ML23356A003

Issuance of Amendments Revise Technical Specification 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

ML23356A003
Person / Time
Site:	Prairie Island
Issue date:	01/17/2024
From:	Ballard B NRC/NRR/DORL/LPL3
To:	Conboy T Northern States Power Company, Minnesota
Ballard B, NRR/DORL/LPL3
References
EPID L-2022-LLA-0184
Download: ML23356A003 (27)

Text

January 17, 2024 Mr. Thomas Conboy Site Vice President Northern States Power Company - Minnesota Prairie Island Nuclear Generating Plant 1717 Wakonade Drive East Welch, MN 55089

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 -

ISSUANCE OF AMENDMENT NOS. 243 AND 231 RE: REVISE TECHNICAL SPECIFICATION 5.6.6, REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (EPID L-2022-LLA-0184)

Dear Mr. Conboy:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 243 to Renewed Facility Operating License No. DPR-42 and Amendment No. 231 to Renewed Facility Operating License No. DPR-60 for the Prairie Island Nuclear Generating Plant, Units 1 and 2, respectively. The amendments consist of changes to the technical specifications (TSs) in response to your application dated December 2, 2022, as supplemented by letters dated March 28, 2023, and December 20, 2023.

The amendments revise Technical Specification 5.6.6, Reactor Coolant System (RCS)

Pressure and Temperature Limits Report (PTLR), to replace the current PTLR method with more recent analytical methods and remove a reference to an American Society of Mechanical Engineers Code Case that is no longer needed with the updated methods.

A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Brent T. Ballard, Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306

Enclosures:

1. Amendment No. 243 to DPR-42

2. Amendment No. 231 to DPR-60

3. Safety Evaluation cc: Listserv

NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 243 Renewed License No. DPR-42

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Northern States Power Company (NSPM, the licensee), dated December 2, 2022, as supplemented by letters dated March 28, 2023, and December 20, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-42 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 243, are hereby incorporated in the renewed operating license. NSPM shall operate the facility in accordance with the Technical Specifications.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Jeffrey A. Whited, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: January 17, 2024 Jeffrey A.

Whited Digitally signed by Jeffrey A. Whited Date: 2024.01.17 15:30:09 -05'00'

NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 231 Renewed License No. DPR-60

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Northern States Power Company (NSPM, the licensee), dated December 2, 2022, as supplemented by letters dated March 28, 2023, and December 20, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-60 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 231, are hereby incorporated in the renewed operating license. NSPM shall operate the facility in accordance with the Technical Specifications.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Jeffrey A. Whited, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: January 17, 2024 Jeffrey A.

Whited Digitally signed by Jeffrey A. Whited Date: 2024.01.17 15:30:55 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NOS. 243 AND 231 RENEWED FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 DOCKET NOS. 50-282 AND 50-306 Replace the following pages of the Renewed Facility Operating License Nos. DPR-42 and DPR-60 with the attached revised pages. The changed areas are identified by a marginal line.

Renewed Facility Operating License No. DPR-42 REMOVE INSERT Page 3 Page 3 Renewed Facility Operating License No. DPR-60 REMOVE INSERT Page 3 Page 3 Technical Specifications Replace the following pages of 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.

REMOVE INSERT 5.6-7 5.6-7 5.6-8 5.6-8 5.6-9 5.6-9 Renewed Operating License No. DPR-42 Amendment No. 243 (3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, NSPM to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (6) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to transfer byproduct materials from other job sites owned by NSPM for the purpose of volume reduction and decontamination.

C.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter l:

Part 20, Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; 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 NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 1677 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 243, are hereby incorporated in the renewed operating license.

NSPM shall operate the facility in accordance with the Technical Specifications.

(3) Physical Protection NSPM shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Renewed Operating License No. DPR-60 Amendment No. 231 (3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, NSPM to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (6) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to transfer byproduct materials from other job sites owned by NSPM for the purposes of volume reduction and decontamination.

C.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter l:

Part 20, Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; 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 NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 1677 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 231, are hereby incorporated in the renewed operating license.

NSPM shall operate the facility in accordance with the Technical Specifications.

(3) Physical Protection NSPM shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains

Reporting Requirements 5.6 Prairie Island Units 1 and 2 5.6-7 Unit 1 +/- Amendment No. 243 Unit 2 +/- Amendment No. 231 5.6 Reporting Requirements (continued) 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) a.

RCS pressure and temperature limits for heat-up, cooldown, low temperature operation, criticality, and hydrostatic testing, OPPS arming, PORV lift settings and Safety Injection Pump Disable Temperature as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3, ³RCS Pressure and Temperature (P/T) Limits';

LCO 3.4.6, ³RCS Loops - MODE 4';

LCO 3.4.7, ³RCS Loops - MODE 5, Loops Filled';

LCO 3.4.10, ³Pressurizer Safety Valves';

LCO 3.4.12, ³Low Temperature Overpressure Protection (LTOP) +/-

Reactor Coolant System Cold Leg Temperature (RCSCLT) > Safety Injection (SI) Pump Disable Temperature';

LCO 3.4.13, ³Low Temperature Overpressure Protection (LTOP) +/-

Reactor Coolant System Cold Leg Temperature (RCSCLT) < Safety Injection (SI) Pump Disable Temperature'; and LCO 3.5.3, ³ECCS - Shutdown'.

b.

The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigating System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-14040-NP-A, Revision 4, ³Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,' May 2004.

2.

WCAP-18124-NP-A, Revision 0, ³Fluence Determination with RAPTOR-M3G and FERRET,' July 2018, and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0, ³Fluence Determination with Raptor-M3G and FERRET +/- Supplement for Extended Beltline Materials,' May 2022, shall be used as an alternative to Section 2.2 of WCAP-14040-NP-A.

