ML25111A239

Issuance of Amendment Nos. 247 and 235 Revise the TS Definition of Reactor Trip System (RTS) Response Time and Apply Response Time Testing to RTS Trip Functions with Time Delay Assumptions

ML25111A239
Person / Time
Site:	Prairie Island
Issue date:	06/09/2025
From:	Wetzel B Plant Licensing Branch III
To:	Northern States Power Co
Ballard, Brent
References
EPID L-2025-LLA-0067
Download: ML25111A239 (29)

Text

June 9, 2025 Werner Paulhardt 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. 247 and 235 RE: REVISE THE TECHNICAL SPECIFICATION DEFINITION OF REACTOR TRIP SYSTEM (RTS) RESPONSE TIME AND APPLY RESPONSE TIME TESTING TO RTS TRIP FUNCTIONS WITH TIME DELAY ASSUMPTIONS (EPID L-2024-LLA-0067)

Dear Mr. Paulhardt:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 247 to Renewed Facility Operating License No. DPR-42 and Amendment No. 235 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 June 3, 2024, as supplemented by letter dated December 9, 2024.

The amendments revise the definition of REACTOR TRIP SYSTEM (RTS) RESPONSE TIME used in TS Surveillance Requirement (SR) 3.3.1.16 to allow allocation of response times in lieu of testing using the methodology proposed in Enclosure (2) of the June 3, 2024, letter.

Additionally, TS Table 3.3.1-1 is revised to apply SR 3.3.1.16 to the RTS trip functions for which the accident analyses include assumptions about time delays.

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

Sincerely,

/RA/

Beth Wetzel, 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. 247 to DPR-42

2. Amendment No. 235 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. 247 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 June 3, 2024, as supplemented by letter dated December 9, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's 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. 247, 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 Ilka Berrios, Acting 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: June 9, 2025 ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2025.06.09 15:34:58 -04'00'

NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 235 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 June 3, 2024, as supplemented by letter dated December 9, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's 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. 235 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 Ilka Berrios, Acting 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: June 9, 2025 ILKA BERRIOS Digitally signed by ILKA BERRIOS Date: 2025.06.09 15:35:49 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NOS. 247 AND 235 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 1.1-5 1.1-5 3.3.1-20 3.3.1-20 3.3.1-21 3.3.1-21 3.3.1-22 3.3.1-22 Renewed Operating License No. DPR-42 Amendment No. 247 (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. 247, 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. 235 (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. 235, 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

Definitions 1.1 Prairie Island Unit 1 - Amendment No. 247 Units 1 and 2 1.1-5 Unit 2 - Amendment No. 235 1.1 Definitions (continued)

PRESSURE AND The PTLR is the unit specific document that provides the reactor TEMPERATURE vessel pressure and temperature limits, including heatup and LIMITS cooldown rates, and the OPPS arming temperature for the current REPORT reactor vessel fluence period. These pressure and temperature limits (PTLR) shall be determined for each fluence period in accordance with Specification 5.6.6. Plant operation within these operating limits is addressed in LCO 3.4.3, RCS Pressure and Temperature (P/T)

Limits, LCO 3.4.12, Low Temperature Overpressure Protection (LTOP) -Reactor Coolant System Cold Leg Temperature (RCSCLT)

> Safety Injection (SI) Pump Disable Temperature, and LCO 3.4.13, Low Temperature Overpressure Protection (LTOP)

- Reactor Coolant System Cold Leg Temperature (RCSCLT)

< Safety Injection (SI) Pump Disable Temperature.

QUADRANT QPTR shall be the ratio of the maximum upper excore detector POWER TILT calibrated output to the average of the upper excore detector RATIO (QPTR) calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED RTP shall be a total reactor core heat transfer rate to the reactor THERMAL coolant of 1677 MWt.

POWER (RTP)

REACTOR The RTS RESPONSE TIME shall be that time interval from when TRIP the monitored parameter exceeds its RTS trip setpoint at the channel SYSTEM (RTS) sensor output until opening of a reactor trip breaker. The response

RESPONSE

time may be measured by means of any series of sequential, TIME overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components have been evaluated in accordance with an NRC approved methodology.

RTS Instrumentation 3.3.1 Prairie Island Unit 1 - Amendment No. 247 Units 1 and 2 3.3.1-20 Unit 2 - Amendment No. 235 Table 3.3.1-1 (page 1 of 8)

Reactor Trip System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

1.

Manual Reactor Trip 1, 2 3(a), 4(a), 5(a) 2 2

B C

SR 3.3.1.14 SR 3.3.1.14 NA NA

2.

Power Range Neutron Flux

a.

High

b.

Low

3.

Power Range Neutron Flux Rate

a. High Positive Rate

b. High Negative Rate

4.

Intermediate Range Neutron Flux 1, 2 1(b), 2 1, 2 1, 2 1(b), 2(c) 4 4

4 4

2 D

D D

D F, G SR 3.3.1.1 SR 3.3.1.2 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.16 SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.16 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11

< 110% RTP

< 40% RTP

< 6% RTP with time constant

> 2 sec

< 8% RTP with time constant

> 2 sec

< 40% RTP (a)

With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(b)

Below the P-10 (Power Range Neutron Flux) interlocks.

(c)

Above the P-6 (Intermediate Range Neutron Flux) interlocks.

RTS Instrumentation 3.3.1 Prairie Island Unit 1 - Amendment No. 247 Units 1 and 2 3.3.1-21 Unit 2 - Amendment No. 235 Table 3.3.1-1 (page 2 of 8)

Reactor Trip System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

5.

Source Range Neutron Flux 2(d) 3(a), 4(a), 5(a) 2 2

H, I I, J SR 3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.11

< 1.0E6 cps

< 1.0E6 cps

6. Overtemperature T 1, 2 4

E SR 3.3.1.1 SR 3.3.1.3 SR 3.3.1.6 SR 3.3.1.7 SR 3.3.1.12 SR 3.3.1.16 Refer to Note 1 (Page 3.3.1-23)

7. Overpower T

8.

Pressurizer Pressure

a.

Low

b.

