ML23096A184

Issuance of Amendment No. 328 Revise and Relocate Reactor Coolant System Pressure and Temperature Limits and Pressurizer Overpressure Protection System Limits to Pressure and Temperature Limits Report

ML23096A184
Person / Time
Site:	Salem
Issue date:	05/09/2023
From:	James Kim Plant Licensing Branch 1
To:	Carr E Public Service Enterprise Group
Kim J
References
EPID L 2022 LLA-0111
Download: ML23096A184 (1)

Text

May 9, 2023 Mr. Eric Carr President and Chief Nuclear Officer PSEG Nuclear LLC - N09 P.O. Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 - ISSUANCE OF AMENDMENT NO. 328 RE: REVISE AND RELOCATE REACTOR COOLANT SYSTEM PRESSURE AND TEMPERATURE LIMITS AND PRESSURIZER OVERPRESSURE PROTECTION SYSTEM LIMITS TO PRESSURE AND TEMPERATURE LIMITS REPORT (EPID L-2022-LLA-0111)

Dear Mr. Carr:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 328 to Renewed Facility Operating License No. DPR-75 for the Salem Nuclear Generating Station, Unit No. 2, in response to your application dated August 7, 2022.

The amendment revised the technical specifications by relocating the pressure-temperature (P-T) limits for the reactor pressure vessel to a licensee-controlled pressure and temperature limits report (PTLR) and replacing the existing reactor vessel heatup and cooldown rate limits and the P-T limit curves with references to the PTLR. The amendment also updated the existing P-T limits to extend their applicability through the period of extended operation to 50 effective full-power years. A copy of the related safety evaluation is also enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

James S. Kim, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-311

Enclosures:

1. Amendment No. 328 to DPR-75

2. Safety Evaluation cc: Listserv

PSEG NUCLEAR LLC CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-311 SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 328 Renewed License No. DPR-75

1.

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

A.

The application for amendment filed by PSEG Nuclear LLC dated August 7, 2022, 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-75 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 328, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

The license amendment is effective as of its date of issuance and shall be implemented within 60 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Hipólito J. González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License and Technical Specifications Date of Issuance: May 9, 2023 Hipolito J.

Gonzalez Digitally signed by Hipolito J. Gonzalez Date: 2023.05.09 11:23:00

-04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 328 SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 RENEWED FACILITY OPERATING LICENSE NO. DPR-75 DOCKET NO. 50-311 Replace the following page of Renewed Facility Operating License No. DPR-75 with the attached revised page as indicated. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert 3

3 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert I

I XVIII XVIII 1-5 1-5 3/4 1-9 3/4 1-9 3/4 4-3 3/4 4-3 3/4 4-4a 3/4 4-4a 3/4 4-27 3/4 4-27 3/4 4-28 3/4 4-28 3/4 4-29 3/4 4-29 3/4 4-31 3/4 4-31 3/4 5-7 3/4 5-7 3/4 5-8 3/4 5-8 6-24b 6-24b 6-24c Renewed License No. DPR-75 Amendment No. 328 (3)

PSEG Nuclear LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4)

PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source or 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; (5)

PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, 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 calibration or associated with radioactive apparatus or components; and (6)

PSEG Nuclear LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level PSEG Nuclear LLC is authorized to operate the facility at steady state reactor core power levels not in excess of 3459 megawatts (thermal).

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 328, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. PSEG Nuclear LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS DEFINED TERMS............................................................................................... 1-1 ACTION............................................................................................................. 1-1 AXIAL FLUX DIFFERENCE................................................................................... 1-1 CHANNEL CALIBRATION.................................................................................... 1-1 CHANNEL CHECK............................................................................................. 1-1 CHANNEL FUNCTIONAL TEST............................................................................. 1-1 CONTAINMENT INTEGRITY................................................................................. 1-2 CORE ALTERATION.......................................................................................... 1-2 CORE OPERATING LIMITS REPORT................................................................... 1-2 DOSE EQUIVALENT I-131................................................................................... 1-2 DOSE EQUIVALENT XE-133................................................................................. 1-3 ENGINEERED SAFETY FEATURE RESPONSE TIME......................................... 1-3 FREQUENCY NOTATION.................................................................................... 1-3 FULLY WITHDRAWN......................................................................................... 1-3 GASEOUS RADWASTE TREATMENT SYSTEM.................................................. 1-3 IDENTIFIED LEAKAGE....................................................................................... 1-3 INSERVICE TESTING PROGRAM......................................................................... 1-4 OFFSITE DOSE CALCULATION MANUAL (ODCM)............................................. 1-4 OPERABLE - OPERABILITY................................................................................. 1-4 OPERATIONAL MODE - MODE............................................................................. 1-4 PHYSICS TESTS................................................................................................ 1-5 PRESSURE BOUNDARY LEAKAGE..................................................................... 1-5 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)............................... 1-5 PROCESS CONTROL PROGRAM (PCP).............................................................. 1-5 PURGE-PURGING............................................................................................. 1-5 QUADRANT POWER TILT RATIO......................................................................... 1-5 RATED THERMAL POWER.................................................................................. 1-5 REACTOR TRIP SYSTEM RESPONSE TIME....................................................... 1-6 REPORTABLE EVENT....................................................................................... 1-6 SHUTDOWN MARGIN........................................................................................ 1-6 SOLIDIFICATION................................................................................................ 1-6 SOURCE CHECK............................................................................................... 1-6 STAGGERED TEST BASIS................................................................................... 1-6 THERMAL POWER............................................................................................ 1-7 UNIDENTIFIED LEAKAGE................................................................................... 1-7 VENTILATION EXHAUST TREATMENT SYSTEM................................................ 1-7 VENTING.......................................................................................................... 1-7 SALEM - UNIT 2 I

Amendment No. 328

INDEX ADMINISTRATIVE CONTROLS

==================================================================

SECTION PAGE 6.1 RESPONSIBILITY..6-1 6.2 ORGANIZATION Onsite and Office Organizations 6-1 Facility Staff...6-1 Shift Technical Advisor 6-7 6.3 FACILITY STAFF QUALIFICATIONS..6-7 6.4 DELETED 6-7 6.5 REVIEW AND AUDIT (THIS SECTION DELETED)..6-8 6.6 REPORTABLE EVENT ACTION.6-16 6.7 SAFETY LIMIT VIOLATION.6-16 6.8 PROCEDURES AND PROGRAMS 6-17 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS 6-20 6.9.2 SPECIAL REPORTS..6-24C 6.10 RECORD RETENTION.6-25 6.11 RADIATION PROTECTION PROGRAM 6-26 6.12 HIGH RADIATION AREA..6-27 6.13 PROCESS CONTROL PROGRAM (PCP).6-28 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM)...6-29 6.15 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATMENT SYSTEMS.6-29 6.16 TECHNICAL SPECIFICATION (TS) BASES CONTROL PROGRAM...6-30 6.17 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM..6-31 SALEM - UNIT 2 XVIII Amendment No. 328

DEFINITIONS PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 of the Updated FSAR, 2) authorized under the provisions of 10CFR50.59, or 3) otherwise by the Commission.