Reporting Requirements 5.6 Prairie Island Units 1 and 2 5.6-8 Unit 1 +/- Amendment No. 243 Unit 2 +/- Amendment No. 231 5.6 Reporting Requirements 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued) c.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

Changes to the curves, setpoints, or parameters in the PTLR resulting from new or additional analysis of beltline material properties shall be submitted to the NRC prior to issuance of an updated PTLR.

5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.8, ³Steam Generator (SG) Program.' The report shall include:

a.

The scope of inspections performed on each SG; b.

The nondestructive examination techniques utilized for tubes with increased degradation susceptibility; c.

For each degradation mechanism found:

1.

The nondestructive examination techniques utilized; 2.

The location, orientation (if linear), measured size (if available),

and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported; 3.

A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and 4.

The number of tubes plugged during the inspection outage.

Reporting Requirements 5.6 Prairie Island Units 1 and 2 5.6-9 Unit 1 +/- Amendment No. 243 Unit 2 +/- Amendment No. 231 5.6 Reporting Requirements (continued) 5.6.8 Steam Generator Tube Inspection Report (continued) d.

An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results; e.

The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG; and f.

The results of any SG secondary side inspections.

5.6.8 EM Report When a report is required by Condition C or I of LCO 3.3.3, "Event Monitoring (EM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 243 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-42 AND AMENDMENT NO. 231 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-60 NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306

1.0 INTRODUCTION

By application dated December 2, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22343A257, as supplemented by letters dated March 28, 2023 (ML23087A250), and December 20, 2023 (ML23354A155), respectively, Northern States Power Company, a Minnesota Corporation (NSPM, the licensee), doing business as Xcel Energy, requested changes to the Technical Specifications (TSs) for Prairie Island Nuclear Generating Plant, Units 1 and 2 (PINGP, Prairie Island).

The supplemental letter(s) dated March 28, 2023, and December 20, 2023, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on March 7, 2023 (88 FR 14184).

The proposed changes would revise Technical Specification (TS) 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), to update the methods used to determine RCS pressure and temperature (P-T) limits for operation of the PINGP, Units 1 and 2 and to revise the P-T limits curves as a result of the analysis of reactor vessel material surveillance testing results.

1.1

System Description

Components of the RCS are designed to withstand effects of loadings due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. Pressure and temperature changes are limited during RCS heatup and cooldown within the design assumptions and the stress limits for normal operation.

Pressure and temperature limits have been established for heatup, cooldown, and inservice leak and hydrostatic testing and are documented in the PINGP PTLR.

2.0 REGULATORY EVALUATION

2.1 Applicable Regulations The regulations in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, Technical Specifications, paragraph (a)(1), requires that each operating license application for a production or utilization facility include proposed TSs and a summary statement of the bases for such specifications.

Paragraph (c)(5) of 10 CFR 50.36, Administrative controls, are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

The regulations in 10 CFR 50.60, Acceptance criteria for fracture prevention measures for light-water nuclear power reactors for normal operation, require that all light-water nuclear power reactors meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in 10 CFR Part 50, Appendices G, Fracture Toughness Requirements, and H, Reactor Vessel Material Surveillance Program Requirements.

The regulations in 10 CFR 50.61, Fracture toughness requirements for protection against pressurized thermal shock (PTS) events, requires, for pressurized water reactors, that the PTS reference temperature (RTPTS), of the reactor vessel shell material be limited at the end of plant life to less than the PTS screening criterion of 270 °F for plates, forgings, and axial weld materials, and 300 °F for circumferential weld materials.1 The regulations in 10 CFR Part 50, Appendix G, establish fracture toughness requirements to provide adequate margins of safety to maintain the integrity of the reactor coolant pressure boundary (RCPB) in nuclear power plants. P-T limit requirements for the reactor vessel are established in paragraph IV.A.2 and Table 1 of this rule. Paragraph IV.A.2 and Table 1 specify that P-T limit curves and minimum temperature requirements for the reactor vessel are defined by the operating condition (i.e., pressure testing or normal operation, including anticipated operational occurrences), the reactor vessel pressure, whether or not fuel is in the vessel, and whether the core is critical. In Table 1, the reactor vessel pressure is defined as a percentage of the preservice system hydrostatic test pressure. The requirements for both the P-T limit curves and the minimum temperature must be met for all normal operating and pressure test conditions.

Additionally, 10 CFR Part 50, Appendix G, requires that applicable surveillance data from reactor vessel material surveillance programs be incorporated into the calculations of the P-T limits and that the P-T limits be generated using a method that accounts for the effects of neutron irradiation on the material properties of the reactor vessel beltline materials.

1 If this criterion cannot be satisfied, 10 CFR 50.61 may allow continued operation if other specified criteria are satisfied.

The regulations in 10 CFR Part 50, Appendix H, require a material surveillance program to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region which result from exposure of these materials to neutron irradiation and the thermal environment.

2.2 Design Criteria Units 1 and 2 of PINGP were designed to comply with the Atomic Energy Commission (AEC)

General Design Criteria (GDC) for Nuclear Power Plant Construction Permits, draft for comment, 32 FR 10214 (July 11, 1967) (1967 Draft GDC). PINGP was not licensed to the 10 CFR Part 50, Appendix A, GDC and was designed and constructed to comply with the Principal Design Criteria (PDC) proposed in NSPMs (then Northern States Power) construction permit application, which reflected or adopted the 1967 Draft GDC. The following discussion addresses the proposed changes with respect to meeting the requirements of the applicable 1967 Draft GDC to which PINGP was licensed. The NRC staff review was performed in consideration of the requirements contained in GDC.