High 1, 2 1(e) 1, 2 4

4 3

E K

E SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.12 SR 3.3.1.16 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16 Refer to Note 2 (Page 3.3.1-24)

> 1845 psig

< 2400 psig (a)

With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(d)

Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(e)

Above the P-7 (Low Power Reactor Trips Block) interlock.

RTS Instrumentation 3.3.1 Prairie Island Unit 1 - Amendment No. 247 Units 1 and 2 3.3.1-22 Unit 2 - Amendment No. 235 Table 3.3.1-1 (page 3 of 8)

Reactor Trip System Instrumentation FUNCTION APPLICABLE MODES OR OTHER SPECIFIED CONDITIONS REQUIRED CHANNELS CONDITIONS SURVEILLANCE REQUIREMENTS ALLOWABLE VALUE

9.

Pressurizer Water Level - High 1(e) 3 K

SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10

< 90%

10. Reactor Coolant Flow-Low 1(f) 3 per loop K

SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

> 91%

11. Loss of Reactor Coolant Pump (RCP)

a.

RCP Breaker Open

b.

Under-frequency 4 kV Buses 11 and 12 (21 and 22)

12. Undervoltage on 4 kV Buses 11 and 12 (21 and

22)

13. Steam Generator (SG)

Water Level -

Low Low 1(f) 1(f) 1(e) 1, 2 1 per RCP 2 per bus 2 per bus 3 per SG M

L L

E SR 3.3.1.14 SR 3.3.1.9 SR 3.3.1.10 SR 3.3.1.9 SR 3.3.1.10 SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16 NA

> 58.2 Hz

> 76% rated bus voltage

> 11.3%

(e)

Above the P-7 (Low Power Reactor Trips Block) interlock.

(f)

Above the P-8 (Power Range Neutron Flux) or P-7 (Low Power Reactor Trips Block) interlocks.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 247 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-42 AND AMENDMENT NO. 235 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-60 NORTHERN STATES POWER COMPANY - MINNESOTA PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306

1.0 INTRODUCTION

By application dated June 3, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24155A220), as supplemented by letter dated December 9, 2024 (ADAMS ML24345A206), Northen States Power Company, a Minnesota corporation (hereafter NSPM or the licensee), submitted a license amendment request (LAR) to change the Technical Specifications (TS) for the Prairie Island Nuclear Generating Plant, Units 1 and 2 (PINGP). The proposed changes would revise the definition of REACTOR TRIP SYSTEM (RTS)

RESPONSE TIME used in TS Surveillance Requirement (SR) 3.3.1.16 to allow allocation of response times to replace testing using the proposed methodology. The LAR would also change TS Table 3.3.1-1 to apply SR 3.3.1.16 to the RTS trip functions for which the accident analyses include assumptions about time delay.

The supplemental letter dated December 9, 2024, 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) staff's original proposed no significant hazards consideration determination as published in the Federal Register on August 6, 2024 (89 FR 63992).

The proposed changes would revise the definition of REACTOR TRIP SYSTEM (RTS)

RESPONSE TIME used in TS Surveillance Requirement (SR) 3.3.1.16 to allow allocation of response times in lieu of testing using the methodology proposed in Enclosure (2) of the June 3, 2024, letter. Additionally, TS Table 3.3.1-1 is revised to apply SR 3.3.1.16 to the RTS trip functions for which the accident analyses include assumptions about time delays.

2.0 REGULATORY EVALUATION

The NRC staff relies on the following NRC regulations to evaluate the information related to the proposed changes to TS Table 3.3.1-1and the response times assumed in the accident analyses:

Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36 specifies the contents of the Technical Specifications, including (1) safety limits, (2) limiting conditions for operation, (3) surveillance requirements, (4) design features, and (5) administration controls. 10 CFR 50.36(c)(2), "Limiting conditions for operation," states that "limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met."

10 CFR 50.36(c)(2)(ii) requires that a technical specification limiting condition for operation (LCO) of a nuclear reactor must be established for each item meeting one or more of the following criteria:

Criterion 1 - Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2 - A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3 - A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4 - A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

10 CFR 50.36(c)(3), Surveillance requirements, states that 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 the safety limits, and the limiting conditions for operation will be met.

PINGP was designed and constructed to comply with the intent of the Atomic Energy Commission (AEC) General Design Criteria (GDC) for Nuclear Power Plant Construction Permits, as proposed on July 10, 1967. Therefore, the PINGP Licensing Basis requires conformance to the AEC GDC, as reflected in the PINGP Updated Safety Analysis Report (USAR). The following AEC GDC is applicable to the proposed changes:

GDC 14, Core Protection Systems - Core protection systems, together with associated equipment, shall be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits.

3.0 TECHNICAL EVALUATION

3.1 Changes to Reactor Trip System Response Time Definition PINPG Units 1 and 2 LCO 3.3.1, requires the RTS instrumentation for each Function in TS Table 3.3.1-1 Reactor Trip System Instrumentation, to be OPERABLE. To assure the LCO is met, SR 3.3.1.16 requires the licensee to verify that RTS RESPONSE TIMES are within limits.

Section 1.1 of the TS definitions for RTS RESPONSE TIME states acceptable means to measure each response time.

The licensee proposed revising the TS 1.1 definition of REACTOR TRIP SYSTEM (RTS)

RESPONSE TIME to allow allocation of response times using an approved methodology in lieu of testing by adding, In lieu of measurement, response time may be verified for selected components provided that the components have been evaluated in accordance with an NRC approved methodology. The licensee proposes to use the previously approved TSTF-569, Revision 2, Methodology 2, as the approved methodology for allocating response times to the PINGP NIS and RPS.

The NRC staff reviewed the proposed change to the TS 1.1 definition of RTS response time and determined that it is acceptable because the proposed change is consistent with TSTF-569, the staff did not identify any material differences in the relevant technical specifications, and 10 CFR 50.36(c)(3) will continue to be met.