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 1.20a The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, and the Overpressure Protection System setpoint and enable temperature, for the current reactor vessel fluence period. The pressure and temperature limits shall be determined for each fluence period in accordance with Technical Specification Section 6.9.1.11.

PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary-to-secondary leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

PROCESS CONTROL PROGRAM (PCP) 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of radioactive waste.

PURGE - PURGING 1.23 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector 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. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3459 MWt.

SALEM - UNIT 2 1-5 Amendment No. 328

REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 At least one charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE.#

APPLICABILITY: MODES 4, 5 and 6.

ACTION:

With no charging pump OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until one charging pump is restored to OPERABLE status.

SURVEILLANCE REQUIREMENTS 4.1.2.3 No additional Surveillance Requirements other than those required by the INSERVICE TESTING PROGRAM.

A maximum of one centrifugal charging pump shall be OPERABLE while in MODE 4 when the temperature of one or more of the RCS cold legs is less than or equal to, the POPS enable temperature specified in the PTLR, MODE 5, or MODE 6 when the head is on the reactor vessel.

SALEM - UNIT 2 3/4 1-9 Amendment No. 328

REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3

a.

At least two of the coolant loops listed below shall be OPERABLE:

1.

Reactor Coolant Loop 21 and its associated steam generator and reactor coolant pump,*

2.

Reactor Coolant Loop 22 and its associated steam generator and reactor coolant pump,*

3.

Reactor Coolant Loop 23 and its associated steam generator and reactor coolant pump,*

4.

Reactor Coolant Loop 24 and its associated steam generator and reactor coolant pump,*

5.

Residual Heat Removal Loop 21,

6.

Residual Heat Removal Loop 22.

b.

At least one of the above coolant loops shall be in operation.**

APPLICABILITY: MODE 4 ACTION:

a.

With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

b.

With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to the POPS enable temperature specified in the PTLR unless 1) the pressurizer water volume is less than 1650 cubic feet (equivalent to approximately 92% of level) or 2) the secondary water temperature of each steam generator is less than 50°F above each of the RCS cold leg temperatures.

All reactor coolant pumps and residual heat removal pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 2) core outlet temperature is maintained at least 10°F below saturation temperature.

SALEM - UNIT 2 3/4 4-3 Amendment No. 328

REACTOR COOLANT SYSTEM COLD SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.4 Two# residual heat removal loops shall be OPERABLE* and at least one RHR loop shall be in operation.**

APPLICABILITY: MODE 5.##

ACTION:

a.

With less than the above required loops operable, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.

b.

With no RHR loop in operation; suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4 At least one residual heat removal loop shall be verified to be in operation and circulating reactor coolant in accordance with the Surveillance Frequency Control Program.

One RHR loop may be inoperable for up to two hours for surveillance testing, provided the other RHR loop is OPERABLE and in operation. Additionally, four filled reactor coolant loops, with at least two steam generators with their secondary side water levels greater than or equal to 5% (narrow range), may be substituted for one residual heat removal loop.

A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to the POPS enable temperature specified in the PTLR unless 1) the pressurizer water volume is less than 1650 cubic feet (equivalent to approximately 92% of level), or 2) the secondary water temperature of each steam generator is less than 50°F above each of the RCS cold leg temperatures.

Systems supporting RHR loop operability may be excepted as follows:

a.

The normal or emergency power source may be inoperable.

The residual heat removal pumps may be de-energized for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 2) core outlet temperature is maintained at least 10°F below saturation temperature.

SALEM - UNIT 2 3/4 4-4a Amendment No. 328

REACTOR COOLANT SYSTEM 3/4.4.10 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.10.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limits specified in the PTLR with:

a.

A maximum heatup rate within the limits specified in the PTLR,

b.

A maximum cooldown rate within the limits specified in the PTLR, and

c.

A maximum temperature change within the limits specified in the PTLR, during hydrostatic testing operations above system design pressure.

APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS Tavg and pressure to less than 200°F and 500 psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.10.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits specified in the PTLR in accordance with the Surveillance Frequency Control Program during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.10.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50, Appendix H. The results of these examinations shall be used to update the P-T Limit Curves specified in the PTLR.

SALEM - UNIT 2 3/4 4-27 Amendment No. 328

THIS PAGE LEFT INTENTIONALLY BLANK SALEM - UNIT 2 3/4 4-28 Amendment No. 328

THIS PAGE LEFT INTENTIONALLY BLANK SALEM - UNIT 2 3/4 4-29 Amendment No. 328

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.10.3 At least one of the following overpressure protection systems shall be OPERABLE:

a.

Two Pressurizer Overpressure Protection System relief valves (POPS) with a lift setting of less than or equal to the value specified in the PTLR, or

b.

The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 3.14 square inches.

APPLICABILITY: When the temperature of one or more of the RCS cold legs is less than or equal to the POPS enable temperature specified in the PTLR, except when the reactor vessel head is removed.

ACTION:

a.

With one POPS inoperable in MODE 4 and the temperature of one or more of the RCS cold legs is less than or equal to the POPS enable temperature specified in the PTLR, restore the inoperable POPS to OPERABLE status within 7 days or depressurize and vent the RCS through a 3.14 square inch vent(s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a vented condition until both POPSs have been restored to OPERABLE status.

b.

With one POPS inoperable in MODES 5 or 6 with the Reactor Vessel Head installed, restore the inoperable POPS to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or complete depressurization and venting of the RCS through at least a 3.14 square inch vent(s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a vented condition until both POPSs have been restored to OPERABLE status.

c.

With both POPSs inoperable, depressurize and vent the RCS through a 3.14 square inch vent(s) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a vented condition until both POPSs have been restored to OPERABLE status.

d.

In the event either the POPS or the RCS vent(s) are used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the POPS or vent(s) on the transient and any corrective action necessary to prevent recurrence.

e.

LCO 3.0.4.b is not applicable when entering MODE 4.