Draft 1967 Criterion 9 - Reactor Coolant Pressure Boundary The reactor coolant pressure boundary shall be designed and constructed so as to have an exceedingly low probability of gross rupture or significant leakage throughout its design lifetime.

Draft 1967 Criterion 34 - Reactor Coolant Pressure Boundary Rapid Propagation Fracture Prevention The reactor coolant pressure boundary shall be designed to minimize the probability of rapidly propagating type failures. Consideration shall be given (a) to the notch-toughness properties of materials extending to the upper shelf of the Charpy transition curve, (b) to the state of stress of materials under static and transient loadings, (c) to the quality control specified for materials and component fabrication to limit flaw sizes, and (d) to the provisions for control over service temperature and irradiation effects which may require operational restrictions.

Draft 1967 Criterion 35 - Reactor Coolant Pressure Boundary Brittle Fracture Prevention Under conditions where reactor coolant pressure boundary system components constructed of ferritic materials may be subjected to potential loadings, such as a reactivity-induced loading, service temperature shall be at least 120° F above the nil ductility transition (NDT) temperature of the component material if the resulting energy release is expected to be absorbed by plastic deformation or 60° F above the NDT temperature of the component material if the resulting energy release is expected to be absorbed within the elastic strain energy range.

2.3 Applicable Guidance Regulatory Guide (RG) 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, dated May 1988 (ML003740284), describes procedures for calculating the adjusted nil-ductility transition reference temperature RTNDT (ART) due to neutron irradiation on the reactor vessel materials.

The guidance in RG 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, (ML010890301) provides guidance on methods for determining reactor pressure vessel fluence that are acceptable to the NRC staff. The guidance in RG 1.190 states that an acceptable neutron fluence calculation has the following attributes:

Fluence estimation using an appropriate calculational methodology, Analytic uncertainty analysis identifying possible sources of uncertainty, Comparisons with benchmark measurements and calculations from applicable test facilities including:

o Plant-specific operating reactor measurements o Pressure vessel simulator measurements o Calculational benchmarks The NRC Generic Letter (GL) 96-03, Relocation of Pressure and Temperature Limit Curves and Low Temperature Overpressure Protection System Limits, (ML031110004) permits relocation of the P-T limits from the TS to a PTLR. GL 96-03 calls for licensees to (1) generate their P-T limits in accordance with an NRC-approved methodology, (2) comply with 10 CFR Part 50, Appendices G and H, (3) reference NRC-approved methodologies in the TS, (4) define the PTLR in TS Section 1.0, (5) develop a PTLR to contain the P-T limit curves, and (6) modify applicable sections of the TS accordingly.

Regulatory Issue Summary (RIS) 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, dated October 14, 2014 (ML14149A165), provides evaluation guidance for P-T curves and PTLRs, including the consideration of neutron fluence and structural discontinuities in the development of P-T curves.

Guidance in NUREG 0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition (SRP), Chapter 16.0, Technical Specifications, Revision 3, dated March 2010 (ML100351425) provides NRC staff guidance for the review of TS.

3.0 TECHNICAL EVALUATION

3.1 Background

The American Society of Mechanical Engineers (ASME) Code,Section XI, Appendix G methodology for generating P-T limit curves is based upon the principles of linear elastic fracture mechanics. The basic parameter of this methodology is the stress intensity factor, KI, which is a function of the stress state in the component and flaw configuration. The ASME Code,Section XI, Appendix G requires a safety factor of 2.0 on stress intensities resulting from reactor pressure during normal and transient operating conditions, and a safety factor of 1.5 on these stress intensities for hydrostatic and pressure testing limits. The ASME Code,Section XI, Appendix G specifies that the P-T limits be generated by postulating a flaw with a depth that is equal to 1/4 of the reactor vessel shell thickness and a length equal to 1.5 times the reactor vessel section thickness. The critical locations in the reactor vessel shell thickness (T) for calculating heatup and cool-down P-T limit curves are the 1/4 thickness (1/4T) and 3/4 thickness (3/4T) locations, which correspond to the maximum depth of the postulated inside surface and outside surface defects, respectively.

The P-T limit curve calculations are based, in part, on the reference nil-ductility temperature (RTNDT) for the material, as specified in the ASME Code,Section XI, Appendix G. The RTNDT is the critical parameter for determining the critical or reference stress intensity factor (fracture toughness, KIC) for the material. As required by 10 CFR Part 50, Appendix G, RTNDT values for materials in the reactor vessel beltline region shall be adjusted to account for the effects of neutron radiation. RG 1.99, Revision 2, contains methodologies for calculating the ART due to neutron irradiation. The ART is defined as the sum of the initial (unirradiated) reference nil-ductility temperature (initial RTNDT), the mean value of the shift in reference temperature caused by irradiation (RTNDT), and a margin term. The RTNDT is a product of a chemistry factor (CF) and a fluence factor. The CF is dependent upon the amount of copper (Cu) and nickel (Ni) in the material and may be determined from the tables in RG 1.99, Revision 2, or from surveillance data. The fluence factor is dependent upon the neutron fluence at the postulated flaw depths described above. The margin term is dependent upon whether the initial RTNDT is a plant-specific or a generic value and whether the CF was determined using the tables in RG 1.99, Revision 2, or surveillance data. The margin term is used to account for uncertainties in the values of the initial RTNDT, the Cu and Ni contents, the neutron fluence and the calculational procedures.

To satisfy the requirements of 10 CFR Part 50, Appendix G, methods for determining neutron fluence are necessary to estimate the fracture toughness of the reactor vessel materials.

Appendix H, Reactor Vessel Material Surveillance Program Requirements, of 10 CFR Part 50, requires the installation of surveillance capsules, including material test specimens and flux dosimeters, to monitor changes in fracture toughness.

