temperature for the bounding LOCA case. The GOTHIC analysis results therefore have discretionary margin added that results in a bounding calculation in the AOR.
Net Positive Suction Head Analysis As noted in the LAR supplement dated March 28, 2025, figure 7.2, the sump temperature response in the post-LOCA recirculation phase is nearly identical to the AOR response, therefore, the NRC staff determines that the net positive suction head (NPSH) AOR for the pumps that draw water from the sump during the LOCA recirculation phase is not affected because the sump temperature is the most dominant parameter that could affect the NPSH.
3.5 Technical Conclusion 3.5.1 ECCS Analysis The NRC staff reviewed the licensees evaluation determining the impact of the proposed change in the SIT pressure band on the ECCS performance AOR for both Westinghouse fuel (CE16STD and CE16NGF) and Framatome fuel (CE16HTP). Each evaluation considered the effects to the AOR for a limiting large break LOCA, a limiting small break LOCA, and post-LOCA long-term cooling. The NRC staff determined that the impact of the proposed change is acceptable because the results of the evaluations remain within the acceptance criteria of 10 CFR 50.46; 10 CFR Part 50, Appendix K; and GDC 35. As a result, 10 CFR 50.36(c)(3) is met because the proposed modification to the SR assures that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met 3.5.2 Containment Analysis The NRC staff reviewed the licensees evaluation determining the impact of the proposed change in the SIT pressure band, using the GOTHIC code, on the containment pressure and temperature AOR. The NRC staff finds no impact because the AOR values of peak containment pressure and vapor temperature documented in Palo Verde UFSAR table 6.2.1-9 are not affected. In addition, the sump temperature response and NPSH analysis documented in UFSAR, section 6.2.2, Containment Heat Removal Systems, are not affected by the proposed change.
Consistent with the requirement in the Kewanuee and Prairie Island amendments (References 7 and 8) stating that the heat transfer MDLM shall not be used for licensing calculations, the LOCA containment pressure and temperature analysis did not use this model. The licensees analysis used the Tagami and Uchida condensing heat transfer correlations as described in Palo Verde UFSAR section 6.2.1.1.3.1, Containment Peak Pressure and Temperature Analysis.
In compliance with GDC 16, the GOTHIC analysis confirms that in the event of the limiting DEDLS break LOCA with maximum ECCS flow, the SI and CS systems cool the reactor core, and return the containment to near atmospheric pressure. The containment, SI, and CS system along with containment isolation system ensure the capability of containing any uncontrolled release of radioactivity.
In compliance with GDC 50, the GOTHIC analysis confirms that the containment structure and its internal compartments, including access openings, penetrations, and the containment heat removal system accommodate the calculated pressure and temperature conditions resulting from the limiting DEDLS break LOCA with maximum ECCS flow, without exceeding the design leakage rate and with a sufficient margin.
In sum, the results of the post-LOCA containment pressure and temperature analysis continue to meet GDC 16 and 50, thereby demonstrating the adequacy of the containment structure design to support the proposed changes to the SIT pressure bands.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Arizona State official was notified of the proposed issuance of the amendments on May 4, 2025. No comments were received from the State official.
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 and change SRs. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, as published in the Federal Register on January 14, 2025 (90 FR 3256), and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the 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 Commission's 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.
Horton, T., Arizona Public Service Company, letter to U.S. Nuclear Regulatory Commission, Palo Verde Nuclear Generating Station Units 1, 2, and 3, Docket Nos.
STN 50-528,59-529, and 50-530 Renewed Operating License Number NPF-41, NPF-51, and NPF-74 License Amendment Request to Revise the Technical Specifications 3.5.1 and 3.5.2 Safety Injection Tank Pressure Bands, and to Use GOTHIC Code, dated August 28, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24241A278.
- 2.
Spina, J., Arizona Public Service Company, letter to U.S. Nuclear Regulatory Commission, Palo Verde Nuclear Generating Station Units 1, 2, and 3 Docket Nos.
STN 50-528,59-529, and 50-530 Renewed Operating License Number NPF-41, NPF-51, and NPF-74 Supplement to License Amendment Request to Revise the Technical Specifications 3.5.1 and 3.5.2 Safety Injection Tank Pressure Bands, and to Use GOTHIC Code, dated March 28, 2025 (ML25090A018).
- 3.
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Chawla, M. L., U.S. Nuclear Regulatory Commission, letter to J. M., Solymossy, Nuclear Management Company, LLC, Prairie Island Nuclear Generating Plant, Units 1 and 2 -
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Principal Contributors: A. Sallman, NRR J. Ambrosini, NRR Date: May 30, 2025
- via eConcurrence NRR-058 OFFICE NRR/DORL/LPL4/PM*
NRR/DORL/LPL4/LA*
NRR/DSS/SNSB/BC NRR/DSS/STSB/BC NAME WOrders PBlechman DMurdock SMehta DATE 5/7/2025 5/13/2025 5/16/2025 5/16/2025 OFFICE OGC NRR/DORL/LPL4/BC*
NRR/DORL/LPL4/PM*
NAME STurk TNakanishi WOrders DATE 5/22/2025 5/30/2025 5/30/2025