SURVEILLANCE REQUIREMENTS 4.4.10.3.1 Each POPS shall be demonstrated OPERABLE by:

SALEM - UNIT 2 3/4 4-31 Amendment No. 328

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - Tavg < 350°F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a.

One OPERABLE centrifugal charging pump# and associated flow path capable of taking suction from the refueling water storage tank and transferring suction to the residual heat removal pump discharge piping and;

1.

Discharging into each Reactor Coolant System (RCS) cold leg.

b.

One OPERABLE residual heat removal pump and associated residual heat removal heat exchanger and flow path capable of taking suction from the refueling water storage tank on a safety injection signal and transferring suction to the containment sump during the recirculation phase of operation and;

1.

Discharging into each RCS cold leg, and; upon manual initiation,

2.

Discharging into two RCS hot legs.

APPLICABILITY: MODE 4.

ACTION:

a.

With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

b.

With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or residual heat removal pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System Tavg less than 350°F by use of alternate heat removal methods.

c.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

d.

LCO 3.0.4.b is not applicable to ECCS high head subsystem A maximum of one safety injection pump or one centrifugal charging pump shall be OPERABLE in MODE 4 when the temperature of one or more of the RCS cold legs is less than or equal to the POPS enable temperature specified in the PTLR, Mode 5, or Mode 6 when the head is on the reactor vessel.

SALEM - UNIT 2 3/4 5-7 Amendment No. 328

EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - Tavg < 350°F SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per applicable Surveillance Requirements of 4.5.2.

4.5.3.2 All safety injection pumps and centrifugal charging pumps, except the above required OPERABLE pump, shall be demonstrated to be inoperable in accordance with the Surveillance Frequency Control Program while in MODE 4 and the temperature of one or more of the RCS cold legs is less than or equal to the POPS enable temperature specified in the PTLR, MODE 5, or MODE 6 when the head is on the reactor vessel by either of the following methods:

a.

By verifying that the motor circuit breakers have been removed from their electrical power supply circuits or,

b.

By verifying that the pump is in a recirculation flow path and that two independent means of preventing RCS injection are utilized.

SALEM - UNIT 2 3/4 5-8 Amendment No. 328

ADMINISTRATIVE CONTROLS

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;

4. The number of tubes plugged during the inspection outage; and

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;

f.

The results of any SG secondary side inspections.

6.9.1.11 REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, POPS enable temperature, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

1. Specification 3.1.2.3, "Charging Pump-Shutdown"

2. Specification 3.4.1.3, Reactor Coolant System Hot Shutdown

3. Specification 3.4.1.4, Reactor Coolant System Cold Shutdown"

4. Specification 3/4.4.10.1, "Reactor Coolant System Pressure/Temperature Limits"

5. Specification 3.4.10.3, "Reactor Coolant System Overpressure Protection Systems"

6. Specification 3/4.5.3, Emergency Core Cooling Systems ECCS Subsystems -

Tavg <350°F

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC.

1. WCAP-14040-A, Rev. 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May, 2004.

2. WCAP-18124-NP-A, Rev 0, Fluence Determination with RAPTOR-M3G and FERRET, July 2018, may be used as an alternative to Section 2.2 of WCAP-14040-A Rev. 4.

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

SALEM - UNIT 2 6-24b Amendment No. 328

ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, with a copy to the Administrator, USNRC Region I within the time period specified for each report.

6.9.3 DELETED 6.9.4 When a report is required by ACTION 1, 4, 8 OR 9 of Table 3.3-11 "Accident Monitoring Instrumentation", a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring for inadequate core cooling, the cause of the inoperability, and the plans and schedule for restoring the instrument channels to OPERABLE status.

SALEM - UNIT 2 6-24c Amendment No. 328

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 328 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-75 PSEG NUCLEAR LLC CONSTELLATION ENERGY GENERATION, LLC SALEM NUCLEAR GENERATING STATION, UNIT NO. 2 DOCKET NO. 50-311

1.0 INTRODUCTION

By letter dated August 7, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22220A248), PSEG Nuclear LLC (the licensee) submitted a license amendment request (LAR) to modify technical specification (TS) Section 1.0, Definitions, Section 3/4, Limiting Conditions for Operation and Surveillance Requirements, and Section 6.0, Administrative Controls, by relocating the pressure-temperature (P-T) limits for the reactor pressure vessel (RPV) to a licensee-controlled pressure and temperature limits report (PTLR), and replacing the existing reactor vessel heatup and cooldown rate limits and the pressure and temperature (P-T) limit curves with references to the PTLR at Salem Nuclear Generating Station (Salem), Unit No. 2. The LAR also proposed to update the existing P-T limits to extend their applicability through the period of extended operation to 50 effective full-power years (EFPY).

Pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 50.90, the licensee proposed, specifically, to (1) modify TS Section 1.0, Definitions, (2) delete TS Figure 3.4-3, Salem Unit 2 Reactor Coolant System Heatup Limitations. and TS Figure 3.4-3, Salem Unit 2 Reactor Coolant System Cooldown Limitations, (3) update references in TS Section 3/4 to the pressurizer overpressure protection system (POPS) setpoint to refer to the PTLR, and (4) modify TS Section 6.0, Administrative Controls, to add 6.9.1.11 describing the approved analytical methods applied to evaluate P-T limit curves. The licensee provided the proposed PTLR in Enclosure 2 to the LAR, including the updated P-T limit curves for 50 EFPY.

2.0 REGULATORY EVALUATION

The regulations in 10 CFR 50.36, Technical specifications, paragraph (a), require 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) of 10 CFR 50.36 requires, in part, that TSs include the following categories related to facility operation: (1) safety limits, limiting safety systems settings, and control settings; (2) limiting conditions for operation, (3) surveillance requirements, (4) design features, and (5) administrative controls.

The regulations in 10 CFR 50.60, Acceptance criteria for fracture prevention measures for light-water nuclear power reactors for normal operation, impose fracture toughness and material embrittlement surveillance program requirements, as set forth in Appendices G and H to 10 CFR Part 50.

The regulations in 10 CFR 50.61, Fracture toughness requirements for protection against pressurized thermal shock events, require, in part, that pressurized-water reactors (PWRs) to project reference temperature for RPV beltline materials at the end of license (RTPTS) to satisfy pressurized thermal shock (PTS) screening criteria, thereby assuring that the fracture toughness of the reactor vessel is maintained below the threshold throughout the operating life of the reactor vessel.