3.2 Proposed Methodologies for Allocating Response Times in Lieu of Testing The reactor trip system (RTS) at PINGP Units 1 and 2 initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and reactor coolant system (RCS) pressure boundary during anticipated operational occurrences (AOO) and to assist the Engineered Safety Features Systems in mitigating accidents. The RTS at PINGP Units 1 and 2 includes the following three subsystems: the nuclear instrumentation system (NIS), the relay protection system (RPS), and the process protection system (PPS).

The licensee proposed to revise the definition of RTS response time used in TS SR 3.3.1.16 for PINGP Units 1 and 2 to allow allocation of response times instead of periodic response time testing (RTT) for the RTS. However, allocation of response times to selected system components or modules requires a methodology approved by the NRC. In the LAR, the licensee first proposed to use the Technical Specifications Task Force (TSTF)-569, Revision 2,, Methodology 2 for allocating response times for the NIS and RPS subsystems and then use a proposed methodology in the LAR for the PPS subsystem.

TSTF-569, Revision 2, Attachment 1, Methodology 2 is based on Westinghouse Owners Group Topical Report (TR) WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests", which provides a technical justification for deletion of periodic RTT of the electronic signal processing hardware between the primary sensor and the final actuated device. This Topical Report with its associated TSTF-569, Revision 2 including Methodology 2 was previously reviewed and approved by the NRC.

PINGP Units 1 and 2 currently utilize the Westinghouse DB-50 reactor trip breakers. Like the approved TR WCAP-14036-P-A, Revision 1, the reactor trip breakers for the PINGP Units 1 and 2 RTS trip functions are not included in the methodology proposed in this LAR, so the reactor trip breakers for the PINGP Units 1 and 2 RTS are still subject to the periodic RTT as required in the TS.

For PINGP Units 1 and 2, the licensee stated in the LAR that the NIS and RPS subsystems were built with a Westinghouse NIS and relay reactor protection system (RRPS), respectively.

The NIS and RRPS signal processing hardware evaluated in the TR WCAP-14036-P-A, Revision 1 applies to the PINGP encore nuclear instrumentation system and associated solid state and relay trip logic circuitry up to the slave relay output. So, the PINGP Units 1 and 2 NIS and RPS subsystems are directly applicable to components described and assessed in the TR WCAP-14036-P-A, Revision 1. In addition, the NRC assessed and approved methodologies including Methodology 2 in TSTF-569, Revision 2 and its associated TR WCAP-14036-P-A, Revision 1 for allocating response times in lieu of periodic testing for applicable components.

Hence, the NRC staff found that the proposed change in the LAR is acceptable to use allocation of response times based on TSTF-569, Revision 2, Attachment 1, Methodology 2 to replace the RTT for the NIS and RPS subsystems at PINGP Units 1 and 2.

For PINGP Units 1 and 2, the original PPS sub-system used the Foxboro H-Line equipment.

The H-Line equipment was not included and assessed in TR WCAP-14036-P-A, Revision 1.

Therefore, the approved methodologies in TSTF-569, Revision 2 and its associated TR WCAP-14036-P-A, Revision 1 could not be directly applied to allocate response times to replace its periodic RTT. However, the Foxboro H-Line equipment was replaced with Curtiss-Wright Scientech NUS modules, which were reverse engineered replacements for the original H-Line modules and are used to retrofit both Foxboro H-Line and Hagan 7100 modules by using similar circuitry. The Hagan 7100 line of equipment was included in the TR WCAP-14036-P-A, Revision

1. Hence, the licensee proposed in the LAR to use a methodology which is like the approved Methodology 2 described in TSTF-569, Revision 2, Attachment 1 to allocate response times for some RTS functions provided by the PINGP Units 1 and 2 PPS sub-system, instead of performing their periodic RTT.

The proposed methodology in the LAR for evaluating whether RTT can be replaced with allocation of response times for some RTS functions by the PPS sub-system consists of the following actions:(1) Analyze the system modules for their function in providing the protection function. (2) Perform a failure modes and effects analysis (FMEA) on the modules that perform a protection function to determine whether individual component degradation has no impact on the response time or whether the individual components may contribute to the system response time degradation. (3) If the individual component potentially impacts the system response time, perform testing to determine the magnitude of the response time degradation, or determine a bounding response time limit for the system or component if the individual component does not impact the system response time. The NRC staff found that the actions taken for the proposed methodology is consistent with Methodology 2 in the approved TSTF-569, Revision 2 and its associated TR WCAP-14036-P-A, Revision 1.

The NRC staff evaluated the above three actions taken for this LAR. For Action 1, the licensee reviewed necessary instrument block diagrams for the PPS sub-system at the PINGP and adequately identified modules that perform the RTS protection functions included in the LAR.

For Action 2, an FMEA on module or component types identified in Action 1 was performed by Curtiss-Wright Scientech, as contracted by the licensee. Like the TR WCAP-14036-P-A, Revision 1 the licensee also identified capacitors and resistors as the dominant response time sensitive components in the FMEA. In the FMEA, the impact on response time was evaluated if components fail or degrade. It is determined in the FMEA that many failures result in a change detectable during a channel calibration while others may affect the response time of the component. The licensee identified components affecting channel calibration, but not the response time in the FMEA. In the FMEA, the following specific components: mV/I Amplifier (RTL501), Time Domain Module (TMD500), Loop Power Supply (LPS500), Summator (MTH500), Function Generator (SGU501), and Comparator (SAM503, DAM503) are identified and analyzed with bounding response times in the FMEA.

However, in both the submitted FMEA as Attachment 6 and Enclosure 2 of the LAR the NRC staff found that there is a lack of bases to support the rising and falling response times for Baseline for the identified RTL501 Module. The NRC staff also did not find adequate information to substantiate the rising and falling response times for Worst-Case Failures for the identified modules RTL501, MTH500, SGU501, and SAM/DAM503. Therefore, the NRC staff issued a request for additional information (RAI) on November 12, 2024 (ML24330A050). In its RAI response dated December 9, 2024 (ML24345A206), the licensee provided the following additional information to support the response times used in the FMEA.