PINGP, Unit 1 and Unit 2, were approved for license renewal for an additional 20 years beyond the original 40-year licensed period, for a total of 60 years of operation. The period after the initial licensing term is known as the period of extended operation (PEO). For PINGP, Unit 1 and Unit 2, the PEO corresponds to 54 effective full-power years (EFPY).

3.2 WCAP-14040-NP-A, Revision 4 Methodology The analytical method proposed by the licensee to be used to determine the revised RCS P-T limits and develop Low Temperature Overpressure Protection System (LTOPS) setpoints for PINGP is described in the NRC-approved Topical Report WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves (ML050120209). The methodology contained in Revision 4 to WCAP-14040-NP-A was approved for licensing applications subject to the following three conditions described Section 4, Conclusion, of the associated NRC safety evaluation (SE):

Condition 1 requires that the licensee provide information consistent with GL 96-03 to demonstrate that the plant maintains a reactor vessel material surveillance program in accordance with Appendix H to 10 CFR Part 50. The purpose of the surveillance program is to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region that result from the exposure of these materials to neutron irradiation and to the thermal environment. Surveillance capsules are located in each vessel between the core and the reactor vessel near the inner vessel wall in the beltline region. They contain material specimens that are consequently exposed to the neutron irradiation and thermal environment.

Appendix H to 10 CFR Part 50 requires that licensees of commercial light-water nuclear power reactors with a peak neutron fluence exceeding 1x1017 n/cm2 (E > 1 MeV) at the end of the design life maintain a reactor vessel material surveillance program that tests irradiated material specimens that are located in surveillance capsules in the vessel. Units 1 and 2 of PINGP exceed these neutron fluence thresholds and therefore are subject to these requirements and must maintain reactor vessel surveillance programs in accordance with 10 CFR Part 50, Appendix H.Section IV.A of Appendix H to 10 CFR Part 50 requires that each surveillance capsule withdrawal and associated test results must be the subject of a summary technical report that is to be submitted to the NRC within 18 months of the date of the capsule withdrawal.

The PINGP PTLR describes the methodology requirements discussed in Provision 2 in the Table of Attachment 1 to GL 96-03, which are related to the reactor vessel material surveillance program. The licensee stated that its surveillance program complies with Appendix H to 10 CFR Part 50. The PINGP PTLR also includes the latest surveillance specimen removal schedules. By letter dated September 3, 2020 (ML20230A051), the NRC approved the current Reactor Vessel Surveillance Capsule Withdrawal Schedule for PINGP, Unit 1 and Unit 2. The most recent summary technical reports documenting the post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens from Capsule N for PINGP, Unit 1 and Unit 2, (ML22067A147 and ML23075A344, respectively) were performed in accordance with 10 CFR Part 50, Appendix H and American Society for Testing and Materials (ASTM)

Specification E185-82. Therefore, the NRC staff finds that the information submitted by the licensee has met the first condition of the SE for WCAP-14040-A, Revision 4.

Condition 2 requires that Revision 4 of WCAP-14040-A reflect the NRC staff conclusion that no exemption is required for licensee use of provisions in ASME Code Cases N-588, Attenuation to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels, N-640, Alternative Reference Fracture Toughness for Development of P-T Limit Curves, or N-641, Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements, in conjunction with the basic methodology contained in WCAP-14040, Revision 3, since these Code Cases are contained in the edition and addenda of the ASME Code incorporated by reference in 10 CFR 50.55a.

The licensee stated that Table A-1, Status of ASME Nuclear Code Cases Associated with the P-T Limit Curve/COMS Methodology, of Appendix A, Relevant ASME Nuclear Code Cases was updated in Revision 4 of WCAP-14040-A to indicate the date, edition, and addenda of ASME Code,Section XI, when the referenced code cases were approved by the ASME. As indicated in Condition 2, the editions and addenda of ASME Code,Section XI, that are listed in Table A-1 have been incorporated by reference in 10 CFR 50.55a.

The PTLR methodology used by PINGP, Unit 1 and Unit 2, to generate P-T limits curves based on WCAP-14040-A, Revision 4 is consistent with the methodology in ASME Code Section XI, Appendix G. Therefore, the NRC staff concludes that the updated information has adequately addressed the second condition in the SE to WCAP-14040-A, Revision 4 and the second condition has been met.

Condition 3 requires that the reactor vessel flange minimum temperature requirements be incorporated into a facilitys P-T curves until Appendix G to 10 CFR Part 50 is revised to modify the existing reactor vessel flange minimum temperature requirement or an exemption request to modify these requirements is approved by the NRC for a specific facility. The licensee referred to WCAP-14040-A, Revision 4, Section 2.9, Closure Head/Vessel Flange Requirements, which re-states this position. Appendix G to 10 CFR Part 50 has not been revised since WCAP-14040-A, Revision 4, including the associated SE, was issued on February 27, 2004.

The licensee proposes to incorporate the RPV flange minimum temperature into the proposed P-T limit curves. Therefore, the NRC staff concludes that the third condition in the SE to WCAP-14040-A, Revision 4, has been adequately addressed. The staff notes that there have been no changes to the requirements of Appendix G to 10 CFR Part 50 regarding the reactor vessel flange minimum temperature requirements, and the licensee is not requesting a plant-specific exemption to the requirements.

The NRC staff finds that the licensee has adequately addressed the three conditions required in the SE for the use of WCAP-14040-A, Revision 4. Therefore, the staff has concluded that by meeting all three conditions, amending the PTLR methodology to WCAP-14040-A, Revision 4, for PINGP, Unit 1 and Unit 2, is acceptable.