The regulations in 10 CFR Part 50, Appendix A, General Design Criteria [GDC] for Nuclear Power Plants, include the following GDCs applicable to fracture prevention of the reactor coolant pressure boundary:

GDC 14, Reactor coolant pressure boundary, which requires that the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, or rapidly propagating failure, and of gross rupture.

GDC 15, Reactor coolant system design, which requires that reactor coolant system and associated auxiliary, control, and protection systems be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during normal operation and anticipated operational occurrences (AOOs).

GDC 30, Quality of reactor coolant pressure boundary, which requires, in part, that components comprising the reactor coolant pressure boundary be designed, fabricated, erected, and tested to the highest quality standards practical.

GDC 31, Fracture prevention of reactor coolant pressure boundary, which requires, in part, that the reactor coolant pressure boundary be designed with sufficient margin to assure that the boundary behaves in a nonbrittle manner, and the probability of rapidly propagating fracture is minimized during normal operation, maintenance, testing and postulated accident conditions. The GDC requires that the design reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining material properties, the effects on irradiation on material properties, residual stresses, and size of the flaws is specifically applicable to this license amendment request.

The regulations in 10 CFR Part 50, Appendix G, Fracture Toughness Requirements, require, in part, that facility P-T limits for the RPV be at least as conservative as those obtained by applying the linear elastic fracture mechanics methodology of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Appendix G, Fracture Toughness Criteria for Protection Against Failure. Also, 10 CFR Part 50, Appendix G requires that reactor vessel beltline materials have Charpy upper-shelf energy (USE) to maintain a specified value in the beginning of commercial operation and throughout the life of the reactor vessel, unless it is demonstrated that lower values of Charpy USE will provide margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASME Code.

The regulations in 10 CFR Part 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, establish requirements for a facilitys surveillance program for monitoring changes in fracture toughness due to neutron irradiation and the thermal environment.

Generic Letter (GL) 96-03, Relocation of Pressure and Temperature Limit Curves and Low Temperature Overpressure Protection System Limits, permits relocation of the P-T limits from the TS to a PTLR (ML031110004). GL 96-03 requires 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 Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials,"

describes procedures for calculating the adjusted nil-ductility transition reference temperature RTNDT (ART) due to neutron irradiation on RPVs (ML003740284).

Regulatory Issue Summary (RIS) 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components,"

provides evaluation guidance for P-T limit curves and PTLRs, including the consideration of neutron fluence and structural discontinuities in the development of P-T limit curves (ML14149A165).

NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" (SRP), Section 5.3.2, "Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock", provides an acceptable method for determining the P-T limits based on the methodology of the ASME Code,Section XI, Appendix G.

RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," provides guidance regarding neutron fluence calculations (ML010890301).

Technical Specifications Task Force (TSTF) Traveler No. TSTF-419-A, Revision 0, "Revise PTLR Definition and References in ISTS 5.6.6, RCS [Reactor Coolant System] PTLR, "

(ML012690234) was approved for use by licensees, by the U.S. Nuclear Regulatory Commission (NRC or Commission) by letter dated March 21, 2002 (ML020800488).

TSTF-419-A allows licensees to use current Topical Reports to support limits in the PTLR without having to submit an amendment to the facility operating license every time the Topical Report is revised, and it assures that approved versions of the referenced Topical Reports or plant-specific methodologies will be used in determining the P-T limits or LTOP System limits.

By letter dated August 4, 2011 (ML110660285), the NRC staff clarified the use of TSTF-419-A.

3.0 TECHNICAL EVALUATION

3.1 Background

In Salem, Unit No. 2, Amendment 224 (ML011350051), the P-T limit curves were calculated using the Westinghouse vessel fluence methodology. Allowable pressure-temperature relationships for various heatup and cooldown rates were calculated using methods derived from Appendix G in Section XI of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50 and these methods are discussed in detail in WCAP-14040-NP-A, Revision 2, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January 1996, and ASME Pressure Vessel Code Case N-640, Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1," approved March 1999. Adjusted reference temperatures at the nil ductility transition values were developed for the RPV materials in accordance with RG 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials. The NRC staff approved revising the RPV P-T limits and extending their validity to 32 effective full-power years. These are the current P-T limits in the Salem, Unit No. 2, TS.

The current Salem, Unit No. 2, P-T limit curves expire at 32 EFPY, which Salem, Unit No. 2, is expected to reach by approximately July 15, 2024. The licensee proposed a set of revised P-T limit curves that were generated through the 60 years end of license extension (EOLE) (i.e.,

50 EFPY). WCAP-18124-NP-A (ML18204A010) methodology provides justification for the Salem, Unit No. 2, fluence determination. WCAP-18124-NP-A methodology and the NRC approved WCAP-18124-NP-A, Rev. 0, Supplement 1-NP-A Rev. 0 (ML22153A139) are used for Salem, Unit No. 2, Heatup and Cooldown Limit Curves for Normal Operation.

The neutron fluence calculations were performed using the three-dimensional discrete ordinates code, RAPTOR-M3G, the BUGLE-96 cross-section library, and the least-squares evaluation FERRET Code for the surveillance capsule dosimetry that were approved and described in WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET (Attachment 3 to Enclosure 1 of the LAR).

3.2 Licensees Neutron Fluence Calculations The methods used to develop the calculated pressure vessel fluence utilize the NRC-approved methodology described in WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET." The neutron transport evaluation methodology described in the WCAP report is based on the guidance of RG 1.190. The NRC staff concluded that the calculational methodology described in WCAP-18124-NP-A, Revision 0 is acceptable for use in calculating RPV neutron fluence provided the following limitations and conditions are met:

1. Applicability of WCAP-18124-NP, Revision 0, is limited to the reactor pressure vessel region near the active height of the core based on the uncertainty analysis performed and the measurement data provided.

Additional justification should be provided via additional benchmarking, fluence sensitivity analysis to response parameters of interest (for example, pressure-temperature limits, material stress and strain), margin assessment, or a combination thereof, for applications of the method to components including, but not limited to, the reactor pressure vessel upper circumferential weld, and reactor coolant system inlet and outlet nozzles and reactor vessel internal components.

2. Least squares adjustment is acceptable if the adjustments to the measured or calculated 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.

A validation of the fluence model is provided in Appendix C of Enclosure 4 of the LAR. The fluence is based upon operation for 50 EFPY. Cycle-specific calculations were performed for Cycles 1 through 24.