For Module RTL501, the licensee stated that the testing documented in Section 4.2 of in the LAR and summarized in its Conclusions Section 5.1 applies to a RTL500 mV/I amplifier. PINGP utilizes a RTL501 mV/I amplifier which is like the RTL500 with the exception of the absence of the R52/C22 filter present in the RTL500. The summary provided in Table 5-1 of Attachment 6 indicates that the baseline and worst-case failure columns were adjusted to eliminate the response time associated with the absence of the filter circuit. For the baseline response time of the RTL501, 18 msec was removed from the RTL500 baseline to account for the absence of the R52/C22 filter (Rising Time: 56 msec - 18 msec = 38 msec; Falling Time: 35.5 msec - 18 msec =17.5 msec). For the worst-case failure response time of the RTL501, 37 msec was removed from the RTL500 response time to account for the absence of the R52/C22 filter, and the measured additional response time (17 msec Rising Time and 13 msec Falling Time) due to failures of applicable components (C2, C4, C11, C15, C20 and C21) was added (Rising Time: 47 msec - 37 msec +17 msec = 27 msec; Falling Time: 49 msec - 37 msec + 13 msec = 25 msec). The total response time is the summation of the baseline and the worst-case failure columns (Rising Time: 38 msec + 27 msec = 65 msec; Falling Time: 17.5 msec + 25 msec = 42.5 msec).

For Module MTH500, the licensee mentioned that the testing documented in Section 4.3 of and summarized in its Conclusions Section 5.2 applies to a MTH500 summator.

Worst-case failure response times for the MTH500 were due to failures of applicable components (R8 and R7) which added 2.675 msec to the falling transient response and 2.215 msec to the rising transient response. In addition, failures of capacitors (C3.1, C54, C14 and C58) contributed an additional 4.735 msec to the rising response and 5.525 msec to the falling response. The times reported on Table 5-1 in Attachment 6 for worst-case failures is the sum of the above (Rising Time 2.215 msec + 4.735 msec = 6.950 msec, Falling Time 2.675 msec +

5.525 msec = 8.200 msec). The total response time is the summation of the baseline and the worst-case failure columns (Rising Time 2.765 msec + 6.950 msec = 9.715 msec, Falling Time 3.215 msec + 8.200 msec = 11.415 msec).

For Module SGU501, the licensee stated that the testing documented in Section 4.4 of and summarized in its Conclusions Section 5.3 applies to a SGU501 function generator. Worst-case failure response times for the SGU501 were due to failures of applicable components (F1.1, R14.1, R8 and RN6) which added 9.8 msec to the falling transient response and 49.6 msec to the rising transient response. In addition, failures of capacitors (VR1.1, R15.1, C3.1, C103, C403, C32, C33, C6 and C29) contributed an additional 49.8 msec to the rising response and 13.6 msec to the falling response. The times reported on Table 5-1 in Attachment 6 for worst-case failures is the sum of the above (Rising Time 49.6 msec + 49.8 msec = 99.4 msec, Falling Time 9.8 msec + 13.6 msec = 23.4 msec). The total response time is the summation of the baseline and the worst-case failure columns (Rising Time 41.0 msec + 99.4 msec = 140.4 msec, Falling Time 14.8 msec + 23.4 msec = 38.2 msec).

For Module SAM/DAM503, the licensee indicated that the testing documented in Section 4.5 of and summarized in its Conclusions Section 5.4 applies to a SAM/DAM503 comparator. Worst-case failure response times for the SAM/DAM503 were due to failures of applicable components (R104, R105, R107 and R108) which added 1 sec to the falling transient response and 28 sec to the rising transient response. In addition, failures of capacitors (C106) contributed an additional 45.5 sec to the rising response and 0.7 sec to the falling response. The times reported on Table 5-1 in Attachment 6 for worst case failures is the sum of the above (Rising Time 28 sec + 45.5 sec = 73.5 sec, Falling time 1 sec + 0.7 sec

1.7 sec). The total response time is the summation of the baseline and the worst-case failure columns (Rising time 385.0 sec + 73.5 sec = 458.5 sec, Falling time 88.5 sec + 1.7 sec

90.2 sec).

For Action 3, the NRC staff found that the FMEA documented baseline response time for each module type used in the PINGP PPS sub-system. Degraded components were substituted in each module type to identify impact on response time and confirm ability to identify degradation (or not) during channel calibration. Impact on response times is documented in the FMEA. A bounding response time for each type of module used in the PINGP PPS sub-system based on testing was provided in the FMEA by using degraded components. This bounding response time was combined with other modules/components in the loop to determine the response time for the overall protection channel.

Therefore, the NRC staff found that the proposed methodology in the LAR is like the approved Methodology 2 described in TSTF-569, Revision 2, Attachment 1 and related TR WCAP-14036-P-A, Revision 1 by using the same actions to perform analyses and determine bounding response times for identified modules. In addition, the NRC staff found that adequate information is provided in the LAR with the request for additional information (RAI) response to justify the use of bounding response times in the FMEA for each type of identified modules.

Hence, the NRC staff found that the proposed methodology in the LAR is acceptable for allocating response times for the RTS trip functions in the LAR provided by the PINGP Units 1 and 2 PPS sub-system, instead of performing their periodic RTT.

By using the bounding response times for individual modules assessed and analyzed by the proposed methodology for the PINGP PPS sub-system and components evaluated in the approved TR WCAP-14036-P-A, Revision 1, the summations of response times for individual modules within a loop were made for all RTS reactor trip functions included in the LAR. The loop for each RTS trip function includes its reactor trip breakers. Although the reactor trip breakers are not included in this LAR and will continue to be subject to the required periodic response time testing in the TS, its bounding response time of 100 msec should be added to the total bounding time delay of each loop for all RTS trip functions. The total bounding time delay for each RTS trip function included in this LAR is calculated below by using the proposed methodology and then is compared to the assumed time delay in the Accident Analysis.

TS Table 3.3.1-1 Function 2.a, Power Range High Neutron Flux: The Power Range High Neutron Flux Trip loop includes the following components with their bounding response time: Summing and Level Amplifier NM-310 (1 msec), Bistable Relay Driver NC-306 (65 msec), and Relay Logic (200 msec). The total bounding time delay including the maximum response time (100 msec) of its reactor trip breakers is 366 msec for this RTS trip function provided by the PPS sub-system. The assumed time delay for this PPS trip function in the Accident Analysis is 450 msec. Therefore, the NRC staff found the total bounding response time delay for this PPS trip function is less than the assumed time delay in the Accident Analysis and is deemed acceptable.