3.3 Pressure-Temperature Limits Methodology to the March 28, 2023, submittal contained WCAP-18746-NP, Revision 2, Prairie Island Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation, (ML23087A250). This report contains an evaluation of the reactor vessel integrity for PINGP, Unit 1 and Unit 2, including evaluation of the P-T limits resulting from the surveillance testing results of Capsule N for PINGP, Unit 1; Capsule N for PINGP, Unit 2; and update to the methodology of WCAP-14040-NP-A, Revision 4.

As described in RIS 2014-11, the beltline definition in 10 CFR Part 50, Appendix G is applicable to all reactor vessel ferritic materials with projected neutron fluence values greater than 1 x 1017 n/cm2 (E > 1 MeV), and this fluence threshold remains applicable for the design life as well as throughout the licensed operating period. The PINGP, Unit 1, beltline materials include Nozzle (Upper) Shell Forging B, Intermediate Shell Forging C, Lower Shell Forging D, Nozzle Shell Forging B to Intermediate Shell Forging C Circumferential Weld W2, and Intermediate Shell Forging C to Lower Shell Forging D Circumferential Weld W3. The Prairie Island, Unit 2 beltline materials include Upper Shell Forging B, Intermediate Shell Forging C, Lower Shell Forging D, Upper Shell Forging B to Intermediate Shell Forging C Circumferential Weld W2, and Intermediate Shell Forging C to Lower Shell Forging D Circumferential Weld W3.

The NRC staff confirmed the best-estimate weight percent values of copper (Cu) and nickel (Ni),

initial RTNDT and initial upper shelf energy (USE) values for the reactor vessel beltline materials provided in Table 3-1 and Table 3-2 of WCAP-18746-NP, Revision 2, which is the Enclosure to the March 28, 2023, supplement, for PINGP, Unit 1 and Unit 2, respectively.

3.4 Surveillance Data The summary technical report for the most recently withdrawn and tested surveillance capsule (Capsule N) for PINGP, Unit 1, was documented in WCAP-18660-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island, Unit 1 Reactor Vessel Radiation Surveillance Program, (ML22067A147). For PINGP, Unit 2, the summary technical report for the most recently withdrawn and tested capsule (Capsule N) was documented in WCAP-18795, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island, Unit 2 Reactor Vessel Radiation Surveillance Program, (ML23075A344). These reports were reviewed by the NRC staff, as documented in a staff assessment dated September 26, 2023 (ML23265A253), and September 29, 2023 (ML23270B902), for Capsule N at PINGP, Unit 1 and Unit 2, respectively.

Position 2.1 of RG 1.99, Revision 2 describes a methodology for incorporating data from a reactor vessel material surveillance program in calculating the chemistry factor for that beltline material. The surveillance capsule base metal material for PINGP, Unit 1, is from Intermediate Shell Forging C; the surveillance weld material was fabricated from weld heat #1752, applicable to Intermediate Shell Forging C to Lower Shell Forging D Circumferential Weld W3. As described in WCAP-18660-NP, Revision 0 and the associated NRC staff assessment, the staff confirmed that the surveillance forging data and weld data were non-credible. Per Position 2.1 of RG 1.99, Revision 2, when the surveillance data is credible, the margin in determining the ART may be reduced. The NRC staff confirmed that since the surveillance data for PINGP, Unit 1, was not found to be credible, the margin term was not reduced.

The surveillance capsule weld material for PINGP, Unit 2, is from Lower Shell Forging D; the surveillance weld material was fabricated from weld heat #2721, applicable to Intermediate Shell Forging C to Lower Shell Forging D Circumferential Weld W3. As described in WCAP-18794-NP, Revision 0 and the associated NRC staff assessment, the staff confirmed that the surveillance forging data and weld data were non-credible. Per Position 2.1 of RG 1.99, Revision 2, as described above, the NRC staff confirmed that since the surveillance data for PINGP, Unit 2, was not found to be credible, the margin term was not reduced.

3.5 Chemistry Factor The surveillance data for PINGP, Unit 1 and Unit 2, are summarized in Table 4-1 and Table 4-2, respectively, of WCAP-18746-NP, Revision 2, which is the Enclosure to the March 28, 2023, supplement. The NRC staff confirmed that the licensee calculated chemistry factors (CFs) of the surveillance materials following the methodology in Regulatory Guide 1.99, Revision 2. The CFs for the surveillance materials are provided in Table 5-1 and Table 5-2 for PINGP, Unit 1 and Unit 2, respectively. The NRC staff confirmed that the CFs included surveillance data for the latest surveillance capsules.

3.6 Adjusted Reference Temperature The NRC staff verified that the 54 EFPY ART values for each reactor vessel beltline material in Tables 7-3, 7-4, 7-5 and 7-6 of WCAP-18746-NP, Revision 2, were calculated following the methodology in RG 1.99, Revision 2. Per these Tables, the limiting ART value for PINGP, Unit 1, is Intermediate Shell to Lower Shell Circumferential Weld W3 (Heat #1752). However, the applied stress and resulting stress intensity factor at the postulated flaw location in the circumferential weld are lower than at the postulated flaw located in the most limiting axial weld.

With consideration of the stress contribution, the limiting ART material does not necessarily define the most limiting P-T limits. The NRC staff confirmed that the most limiting P-T curve for PINGP, Unit 1, with an axially oriented flaw, is Intermediate Shell Forging C. The 1/4T and 3/4T ART values for the PINGP, Unit 1, Intermediate Shell Forging C at 54 EFPY are 140.7°F and 126.9°F, respectively. These values are less than the ART values of 154°F and 136°F, respectively, used to generate the P-T curves in WCAP-14780 and the current PTLR. PINGP, Unit 1 and Unit 2, utilize one set of P-T limit curves. By letter dated November 7, 2002 (ML023230354), the licensee submitted Revisions 2 and 3 to the PTLR which states, in part, that: The results of the analysis of the Units 1 and 2 reactor vessel material surveillance capsule tests show that the limitations for Unit 1 are the most restrictive and conservative. For simplicity these results have been applied to both units.