In the LAR, the calculated fast neutron fluences at the end of plant life (50 EFPY) are provided below:

Parameter Fluence (n/cm2)

Peak Surface (Intermediate Shell Plates) 2.05x1019 Peak 1/4 T (Intermediate Shell Plates) 1.22x11019 Limiting Beltline Material Peak Surface (Intermediate Shell Longitudinal Weld Seams 2-442 B & C) 1.47x1019 Limiting Beltline Material Peak 1/4 T (Intermediate Shell Longitudinal Seams 2-442 B & C) 0.876x1019 3.3 Licensee Proposed P-T Limits The proposed Salem, Unit No. 2, PTLR is presented in Enclosure 2 of the LAR. The licensee stated that it prepared the P-T limit curves and PTLR in accordance with GL 96-03; ASME Code,Section XI, Appendix G; and WCAP-14040-A. Specifically, the licensee calculated:

(1) the P-T limits in accordance with WCAP-14040-A, Revision 4, and ASME Code,Section XI, Appendix G and (2) the adjusted reference temperature (ART) values for the limiting beltline materials in accordance with RG 1.99, Revision 2 (as discussed in GL 96-03).

The development of the 60-year Salem, Unit No. 2, P-T limit curves is documented in detail in of the LAR, WCAP-18502-NP. The licensee stated that all vessel materials subject to a fluence greater than 1x1017 n/cm2 were considered in developing the P-T curves, consistent with RIS 2014-11. The licensee stated that the inlet and outlet nozzles were considered as part of the P-T limit curve evaluation due to the stress concentration at the nozzle corner. Table 3-1 of WCAP-18502-NP shows copper and nickel content of the vessel materials, as well as unirradiated reference nil-ductility temperature (RTNDT.)

Section 4 of WCAP-18502-NP describes the available data for calculation of a chemistry factor (CF) according to RG 1.99, Revision 2, Regulatory Position 2.1. The licensee presented relevant sister plant surveillance data. Section 5 of WCAP-18502-NP describes the calculation of the CFs.

3.4.

NRC Staff Evaluation

The NRC staff evaluated the licensees neutron fluence calculations, the proposed PTLR implementation, the proposed P-T limit curves, the ART calculation, the PTS calculation, the USE calculation, and the proposed changes to the plant technical specifications.

3.4.1 Neutron Fluence Evaluation The NRC staff reviewed the LAR and associated documentation referred to in the LAR for neutron fluence projections. Based on the regulatory evaluation in section 2.0 of this safety evaluation (SE), the staff reviewed the neutron fluence projected at 60 years (50 EFPY) for the Salem, Unit No. 2, RPV materials that should be considered in the development of P-T limit curves due to a predicted neutron fluence exposure greater than 1.0 x 1017 n/cm2 (E > 1.0 MeV). This review included the additional materials that are outside the traditional beltline region but that have the potential to exceed this fluence threshold, referred to as the extended beltline materials.

To apply the computer codes and cross-section library as described in WCAP-18124-NP-A to the calculation of neutron fluence, the licensee performed an applicability evaluation of WCAP-18124-NP-A and presented the results in LAR Section 3.2 for neutron fluence calculations.

The licensee concluded that WCAP-18124-NP-A is applicable to Salem, Unit No. 2.

Specifically, the licensee applied the methodology described in WCAP-18124-NP-A to Salem, Unit No. 2, traditional and extended beltline (i.e., non-beltline) materials including reactor coolant system inlet and outlet nozzles. In performing the fast neutron exposure evaluations for Salem, Unit No. 2, RPV, the licensee conducted a series of fuel-cycle-specific forward transport calculations by using the three-dimensional discrete ordinates computer code RAPTOR-M3G, the BUGLE-96 cross-section library, and the least-squares evaluation FERRET code. The NRC staff reviewed the relevant information provided by the licensee and agrees that WCAP-18124-NP-A is applicable and can be used for Salem, Unit No. 2.

For Salem, Unit No. 2, transport calculations, the licensees reactor model was constructed to include the necessary RPV details encompassing the traditional beltline region as well as the inclusion of the surveillance capsules and associated support structures (as described in Section 2.2 of WCAP-18502-NP). The NRC staff determined that the spatial mesh and angular quadrature and the pointwise inner iteration flux convergence criterion as utilized with this reactor model for WCAP-18502-NP are in conformance with RG 1.190 and are therefore acceptable.

Regarding the development of source distribution used in the transport calculations, the relevant information was outlined in Section 2.2 of WCAP-18502-NP. The NRC staff determined that the preparation of core neutron source for the transport calculations is in conformance with RG 1.190 and is, therefore, acceptable. The NRC staff finds that the results from the licensees neutron transport calculations provided acceptable data in terms of fuel cycle-averaged neutron flux, which when multiplied by the appropriate fuel cycle length, would generate reliable incremental fast neutron exposure for each fuel cycle until the 60 years EOLE.

Based on its review of WCAP-18502-NP, WCAP-14040-A, and associated references, the NRC staff determined that an evaluation of the dosimetry sensor sets from the surveillance capsules withdrawn from Salem, Unit No. 2, was provided. The documented dosimetry analyses showed that the +/-20 percent (1) acceptance criterion specified in RG 1.190 is met.

In addition to the traditional beltline materials, the licensee also identified the RPV non-beltline materials and their locations in Tables 1 and 2 of the LAR, Enclosure 1, Attachment 3, for the development of P-T limits. Among them, the licensee explicitly considered the RPV materials with structural discontinuities, such as nozzles, as mentioned in RIS 2014-11.

The NRC staff determined from WCAP-18502-NP that the model used for the transport calculations as mentioned above had been expanded to axially and azimuthally encompass the non-beltline materials to calculate the neutron fluence to 50 EFPY. The licensee then performed an uncertainty analysis to demonstrate whether the +/-20 percent (1) acceptance criterion specified in RG 1.190 is met. The methodology used to generate WCAP-18502-NP (i.e., WCAP-18124-NP-A) is based on RG 1.190. Since RG 1.190 is guidance to project neutron fluence for the beltline region and is not directly applicable to the extended beltline (i.e., non-beltline) region, the staff concluded the calculational methodology described in WCAP-18124-NP-A, is acceptable for use in calculating RPV neutron fluence.

The methodology described in WCAP-18124-NP-A was applied to Salem, Unit No. 2, extended beltline materials and RCS inlet and outlet nozzles. Additional justification for the Salem, Unit No. 2, application is provided in Attachment 3 to Enclosure 1. Based on the fluence uncertainty analysis, benchmarking, and margin assessment described in Attachment 3 of Enclosure 1 to the LAR, the applicability of the RAPTOR-M3G fluence determination methodology is justified for the Salem, Unit No. 2, RPV extended beltline region and RCS inlet and outlet nozzle fluence determination. The NRC staff has determined that Limitation and Condition 1 described in WCAP-18124-NP-A is met for Salem, Unit No. 2.