TS Table 3.3.1-1 Function 2.b, Power Range Low Neutron Flux: The Power Range Low Neutron Flux Trip loop includes the following components with applicable bounding response times: Summing and Level Amplifier NM-310 (1 msec), Bistable Relay Driver NC-305 (65 msec), and Relay Logic (200 msec). The total bounding time delay including the maximum response time (100 msec) of its reactor trip breakers is 366 msec for this RTS trip function. The assumed time delay for this PPS trip function in the Accident Analysis is 450 msec. Therefore, the NRC staff found the total bounding response time delay for this PPS trip function is less than the assumed time delay in the Accident Analysis and is acceptable.

TS Table 3.3.1-1 Function 3.a, Power Range High Positive Rate: The Power Range High Positive Rate Trip loop includes the following components with applicable bounding response times: Summing and Level Amplifier NM-310 (1 msec), Lag Network NM-311 (135 msec), Bistable Relay Driver NC-303 (65 msec), and Relay Logic (200 msec).

However, the NRC staff did not locate the basis for the time delays for the Lag Network NM-311 (135 msec) and Bistable Relay Driver NC-303 (65 msec) in both the LAR and TR WCAP-14036-P, Revision 1 and therefore issued a Request for Additional Information (RAI) on November 12, 2024 (ML24330A050). In the licensees RAI response dated December 9, 2024 (ML24345A206), the licensee stated that TR WCAP-14036-P, Section 4.6 and Table 8-1 identify the response time for level trips as 65 msec and for flux rate trips as 200 msec. As stated in Attachment 5 to the Methodology in the LAR, the components making up a level trip and a flux rate trip are identical except that the rate trip includes a lag network (NM-311). The nuclear instrumentation system (NIS) channel is comprised of a detector (response time not applicable), a summing and level amplifier (NM-310; response time <1 msec per TR WCAP-14036-P, Section 4.6), a lag network (only applicable for rate trips), and a bistable relay driver NC-303. The response time of the stable relay driver (NC-303) is identical to all bistable relay drivers within the NIS and is identified in the approved TR WCAP-14036-P as a Level Trip Response time of 65 msec. Since the only difference between the level trip and the rate trip is the lag network (NM-311), the response time of the lag network can be derived by subtracting the known response time of the level trip from the total response time of the rate trip (200 msec - 65 msec = 135 msec). Hence, the NRC staff found that the bases provided in this RAI response for the time delays for the Lag Network NM-311 (135 msec) and Bistable Relay Driver NC-303 (65 msec) used in the LAR are adequate.

Therefore, the total bounding time delay including the maximum response time (100 msec) of the PINGP Units 1 and 2 reactor trip breakers is calculated to be 501 msec for this RTS trip function. The assumed time delay for this PPS trip function in the Accident Analysis is 600 msec. Therefore, the NRC staff found the total bounding response time delay for this PPS trip function is less than the assumed time delay in the Accident Analysis and is considered acceptable.

TS Table 3.3.1-1 Function 6, Overtemperature Delta Temperature: The Overtemperature Delta Temperature (OTdT) Trip loop includes the following components with applicable bounding response times: OTdT Trip receives input from Pressurizer Pressure and the only component contributing to Time Response is the Power Supply (PQ-429 NUS Model LPS500) (5 msec). The remainder of the OTdT Loop is shown on Attachment 2.

Thot and Tcold temperature inputs are applied through TT-401A and TT-401B (NUS Model RTL501 mV/I converter) (65 msec) simultaneously, so only a single time delay is assumed. Next the derivation of Delta-T and Tavg occurs simultaneously by TM-405R and TM-401BB (NUS Model TMD500 Time Domain Modules) (50 msec), so only a single time delay is also assumed. The Upper and Lower Neutron Flux Signals are provided from the NIS system through an Isolation Amplifier simultaneously, so again only a single time delay is assumed. The flux signals then are processed by TM-401T (NUS Model SGU501 Function Generator) (141 msec). The Tavg, Pressurizer Pressure, and Delta-Flux Signals are combined by module TM-401B (NUS Model TMD500 Time Domain Module) (50 msec) to develop the OTdT Setpoint. The OTdT Setpoint is compared to the Delta-T Signal by TC-405C/D (NUS Model DAM503 Comparator) (1 msec) to initiate a Channel Trip Signal. The output of the comparator is combined with other channels in a 2 out of 4 coincidence relay logic (200 msec). The total bounding time delay including the maximum response time (100 msec) of its reactor trip breakers and an isolation amplifier (1 msec) is 613 msec. The assumed time delay for this PPS trip function in the Accident Analysis is 6000 msec. Therefore, the NRC staff found the total bounding response time delay for this PPS trip function is less than the assumed time delay in the Accident Analysis and is deemed acceptable.

TS Table 3.3.1-1 Function 7, Overpower Delta Temperature: The Overpower Delta Temperature (OPdT) Trip Loop the following components with applicable bounding response times: Thot and Tcold temperature inputs are applied through TT-401A and TT-401B (NUS Model RTL501 mV/I converter) (65 msec) simultaneously, so only a single time delay is assumed. Next the derivation of Delta-T and Tavg occurs simultaneously by TM-405R and TM-401BB (NUS Model TMD500 Time Domain Modules) (50 msec), so only a single time delay is assumed. The Tavg Signal is provided to TM-401O (NUS Model TMD500) (50 msec) to develop f(Tavg). F(Tavg) is processed through summer TM-401V (NUS Model MTH500 Summator) (12 msec) to develop the OPdT Setpoint. The OPdT Setpoint is compared to the Delta-T Signal by TC-405A/B (NUS Model DAM503 Comparator) (1 msec) to initiate a Channel Trip Signal. The output of the comparator is combined with other channels in a 2 out of 4 coincidence relay logic (200 msec). The total bounding time delay including the maximum response time (100 msec) of its reactor trip breakers is 478 msec for this RTS trip function. The assumed time delay for this PPS trip function in the Accident Analysis is 6000 msec. Therefore, the NRC staff found the total bounding response time delay for this PPS trip function is less than the assumed time delay in the Accident Analysis and is deemed acceptable.