Due to the stress contribution, a material with a circumferential flaw may not produce the most limiting P-T limit curves. This is illustrated in the figures in Section 9 of WCAP-18746-NP, Revision 2. The ART values considered in the P-T limit curves development are summarized in Table 8-1 of WCAP-18746-NP, Revision 2. Therefore, as discussed above, the NRC staff verified the ART values for the reactor vessel beltline materials to be acceptable because the licensee followed the guidance of RG 1.99, Rev. 2.

3.7 Pressure-Temperature Limits The NRC staff verified that the P-T limit curves for 54 EFPY were calculated for the reactor vessel beltline region following the approved methodology presented in WCAP-14040-A, Revision 4. NRC RIS 2014-11 states that:

In determining P-T limits, reactor vessel materials with the highest reference temperature may not always produce the most limiting P-T limits because the consideration of stress levels from structural discontinuities (such as nozzles) may produce a lower allowable pressure. All addressees should ensure that P-T limits (including those implementing NRC-approved PTLR methodologies) sufficiently address all ferritic materials of the reactor vessel, including the impact of structural discontinuities, and address the impact of neutron fluence accumulation in accordance with the requirements of 10 CFR Part 50, Appendix G.

For the inlet and outlet nozzles, the licensee cited PWROG-15109-NP-A, Revision 0 (ML20024E573) as an acceptable basis for satisfying the fracture toughness requirements in Appendix G to 10 CFR Part 50 for nozzles which have neutron fluence exposures at the nozzle corners less than the screening criterion of 4.28 x 1017 n/cm2. The NRC staff verified that the PINGP, Unit 1 and Unit 2, inlet and outlet nozzles have neutron fluence projections below this screening criterion at 54 EFPY, therefore the staff concludes that PWROG-15109NP-A is an acceptable basis for meeting the 10 CFR Part 50, Appendix G, P-T limits requirements for the inlet and outlet nozzles.

PINGP, Unit 1 and Unit 2, conservatively utilize the same P-T curves based upon the most limiting data and analysis for both units. The currently approved P-T limit curves for PINGP, Unit 1 and Unit 2, contained in the PTLR were generated based on the methodology in WCAP-14040-A, Revision 2, as documented in WCAP-14780. The updated P-T limit curves were generated based on the methodology in WCAP-14040-A, Revision 4 and incorporate recent surveillance data. The NRC staff confirmed that the resulting updated P-T limit curves are bounded by the current PTLR curves. This is illustrated in Figure 9-1 and Figure 9-2 of WCAP-18746-NP, Revision 2. Since the updated P-T limit curves are bounded by the current PTLR curves, the NRC staff finds that the current PTLR curves remain applicable through 54 EFPY.

3.8 Upper-Shelf Energy Guidance in 10 CFR Part 50, Appendix G requires the maintenance of an Charpy upper-shelf energy for reactor vessel beltline materials throughout the life of the vessel of no less than 50 ft-lb (68 J), unless it is demonstrated in a manner approved by the Director, Office of Nuclear Reactor Regulation, that lower values of Charpy upper-shelf energy will provide margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASME Code. The NRC staff verified that projected upper-shelf energy values were calculated in accordance with the methodology in RG 1.99, Revision 2. Therefore, the NRC staff finds that all reactor vessel beltline materials for PINGP, Unit 1 and Unit 2, are projected to remain above the 50 ft-lb criterion through 54 EFPY (PEO).

The NRC staff assessed of the upper-shelf energy evaluation in Appendix C in WCAP-18746-NP, Revision 2, and confirmed that all reactor vessel materials for PINGP, Unit 1 and Unit 2, are projected to remain above the 50 ft-lb screening criterion per 10 CFR Part 50 Appendix G through 54 EFPY.

3.9 Pressurized Thermal Shock Appendix D of WCAP-18746-NP contains updated pressurized thermal shock (PTS) projections at 54 EFPY for the reactor vessel beltline materials. The NRC staff verified that the RTPTS values for PINGP, Unit 1 and Unit 2, were calculated in accordance with the requirements of 10 CFR 50.61 and the guidance of RG 1.99, Revision 2. Using Position 2.1 in RG 1.99, Revision 2, the limiting RTPTS value for base metal at 54 EFPY is 146.3°F, which corresponds to PINGP, Unit 1, Intermediate Shell Forging C. Using Position 2.1 in RG 1.99, Revision 2, the limiting RTPTS value for circumferentially oriented welds at 54 EFPY is 183.9°F, which corresponds to the Unit 1 Intermediate to Lower Shell Circumferential Weld W3, Heat #1752.

These values are conservative for Unit 2 materials. Therefore, the NRC staff finds that the PINGP, Unit 1 and Unit 2, reactor vessel materials are below the RTPTS screening criteria of 270°F for base metal and/or longitudinal welds, and 300°F for circumferentially oriented welds through 54 EFPY and no further action regarding PTS is necessary.

The PTS evaluation in WCAP-18746-NP, Appendix D, demonstrated that all reactor vessel beltline materials have projected RTPTS values below the screening criteria set forth in 10 CFR 50.61 through 54 EFPY. Therefore, the NRC staff concludes that the requirements in 10 CFR 50.61 have been met for the PEO for PINGP, Unit 1 and Unit 2.