Limitation and Condition 2 in WCAP-18124-NP-A only applies in situations where the least squares analysis is used to adjust the calculated values of neutron exposure. For Salem, Unit No. 2, Limitation and Condition 2 does not apply as the least-squares procedures were not used to adjust the calculated fast neutron (E > 1.0 MeV) fluence values for RPV materials evaluated in the updated reactor vessel integrity analysis (Enclosure 4). It is noted that for Salem, Unit No. 2, the least-squares results were only used to compare the calculations and measurements from the evaluated dosimetry and validate the neutron transport models, and those comparisons showed satisfactory results.

Based on the radiation transport calculation results and the traditional beltline and extended beltline (i.e., non-beltline) materials information (see section 3.2 of this safety evaluation), the NRC staff confirmed that the licensee had tabulated and provided the fast neutron (E > 1.0 MeV) fluence projections to 50 EFPY for both beltline and extended beltline materials from WCAP-18502-NP to the LAR for Salem, Unit No. 2. The staff concludes that the projection meets the acceptance criterion specified in 10 CFR Part 50, Appendix G, and conforms with the guidance in RIS 2014-11 because the licensee applied the approved methodology of WCAP-18124-NP-A to Salem, Unit No. 2, RPV materials that have the potential to experience a neutron fluence exposure greater than 1.0 x 1017 n/cm2 (E > 1.0 MeV) at EOLE for the development of the P-T limit curves.

3.4.2 PTLR Implementation The NRC staff evaluated the proposed Salem, Unit No. 2, PTLR in accordance with the seven criteria in Attachment 1 to GL 96-03 as discussed below.

Criterion 1 requires that the PTLR methodology describes the transport calculation methods including computer codes and formula used to calculate neutron fluence values.

The proposed Salem, Unit No. 2, PTLR (Enclosure 2 to the LAR) describes the neutron fluence calculation methodology in Section 4.3. Therefore, the NRC staff finds that the licensee has satisfied Criterion 1 of GL 96-03.

Criterion 2 requires that (1) the RPV material surveillance program comply with 10 CFR 50, Appendix H, (2) surveillance capsule removal schedule be provided, and (3) test results of surveillance specimens be used to update the P-T limit curves in the PTLR.

The Salem, Unit No. 2, surveillance program is described in Appendix A of the proposed PTLR and in updated final safety analysis report (UFSAR), section 5.2.4.4. In accordance with 10 CFR Part 50, Appendix H, the licensee has removed the following four surveillance capsules from the Salem, Unit No. 2, RPV -- Capsule T in 1983, Capsule U in 1986, Capsule X in 1991, and Capsule Y in 2000. The test results of the specimens from the removed capsules are presented in various reports as shown in Appendix A of the proposed PTLR. Appendix C of the proposed PTLR presents how the test results of surveillance data were analyzed to determine the credibility of the surveillance data which were used to update the P-T curves in the proposed PTLR. The licensee stated that Capsule S is scheduled to be removed at 32 EFPY, tentatively in May 2023. The licensee further stated that Capsules V, W and Z are on a standby status. The NRC staff determined that (1) the Salem, Unit No. 2, surveillance program is consistent with 10 CFR Part 50, Appendix H, (2) the licensee has used the surveillance data to develop the proposed P-T limit curves as further discussed in this safety evaluation, and (3) the Salem, Unit No. 2, surveillance program includes the capsule removal schedules and the adequate number of capsules. Therefore, the NRC staff finds that the proposed PTLR has satisfied Criterion 2.

Criterion 3 requires that the PTLR methodology describes how the low temperature overpressure protection system limits are calculated applying system/thermal hydraulics and fracture mechanics.

The licensee described the POPS (i.e., low temperature overpressure protection system) in Section 4.14 of the proposed Salem, Unit No. 2, PTLR (Enclosure 2 to the LAR). The licensee referenced the NRC-approved POPs methodology and described the existing POPs setpoint of 375 psig. The licensee stated that the new setpoint as analyzed by the approved methodology is 434 psig. Given the use of the approved methodology, the staff finds that the licensee may update the PTLR to the new setpoint without submission to the NRC for approval, provided that the NRC approves the use of the proposed PTLR in this SE. The NRC staff finds that the licensee has satisfied Criterion 3 of GL 96-03 because of the discussion included in Section 4.14 of the proposed PTLR.

Criterion 4 requires that the PTLR methodology describes the method for calculating the ART values using RG 1.99, Revision 2, and that the licensee identify the RTPTS value in accordance with 10 CFR 50.61.

The licensee described the calculation of the ART values in Section 4.10 of the proposed Salem, Unit No. 2, PTLR (Enclosure 2 to the LAR). The licensee referenced WCAP-18502-NP (Enclosure 3 to the LAR) as describing the methodology to calculate ART. The NRC staff verified that WCAP-18502-NP implements the RG 1.99, Revision 2, methodology for calculating ART. Further, the staff reviewed the ART values given in Table 7-2 of WCAP-18502-NP. The staffs compared these values with values available for similar plants, specifically Salem, Unit No. 1, and confirmed that the licensee correctly implemented the guidance of RG 1.99, Revision 2 and that the licensee has provided the limiting ART values in Table 4-4 of the proposed PTLR.

The licensee described the calculation of RTPTS in Section 4.11 of the proposed Salem, Unit No. 2, PTLR (Enclosure 2 to the LAR). The licensee referenced WCAP-18502-NP, Tables G-1 and G-2. The licensee stated that Lower Shell Longitudinal Weld Seam 3-442 A&C exceeded the 10 CFR 50.61 pressurized thermal shock screening criteria of 270 °F at 50 EFPY. The NRC staff finds that the licensee has satisfied Criterion 4 of GL 96-03 because of the discussion included in Sections 4.10 and 4.11 of the proposed PTLR.

Criterion 5 requires that the PTLR methodology describes the application of fracture mechanics in the construction of P-T limits based on the ASME Code,Section XI, Appendix G, and SRP Section 5.3.2.