TS Table 3.3.1-1 Function 8.a, Pressurizer Low Pressure: The Low Pressurizer Pressure Reactor Trip loop includes the following components with their bounding response times:

Pressurizer Pressure Signal is reliant on Loop Power Supply PQ-429 (NUS Model LPS500) (5 msec). The Pressurizer Pressure signal is processed through PM-429B (NUS Model TMD500) (50 msec). The signal is compared to Trip Setpoint by module PC-429E (NUS Model SAM503 Comparator) (1 msec). The output of the comparator is combined with other channels in a 2 out of 4 coincidence relay logic (200 msec) in the relay reactor protection system for the total response time. The total bounding time delay including maximum response time of its reactor trip breakers is 356 msec for this RTS trip function. The assumed time delay for this PPS trip function in the Accident Analysis is 1000 msec. Therefore, the NRC staff found the total bounding response time delay for this PPS trip function is less than the assumed time delay in the Accident Analysis and is deemed acceptable.

TS Table 3.3.1-1 Function 8.b, Pressurizer High Pressure: The High Pressurizer Pressure Reactor Trip Loop includes the following components with their bounding response times: Pressurizer Pressure Signal is reliant on Loop Power Supply PQ-429 (NUS Model LPS500) (5 msec). The signal is compared to Trip Setpoint by module PC-429A (NUS Model SAM503 Comparator) (1 msec). The output of the comparator is combined with other channels in a 2 out of 3 coincidence relay logic (200 msec). The total bounding time delay including maximum response time (100 msec) of its reactor trip breakers is 306 msec. The assumed time delay for this PPS trip function in the Accident Analysis is 1000 msec for this RTS trip function. Therefore, the NRC staff found the total bounding response time delay for this PPS trip function is less than the assumed time delay in the Accident Analysis and is considered acceptable.

TS Table 3.3.1-1 Function 10, Reactor Coolant Low Flow: The Low Reactor Coolant Flow Reactor Trip Loop includes the following components with their bounding response times: Reactor Coolant Flow Signal is reliant on Loop Power Supply FQ-411 (Foxboro Model 610AC-O) (5 msec). The signal is compared to Trip Setpoint by module FC-411 (NUS Model SAM503 Comparator) (1 msec). The output of the comparator is combined with other channels in a 2 out of 3 coincidence relay logic (200 msec) in the Relay Reactor Protection System for the total response time. The total bounding time delay including maximum response time (100 msec) of its reactor trip breakers is 306 msec for this RTS trip function. The assumed time delay for this PPS trip function in the Accident Analysis is 1200 msec. Therefore, the NRC staff found the total bounding response time delay for this PPS trip function is less than the assumed time delay in the Accident Analysis and is deemed acceptable.

TS Table 3.3.1-1 Function 13, Steam Generator Low-Low Level: The Low-Low Steam Generator Level Reactor Trip Loop includes the following components with their bounding response times: Steam Generator Level Signal is reliant on Loop Power Supply LQ-461 (NUS Model LPS500) (5 msec). The signal is compared to Trip Setpoint by module LC-461A/B (NUS Model DAM503 Comparator) (1 msec). The output of the comparator is combined with other channels in a 2 out of 3 coincidence relay logic (200 msec) in the Relay Reactor Protection System for the total response time. The total bounding time delay including maximum response time (100 msec) of its reactor trip breakers is 306 msec for this RTS trip function. The assumed time delay for this PPS trip function in the Accident Analysis is 1500 msec. Therefore, the NRC staff found the total bounding response time delay for this PPS trip function is less than the assumed time delay in the Accident Analysis and is deemed acceptable.

3.3 RTS Trip Functions in Current TS Table 3.3.1-1 That Include SR 3.3.1.16 Current TS Table 3.3.1-1 requires the following reactor trip system (RTS) functions include SR 3.3.1.16, which states that to verify RTS RESPONSE TIME is within limits.

1. 2.a. - Power Range Neutron Flux - High

2. 2.b. - Power Range Neutron Flux - Low

3. 3.a. - Power Range Neutron Flux Rate - High Positive Rate

4. 3.b. - Power Range Neutron Flux Rate - High Negative Rate

5. 5. - Source Range Neutron Flux

6. 6. - Overtemperature Delta Temperature (OTT)

7. 7. - Overpower Delta Temperature (OPT)

The licensee proposed to maintain TS Table 3.3.1-1 for the above-mentioned functions 2.a, 2.b, 3.a, 6, and 7 unchanged, therefore, the proposed TS continued to be acceptable.

3.3.1 RTS Trip Functions in TS Table 3.3.1-1 That Were Proposed to Delete SR 3.3.1.16 The licensee proposed to delete SR 3.3.1.16 from the following RTS functions:

1. 3.b. - Power Range Neutron Flux Rate - High Negative Rate The licensee indicated on page 7 of Enclosure 1 to the LAR that function 3.b trip was discussed in the analysis for the dropped rod event. The dropped rod analysis took no credit for a reactor trip and, therefore, there was no trip response time assumed in the accident analysis for function 3.b.

2. 5. - Source Range Neutron Flux The licensee indicated on page 7 of Enclosure 1 to the LAR that function 5 was discussed in the analysis for the rod withdrawal from subcritical (RWFS) event. The analysis of the RWFS event credited function 2.b, Power Range Neutron Flux - Low setpoint, for the reactor trip. The RWFS analysis took no credit of function 5 for the reactor trip, therefore, there was no trip response time assumed in the accident analysis for function 5.

In accordance with its TS Bases, SR 3.3.1.16 requires verifying that the individual channel/train actuation response times are less than or equal to the maximum values assumed in the accident analyses. Based on the TS Bases for SR 3.13.1.16 and the licensees accident analyses that did not credit the trip response times for the above two-mentioned RTS functions, the NRC staff determined that deletion of SR 3.3.1.16 from the above-mentioned RTS functions met the TS Bases for SR 3.3.1.16.