3.10 Neutron Fluence Methodology In its submittal, the licensee stated that the neutron fluence calculations were performed in a manner consistent with the guidance described in RG 1.190. The NRC staff reviewed the modeling approach described by the licensee and determined that the neutron fluence calculations are consistent with RG 1.190 guidance. RAPTOR-M3G is a three-dimensional discrete ordinates radiation transport code that approximates a solution to the Boltzmann transport equation. The code methodology and qualification data are documented in the topical report WCAP-18124-NP-A, Revision 0, Supplement 1-P and WCAP-18124-NP-A, Revision 0, Supplement 1-NP, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials (ML22153A139). As indicated in the NRC safety evaluation bound into WCAP-18124-NP-A and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0, the NRC staff has found the calculational fluence methodology described in that topical report, as supplemented, acceptable. This topical report as supplemented has been approved by NRC staff for calculation of reactor pressure vessel neutron fluence provided that limitations and conditions in the topical report are met.

The first limitation and condition states applicability of WCAP-18124-NP, Revision 0, is limited to the RPV region near the active height of the core based on the uncertainty analysis performed and the measurement data provided. In this regard, the SE on WCAP-18124-NP, Revision 0, indicates that additional justification should be provided via additional benchmarking, fluence sensitivity analysis to response parameters of interest (e.g., pressure -temperature limits, material stress/strain), margin assessment, or a combination thereof, for applications of the method to components including, but not limited to, the RPV upper circumferential weld and reactor coolant system inlet and outlet nozzles and reactor vessel internal components.

As described in RIS 2014-11, structural discontinuities in the RPV can induce additional stresses to those experienced in the traditional beltline. Materials that are not adjacent to the active core yet are predicted to accrue fluence levels greater than 1.0 x 1017 n/cm2 (E > 1.0 MeV) are now commonly referred to as extended beltline materials.

WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials, provided the justification necessary to narrow limitation and condition 1 and allow licensees to apply the RAPTOR-M3G method in the extended beltline regions of RPVs on a generic basis. The NRC staff has determined that the fluence methods and qualifications described in WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A are acceptable for referencing in licensing applications. Therefore, the NRC staff determined that the justification for application of the RAPTOR-M3G fluence methodology in extended beltline regions is acceptable, and that the first limitation and condition in the NRC staff SE for WCAP-18124-NP, Revision 0, is satisfied.

The second limitation and condition states least-squares adjustment is acceptable if the adjustments to the measured to calculated (M/C) ratios and to the calculated spectra values are within the assigned uncertainties of the calculated spectra, the dosimetry measured reaction rates, and the dosimetry reaction cross sections. Should this not be the case, the user should re-examine both measured and calculated values for possible errors. If errors cannot be found, the particular values causing the inconsistency should be disqualified.

The licensee stated that the second limitation and condition does not apply because the least-squares procedures were not used to adjust the calculated fast neutron fluence values for RPV materials evaluated in the reactor vessel integrity analysis. Since the licensee did not use the FERRET least-squares adjustment methods in its estimation of the fluence values in the reactor vessel integrity analysis, the NRC staff determined that this limitation does not apply, and the licensee has therefore addressed the limitation acceptably.

Based on the considerations discussed above, NRC staff has determined that use of WCAP-18124-NP-A, Revision 0, Supplement 1-P and WCAP-18124-NP-A, Revision 0, Supplement 1-NP, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials to calculate neutron fluence in the RPV for the purpose of determining RCS P-T limits for Prairie Island, Units 1 and 2, up to 54 effective full-power years is acceptable. WCAP-18124-NP-A, Revision 0, Supplement 1-P and WCAP-18124-NP-A, Revision 0, Supplement 1-NP, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials have been generically approved by NRC staff for this purpose. The licensee has acceptably addressed all the limitations and conditions discussed in the NRC staff SE of WCAP-18124-NP-A, revision 0. Further, the NRC staff finds that the neutron fluence calculation is consistent with the guidance in RG 1.190 and therefore meets the provisions of the Prairie Island PDC corresponding to draft GDC 9, 34 and 35, published in July 1967.

3.11 Low Temperature Overpressure Protection System Setpoint The LTOPS provides pressure relief capability to the RCS during low temperature operation to mitigate the potential challenge to vessel integrity. In enclosure 5 (proprietary version) to the LAR, the licensee provided an analysis for the LTOPS for potential RCS overpressure transients that may occur at low temperature conditions. The licensee stated that the LTOPS Power Operated Relief Valves (PORVs) setpoints are selected to meet the TS LCOs 3.4.12 and 3.4.13 such that the peak pressure during design basis mass injection (MI) and heat injection (HI) transients will not exceed the P-T limits.

As discussed in section 3.2 of this SE, the analytical method proposed by the licensee to be used to develop LTOPS setpoints for PINGP is described in the NRC-approved Topical Report WCAP-14040-NP-A, Revision 4. Because the methodology is approved and accepted for use by the NRC, it provides an acceptable means of satisfying 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, which governs the development of P-T limits and LTOPS setpoints. As explained in section 3.2 of this SE, the NRC staff finds that the licensee has adequately addressed the three conditions required for the use of WCAP-14040-A, Revision 4, and that use of the new methodology will provide the appropriate operating curves and limits for operation to the end of the renewed facility operating licenses at the licensed power level.

Therefore, the NRC staff has determined that use of WCAP-14040-A, Revision 4, to develop LTOPS setpoints is acceptable.

3.12 Evaluation of Changes to Technical Specifications In sections 3.2 and 3.10 of this SE, the staff evaluated the NRC approved methodologies proposed for use by the licensee to be applicable to PINGP, Units 1 and 2, and explained that the licensee satisfactorily addressed the limitations and conditions established for these methodologies, and that they are therefore, acceptable for determining the pressure and temperature limits and LTOPS setpoints for PINGPs Units 1 and 2.