The licensee described the proposed operating limits, including the P-T limits, in Section 3 of the proposed Salem, Unit No. 2, PTLR (Enclosure 2 to the LAR). The license provided the underlying technical background of the operating limits in Section 4 of the proposed PTLR. The licensee stated that the NRC-approved methodology is WCAP-14040-A, Revision 4 (ML050120209). The NRC staff reviewed the proposed EFPY 50 P-T limits and found that the proposed curves were reasonable. The NRC staff finds that the licensee has satisfied Criterion 5 of GL 96-03 because of the discussion included in Sections 3 and 4 of the proposed PTLR.

Criterion 6 requires that the PTLR methodology describes how the minimum temperature requirements such as minimum bolt-up temperature and hydrotest temperature in Appendix G to 10 CFR Part 50 are applied to P-T curves.

The licensee stated that minimum bolt-up temperature requirements are established in Appendix G to 10 CFR Part 50. WCAP-14040, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, (ML050120209) specifies that the minimum bolt-up temperature should be 60 °F or the limiting unirradiated RTNDT of the RPV closure flange region, whichever is higher. Section 4.9 of the proposed PTLR states that the minimum bolt-up temperature for the Salem, Unit No. 2, RPV is 60 °F without margins for instrument uncertainties and 62 °F with margins for instrument uncertainties. The NRC staff determined that the minimum temperatures of the proposed P-T limit curves satisfy the required minimum temperatures of table 1 of 10 CFR Part 50, Appendix G, as discussed further in the safety evaluation. The NRC staff finds that the proposed PTLR has satisfied Criterion 6.

Criterion 7 requires that the PTLR describes (1) how increase in RTNDT is calculated using RPV surveillance data based on RG 1.99, Revision 2, and (2) if surveillance data from other nuclear plants are used, provide supplemental data to demonstrate credibility of the data.

Appendix C of the proposed PTLR discusses credibility of surveillance data based on the five credibility criteria of RG 1.99, Revision 2. The five credibility criteria are: (1) use of controlling surveillance materials with regard to radiation embrittlement, (2) scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small to determine the 30 ft-lb temperature and upper-shelf energy, (3) the scatter of the RTNDT value should be in consistent with the guidance of RG 1.99, Revision 2, (4) the irradiation temperature of the Charpy specimens in the surveillance capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25 °F, and (5) the surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material.

The licensee used the five criteria to demonstrate the credibility of the surveillance data from Diablo Canyon Power Plant (Diablo Canyon), Unit No. 2, weld Heat #21935/12008 and from Salem, Unit No. 2, RPV shell plate B-4712-2. The licensee stated that the Diablo Canyon, Unit No. 2, surveillance weld data is applicable to Salem, Unit No. 2, lower shell longitudinal weld seams 3-442 A, B, and C because the Diablo Canyon, Unit 2, and Salem, Unit No. 2, RPV weld 3-442A, B, and C were fabricated with the same wire Heat #21935/12008 using Linde 1092 weld flux. The Diablo Canyon, Unit 2, surveillance data for Heat #21935/12008 is reported in WCAP-17315-NP, Diablo Canyon Units 1 and 2 Pressurized Thermal Shock and Upper-Shelf Energy Evaluations. The Diablo Canyon, Unit No. 2, surveillance weld data will be discussed further in this safety evaluation.

The licensee further stated that this approach is similar to that taken for the McGuire Nuclear Station (McGuire), Units 1 and 2, Measurement Uncertainty Recapture (MUR) Power Uprate submittal, evaluated in WCAP-17455-NP, McGuire Units 1 and 2 Measurement Uncertainty Recapture (MUR) Power Uprate: Reactor Vessel Integrity and Neutron Fluence Evaluations.

The NRC staff approved the McGuires power uprate submittal in its SE (ML13073A041).

McGuire also used Diablo Canyon, Unit No. 2, surveillance data to evaluate the McGuire, Unit 1, RPV weld 3-442 A, B, and C with wire Heat #21935/12008.

Appendix C of the proposed PTLR also presents the credibility analysis of Salem, Unit No. 2, RPV shell plate B-4712-2 based on specimen data from four surveillance capsules in the Salem, Unit No. 2, RPV. The licensee showed that the scatter of the RTNDT values for plate B-4712-2 specimens is within the range of 17 °F as specified in RG 1.99, Revision 2 and therefore, surveillance data from the Salem, Unit No. 2, RPV for plate B-4712-2 are credible.

The NRC staff noted that the ART of plate B-4712-2 is lower than the ART of weld 3-442A & C; therefore, plate B-4712-2 is not considered in the development of the P-T limit curves.

The NRC staff determined that the licensees proposed PTLR has demonstrated the credibility of Diablo Canyon, Unit No. 2, surveillance data and Salem, Unit No. 2, plate B-4712-2 based on criteria of RG 1.99, Revision 2 in calculating the ART. Therefore, the proposed PTLR has satisfied Criterion 7.

In conclusion, the NRC staff finds that the licensee has satisfied all seven criteria of to GL 96-03.

3.4.3 Evaluation of Proposed P-T Limit Curves The licensee developed the proposed P-T limit curves for the EOLE, (i.e., 60 years of operation (equivalent to 50 EFPY)) based on the KIc methodology detailed in the 1998 through the 2000 Addenda Edition of the ASME Code,Section XI, Appendix G, as shown in Figures 6-1 and 6-2 of WCAP-16982. The formulas and methodology for calculating the EOLE ART values, which are based on fluence calculated at 50 EFPY, are provided in WCAP-18571. The licensee stated that Salem, Unit No. 2, had generated P-T limit curves as shown in WCAP-16982 as part of the Salem license renewal application process but the P-T limit curves developed for the 50 EFPY operation were not submitted for approval at the time. To implement the proposed P-T limit curves in Figures 3-1 and 3-2 of the proposed PTLR, the licensee evaluated the applicability of the proposed P-T limit curves by comparing the 50 EFPY limiting 1/4T and 3/4T ART values contained in WCAP-16982 with those in Tables 6-2 and 6-3 of WCAP-18571. The NRC staff verified that the 1/4T and 3/4T ART values of the limiting material, weld 3-442A and C, were used in the proposed P-T limit curves in the proposed PTLR. After verifying that the limiting ART values were used in the proposed P-T limit curves, the NRC staff evaluated the development of the P-T limit curves.

The NRC staff noted that proposed leak test limit curve in figure 3-1 of the proposed PTLR for inservice leak and hydrostatic testing was developed based on pressure only because the temperature gradient was not significant during leak testing. As such, the NRC staff determined that the Klt can be set to zero in the fracture mechanics calculation because temperature gradient in the RPV shell wall is too small to generate any significant thermal stresses during the leak tests. The ASME Code,Section XI, Appendix G, requires that the safety factor for the KIm is set to 1.5 for the leak test curve. The NRC staff performed an independent calculation and verified that the P-T limit curve for leak test is consistent with the methodology of the ASME Code,Section XI, Appendix G, and, therefore, is acceptable.