In addition, the NRC staff found that the trip response times referred to in SR 3.3.1.16 for the above-mentioned two RTS functions did not apply to any of the following four criteria in 10 CFR 50.36(c)(2)(ii) required for establishing a TS limiting condition for operation (LCO) of a nuclear reactor:

Criterion 1 - Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2 - A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3 - A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4 - A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

Thus, the NRC staff determined that SR 3.3.1.16 does not need to be included in the TS and the proposed deletions of SR 3.3.1.16 from the above-mentioned two RTS functions meets the criteria in 10 CFR 50.36(c)(2)(ii).

Therefore, the NRC staff concluded that the proposed TS deletions were acceptable.

3.3.3 RTS Trip Functions in TS Table 3.3.1-1 That Were Proposed to Add SR 3.3.1.16 The licensee proposed to add SR 3.3.1.16 to the following RTS functions:

1.

8.a. - Pressurizer Pressure - Low

2.

8.b. - Pressurizer Pressure - High

3.

10. - Reactor Coolant Flow - Low

4.

13. - SG Water Level - Low Low The licensee stated in Section 2.3.2 of Enclosure 1 to the LAR that it confirmed the above-mentioned four RTS functions and the associated RTS trip response times were assumed in the accident analyses. The table in Section 3.1.5 of Enclosure 1 to the LAR showed that the RTS delay times assumed in the accident analyses were 1.0, 1.0, 1.2 and 1.5 seconds for functions 8.a, 8.b, 10, and 13 respectively.

In accordance with its TS Bases, SR 3.3.1.16 requires verifying that the individual channel/train actuation response times are less than or equal to the maximum values assumed in the accident analyses. Based on the TS Bases for SR 3.13.1.16 and the licensees accident analyses assuming the response times for the above-mentioned four RTS functions, the NRC staff finds that the proposed additions of SR 3.3.1.16 regarding RTS trip response times are consistent with the TS Bases. Also, since the trip response times for the above-mentioned RTS functions were included in the accident analysis, the NRC staff has determined that the proposed additions meet 10 CFR 50.36(c)(2)(ii), which requires that a TS LCO of a nuclear reactor must be established for each item meeting one or more of the four criteria. Specifically, Criterion 3 states that a structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

In addition, since the trip response times for the above-mentioned four RTS functions were included in the accident analysis and SR 3.3.1.16 requires verifying that the individual channel/train actuation response times are less than or equal to the maximum values assumed in the accident analyses, the NRC staff has determined that the proposed additions of SR 3.3.1.16 to TS Table 3.3.1-1 meet 10 CFR 50.36(c)(3), which states that surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that the limiting conditions for operation will be met. Therefore, the NRC staff concludes that the proposed TS additions are acceptable.

3.3.4 RTS Response Times Based on Proposed Methodology Bounded by That Assumed in Accident Analyses The licensee summarized in a table in Section 3.1.5 of Enclosure 1 to the LAR the trip response times resulting from the proposed methodologies described in this LAR as compared to the trip response times assumed in the accident analyses for the RTS trip functions that are proposed to be subject to SR 3.3.1.16 for the following RTS functions:

1. 2.a. - Power Range Neutron Flux - High

2. 2.b. - Power Range Neutron Flux - Low

3. 3.a. - Power Range Neutron Flux Rate - High Positive Rate

4. 6. - Overtemperature Delta Temperature (OTT)

5. 7. - Overpower Delta Temperature (OPT)

6. 8.a. - Pressurizer Pressure - Low

7. 8.b. - Pressurizer Pressure - High

8. 10. - Reactor Coolant Flow - Low

9. 13. - SG Water Level - Low Low The licensee indicated that the trip response times from the proposed methodology in all cases were less than the response times assumed in the accident analyses. During the NRC staffs review, the staff found that in the table of Section 3.1.5 the trip response time for the above-mentioned RTS function 3.a, the positive neutron flux rate trip (PFRT), assumed in the accident analysis was 0.60 seconds. This response time of 0.60 seconds was inconsistent with the equivalent response time of 0.50 seconds assumed in the analysis of record (AOR) listed on page 14.3-4 of PINGP updated safety analysis report (USAR) Section 14, and the staff requested additional information on this matter in its RAI of November 12, 2024 (ML24330A050). In its response to the RAI, (ML24345A206), the licensee indicated that it recognized the need to analyze a longer assumed trip delay time for the PINGP PFRT and performed an impact analysis to determine the effect of an increase in the response time from 0.50 seconds to 0.60 seconds for the PFRT function on the AOR. The discussion of the impact analysis was included in the LAR as Attachment 7. The NRC staff reviewed the information in and found that the discussion was for the affected uncontrolled rod cluster control assembly withdrawal at power (RWAP) event. The NRC staff determined that the approach was acceptable, since the AOR for the RWAP event described in Section 14.4.2 of the PINGP USAR credited the PFRT function. The AOR considered multiple cases and verified the applicable safety analysis limits for the minimum departure from nucleate boiling ratio (DNBR), peak RCS pressure, and peak main steam system (MSS) pressure. The PFRT function was credited in the AOR in combination with high neutron flux (HNF) (high setting), overtemperature T (OTT),

and high pressurizer pressure reactor trip functions. The impact analysis was performed for the following three categories of the cases: (1) RWAP cases for minimum DNBR, (2) RWAP cases for peak RCS pressure, and (3) RWAP cases for peak MSS pressure to identify the minimum margin to the safety limits of the DNBR, RCS pressure and MSS pressure, respectively. The NRC staff found that the approach was consistent with that in the AOR, and determined it was acceptable. For all the impact analysis cases the replacement steam generators (RSGs) were modelled, since both PINGP units have RSGs installed.