In an August 4, 2011, letter from the NRC to the Technical Specifications Task Force (TSTF)

(ML110660285), the NRC informed the TSTF of the following:

The NRC staff will no longer accept license amendment requests (LARs) to implement technical specification (TS) changes in accordance with Traveler TSTF-363. The standard TS (STS) in NUREGs-1430, -1431, -1432, -1433, and -1434 will be revised in Revision 4 to reflect the TS the way they were prior to approval of TSTF-363.

The NRC staff will continue to accept LARs to implement the TS changes in accordance with Travelers TSTF-408 [Relocation of LTOP Enable Temperature and PORV Lift Setting to the PTLR] and TSTF-419 [Revise PTLR Definition and References in ISTS 5.6.6 RCS PTLR] with modification. Currently the STS show the full topical report or methodology citation as located in the PTLR. In order for NRC staff to approve LARs for these two travelers, the full topical report or methodology citation will need to be included in the TS, not in the PTLR.

Since the licensees adoption of specific reports listed in the proposed changes are identified by the report number, title, revision, and date, the changes are consistent with the NRC staffs position in the NRC August 2011 letter, as discussed above.

PINGP, Units 1 and 2, have adopted the standard TSs, which provide guidelines for a licensee to list the analytical methods used to determine RCS P-T limits. Accordingly, the NRC staff compared the proposed changes with the guidance in the latest revision (Revision 5) of applicable Standard Technical Specifications (STS), NUREG-1431, Standard Technical Specifications, Westinghouse Plants, Revision 5.0, Volume 1, dated September 2021 (ML21259A155) and the August 4, 2011, letter to the TSTF. Because the licensee is adopting WCAP-18124-NP-A, Revision 0, for the PINGP units as the neutron fluence calculational methodology for the evaluation of reactor vessel specimens to support the determination of the RCS P-T limits, and WCAP-14040-A, Rev. 4 as the methodology for calculating LTOPS setpoints and RCS heatup and cooldown limit curves, it is appropriate for these methodologies to be added to TS 5.6.6.b. Further, the licensee is adopting WCAP-18124-NP-A Rev. 0 Supplement 1-NP-A, Rev. 0 to adjust the applicability of the neutron fluence calculational methodology for the region of the reactor pressure vessel (RPV) near the active core. The licensee will use these methodologies for PINGP, Units 1 and 2, as part of the neutron fluence calculational methodology for the evaluation of reactor vessel specimens to support the determination of the RCS P-T limits: therefore, it is appropriate that they be added to TS 5.6.6.b for each unit. For each of the methodologies discussed above, the licensee has proposed adding the complete methodology citation in TS 5.6.6. This is consistent with the guidance in the STS and in the August 4, 2011, letter. Therefore, these changes are acceptable.

The NRC staff reviewed the proposed TS changes for technical clarity and consistency with the requirements for customary terminology and formatting. The staff finds that the proposed TS changes are consistent with Chapter 16.0 of the SRP and NUREG-1431. In addition, 10 CFR 50.36(c)(5) requires that TS include administrative controls. The NRC staffs review of TS 5.6.6, as amended, will continue to include provisions and reporting requirements necessary to assure operation of the facility in a safe manner. Based on this, the amended TS 5.6.6 will continue to meet 10 CFR 50.36(c)(5) and is acceptable.

3.13 Technical Conclusion The NRC staff has determined that use of WCAP-18124-NP-A, Revision 0, Supplement 1-P and WCAP-18124-NP-A, Revision 0, Supplement 1-NP, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials to calculate neutron fluence in the RPV for the purpose of determining RCS P-T limits for Prairie Island, Units 1 and 2, up to 54 effective full-power years is acceptable. Specifically, WCAP-18124-NP-A, Revision 0, Supplement 1-P and WCAP-18124-NP-A, Revision 0, Supplement 1-NP, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials have been generically approved by NRC staff for this purpose. The licensee has addressed all the limitations and conditions discussed in the NRC staff SE of WCAP-18124-NP-A, revision 0.

Further, the NRC staff finds that the neutron fluence calculation is consistent with the guidance in RG 1.190 and therefore meets the provisions of the Prairie Island PDC corresponding to draft GDC 9, 34 and 35, published in July 1967. In addition, the NRC staff has determined that use of WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves to develop P-T limits and LTOPS setpoints is acceptable given that the licensee has addressed all limitations and conditions to demonstrate applicability of this topical to PINGP.

The NRC staff also finds that the updated P-T limit curves for PINGP, Unit 1 and Unit 2, meet the criteria for using WCAP-14040, Revision 4. Development of the P-T limit curves incorporated results for the test results from the reactor vessel material surveillance program, including the most recently withdrawn and tested Capsule N for PINGP, Unit 1 and Unit 2. As discussed above, the NRC staff concluded that the updated P-T limit curves are bounded by the currently approved P-T limit curves in the PTLR, thus the currently approved P-T limit curves in the PTLR remain applicable through 54 EFPY.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations on, the Minnesota State official was notified of the proposed issuance of the amendments on December 8, 2023. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change the requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or change the surveillance requirements. 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 and there has been no public comment on such finding (88 FR 14184). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

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

Principal Contributors: Carolyn Fairbanks, NRR Christopher Jackson, NRR Ravinder Grover, NRR Date of Issuance: January 17, 2024

ML23356A003 NRR-058 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DNRL/NVIB/BC NRR/DSS/SNSB/BC NAME BBallard SRohrer (SLent for)

ABuford (JTsao for)

PSahd DATE 12/21/2023 12/22/23 12/6/2023 10/23/2023 OFFICE NRR/DSS/STSB/BC (A) OGC - NLO NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME SMehta RWeisman JWhited BBallard DATE 11/5/2023 1/12/2024 1/17/2023 1/17/2024