3.3.4.1 Analysis for DNBR Cases The licensee verified by reviewing the AOR for the RWAP event that a trip response time of 0.50 seconds was assumed for the PFRT. The AOR was performed for the RWAP cases initiated from different power levels (100 percent, 60 percent, and 10 percent) with both maximum and minimum reactivity feedback models and a range of reactivity insertion rates (from 1 to 110 pcm/sec). The PFRT did not occur in any of the cases analyzed with maximum feedback cases. For some cases analyzed with minimum reactivity feedback with relatively high reactivity insertion rates the PFRT would occur. Maximum reactivity feedback and minimum reactivity feedback cases with relatively low reactivity insertion rates would trip on either the OTT or HNF signal. For each of the three initial power levels analyzed, the limiting minimum DNBR for cases with the minimum reactivity feedback would result in a reactor trip on HNF or OTT, but not PFRT. The AOR showed that there was significant DNBR margin between the results for cases that tripped on PFRT and the limiting cases that did not trip on PFRT, such that the PFRT response delay time could be increased by 0.10 seconds to 0.60 seconds without affecting the conclusion that the cases that tripped on PFRT were not limiting. Therefore, the AOR for the RWAP events showed that the impacted cases that tripped on PFRT continued to be bounded by the results of other cases.

The AOR for the RWAP event conservatively modeled a higher setpoint of 9 percent as compared to 6.06 percent for PFRT safety analysis limit setpoint. In all RSG cases with a reactor trip on PFRT, a setpoint of 6.06 percent was reached more than 0.10 seconds before the modeled setpoint of 9 percent was reached. The licensee confirmed this result through a reanalysis of the RSG case that would have the earliest reactor trip on PFRT, which was the minimum reactivity feedback case initiated from 100 percent power with a reactivity insertion rate of 110 pcm/sec. This case would have the fastest nuclear flux rate increase and was therefore, most susceptible to being adversely impacted by the additional time to reach the 9 percent PFRT setpoint. The licensee reanalyzed the case to obtain the results for the rate-lagged indicated nuclear power. The results showed that the setpoint of 9% was exceeded with a value of 9.308 percent at 0.75 seconds transient time, and rod motion would occur 0.50 seconds later at 1.25 seconds transient time. The rate-lagged indicated nuclear power 0.10 second earlier, i.e., at 0.65 seconds transient time was 7.97 percent. This result of the reanalysis demonstrated that the setpoint of 6.06 percent was reached more than 0.10 seconds earlier than the time the PFRT signal generation occurred in the analysis. Based on the above discussion of impact analysis, the NRC staff determined that the setpoint margin inherent in the analyses was sufficient to offset an additional 0.10 seconds, and therefore, the staff concluded that the AOR for the RWAP with the minimum DNBR event, which met the requirements of AEC GDC 14, would not be affected by an increase from 0.50 seconds to 0.60 seconds in the PFRT response time, and remained valid and acceptable.

3.3.4.2 Analysis for RCS Over-pressurization Cases A bounding PFRT response time of 3.0 seconds was assumed in the analysis of the peak RCS pressure cases for the RWAP event in Section 14.4.2 of the PINGP USAR. Therefore, the NRC staff concluded that the AOR for the RCS over-pressurization events, which met the requirements of AEC GDC 14, would not be affected by an increase in the PFRT delay time from 0.50 seconds to 0.60 seconds, and remained valid and acceptable.

3.3.4.3 Analysis for MSS Over-pressurization Cases Sensitivity analyses showed that the acceptance criterion was not challenged for the RWAP event. The AOR also showed that the MSS over-pressurization cases that tripped on PFRT were for higher reactivity insertion rates and would result in an earlier reactor trip generation than the limiting MSS over-pressurization cases, which did not trip on PFRT. Therefore, the NRC staff concluded that the AOR for the peak MSS pressure case, which met the requirements of AEC GDC 14, would not be affected by an increase in the PFRT response time from 0.50 seconds to 0.60 seconds, and remained valid and acceptable.

Based on the above discussion of the cases for DNBR, RCS and MSS over-pressurization, the NRC staff found that when the PFRT delay time increased from 0.50 seconds to 0.60 seconds, the results of the analysis for the affected RWAP event met the AEC GDC 14 requirements, and theredore, would remain valid and acceptable.

3.4 Technical Conclusion The NRC staff has reviewed the information provided in the licensees LAR submittal dated June 3, 2024, as supplemented by letter dated December 9, 2024. Based on the NRC staffs evaluation of the licensees proposed deletions and additions of SR 3.3.1.16 from and to the applicable RTS functions in TS Table 3.3.1-1 as discussed in Sections 3.2 and 3.3 of this SE, the NRC staff has concluded that the proposed TS changes are consistent with the TS Bases for SR 3.3.1.16, and meet the requirements of 10 CFR 50.36(c)(2)(ii) and 10 CFR 50.36(c)(3)..

Therefore, the NRC staff has determined that the proposed TS changes are acceptable for PINGP.

The NRC staff also has reviewed the accident analyses supporting the proposed RTS response times from the methodologies described in TSTF-569, Revision 2 and WCAP 14036-P-A, Revision 1. Based on the NRC staffs evaluation in Section 3.2 of this SE, the NRC staff has concluded that: (1) the accident analyses were acceptable, since the analyses continued to meet AEC GDC 14 that requires core protection systems be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits, and (2) the accident analyses showed for all applicable cases that the values of the RTS response times assumed in the analyses bounded the proposed values derived from the methodologies described in this LAR. Therefore, the NRC staff has determined that the proposed RTS response times derived from the methodologies described in this LAR were adequately supported by the accident analyses and were acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Minnesota State official was notified of the proposed issuance of the amendments on January 24, 2025. 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 (89 FR 63992). 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 NRC staff has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: SSun, NRR TSweat, NRR JZhao, NRR Date of Issuance: June 9, 2025

ML25111A239 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DEX/EICB/BC NRR/DSS/SNSB/BC NAME BWetzel SRohrer FSacko DMurdock DATE 4/21/2025 4/23/2025 05/07/2025 5/05/2025 OFFICE NRR/DSS/STSB/BC OGC (NLO)

NRR/DORL/LPL3/BC (A) NRR/DORL/LPL3/PM NAME SMehta STurk IBerrios BWetzel DATE 5/05/2025 5/23/2025 6/05/2025 6/06/2025