Figure 3-1 of the proposed PTLR contains two heatup curves for operation when the core is not critical. One heatup curve is limited for a heatup rate of less than or equal to () 60 °F/hour and the other heatup curve is for a heatup rate of less than or equal to () 100 °F/hour. The ASME Code,Section XI, Appendix G, requires that the safety factor of 2.0 and 1.0 for the KIm and Klt, respectively, be used to develop the heatup curve. The NRC staff performed an independent calculation and verified that both heatup curves when the core is not critical are consistent with the methodology of ASME Code,Section XI, Appendix G, and, therefore, are acceptable.

Figure 3-1 of the proposed PTLR contains two critical P-T limit curves for normal operation when the core is critical. One critical P-T limit curve is for a heatup rate of less than or equal to

() 60 °F/hour and the other critical P-T limit curve is for a heatup rate of less than or equal to

() 100 °F/hour. The regulations in 10 CFR Part 50, Appendix G, require that when the core is critical, the critical P-T limit curve shall be, at a minimum, 40 °F higher than the heatup curve when the core is not critical under all pressure regimes. The NRC staff verified that the temperature data points of the two critical limit curves (when the core is critical) are at least 40 °F higher than the temperatures of the two heatup curves (when the core is not critical). The NRC staff determined that both critical limit curves are consistent with the requirements of 10 CFR Part 50, Appendix G, and the ASME Code,Section XI, Appendix G, and, therefore, are acceptable.

Figure 3-2 of the proposed PTLR contains five cooldown curves. Each of the cooldown curves was developed based on the corresponding cooldown rates of less than or equal to () 0, 20, 40, 60, and 100 °F/hour. The ASME Code,Section XI, Appendix G, specifies that the safety factors of 2.0 and 1.0 for the Kim and Klt, respectively, be used to develop the cooldown curves.

The NRC staff performed independent calculations and verified that all five cooldown curves are consistent with the methodology of ASME Code,Section XI, Appendix G, and, therefore, are acceptable.

Inlet and Outlet Nozzles P-T Limit Curves NRC RIS 2014-11 requires that the P-T limit curves account for the high stresses in the nozzle corner region due to the potential for more restrictive P-T limits, even if the RTNDT for these components are not as high as those of the RPV beltline shell materials that have simpler geometries. The licensee stated that Pressurized Water Reactor Owners Group (PWROG)

Report, PWROG-15109-NP-A, PWR Pressure Vessel Nozzle Appendix G Evaluation, January 2020, (ML20024E573), addresses the RIS 2014-11 concern generically for the U.S.

PWR operating fleet. PWROG-15109-NP-A addresses the potential for P-T limit curves for inlet or outlet nozzle comers of U.S. PWRs that could be more limiting than the current NRC-approved P-T limits of the shell (and associated welds) in the beltline region and closure flange regions, as applicable of the RPV below a fluence threshold of 4.28 x 1017 n/cm2 (E > 1.0 MeV). PWROG-15109-NP-A proposes that if the neutron fluence at the RPV nozzle regions is below 4.28 x 1017 n/cm2 (E > 1.0 MeV), the P-T limit curves need not to be developed for the RPV nozzle regions.

By letter dated October 31, 2019 (ML19301D063 and ML19301D160), the NRC staff approved PWROG-15109-NP-A as an acceptable means to address the concerns of RIS 2014-11 for the nozzle regions experiencing neutron fluence below a threshold of 4.28 x 1017 n/cm2 (E > 1.0 MeV).

The licensee stated that the results and conclusions of PWROG-15109-NP-A are applicable to the RPV nozzle regions at Salem, Unit No. 2, which experience a neutron fluence less than the fluence threshold of 4.28 x 1017 n/cm2 (E > 1.0 MeV). The NRC staff verified that the neutron fluence values for the inlet and outlet nozzles attached to the Salem, Unit No. 2, RPV as shown in Table 2-2 of WCAP-18571 are lower than the fluence threshold of 4.28 x 1017 n/cm2. As such, the NRC staff determined that the Salem, Unit No. 2, RPV nozzles are not the limiting material in the development of P-T limit curves. The NRC staff finds that the licensee has satisfactorily addressed the information in RIS 2014-11 based on PWROG-15109-NP-A because the licensee has sufficiently addressed all ferritic materials of the reactor vessel, including the impact of structural discontinuities from components such as nozzles, and addressed the impact of neutron fluence accumulation in accordance with the requirements of 10 CFR Part 50, Appendix G.

10 CFR Part 50, Appendix G Minimum Temperature Requirements Appendix G to 10 CFR Part 50, table 1, requires that specific minimum RCS temperatures be achieved when RCS pressures reach a certain level as part of heatup, cooldown and hydrotest operations. The minimum temperatures and pressures provide assurance that the RPV shell plates and welds are not overstressed.

Appendix G to 10 CFR Part 50, Table 1, with associated footnote 1, specifies a limit of 20 percent of the preservice system hydrostatic test pressure. The licensee stated that the preservice hydrostatic test pressure is 3106 psig as shown in Section 6.1 of WCAP-16982. The calculated 20 percent of 3106 psig is 621 psig. However, the licensee included a measurement uncertainty of 61 psig as shown in Section 3.0 of the proposed PTLR. Therefore, the 20 percent pressure limit of 621 psig is reduced to 560 psig (621 psig - 61 psig). The NRC staff finds that the application of an uncertainty of 61 psig is acceptable to lower the operating pressure prior to heatup or cooldown. The NRC staff finds that a lower pressure protects the RPV shell materials from overstress and therefore, a pressure of 560 psig is acceptable when compared to a pressure of 621 psig.

Appendix G to 10 CFR Part 50, table 1, items 1.a and 2.a, requires that when the operating pressure is less than or equal to () 20 percent of the preservice system hydrostatic test pressure, the minimum temperature on the P-T curve must be the highest reference temperature of the material in the closure flange region that is highly stressed by the bolt preload. The licensee stated that the closure flange reference temperature is 60 °F as discussed in Section 6-2 of WCAP-16982. The licensee added 2 °F to the closure flange reference temperature to account for the margin. As such, the minimum temperature on the P-T curve is 62 °F. The NRC staff verified that the P-T curves in Figures 3-1 and 3-2 of the proposed PTLR satisfy the requirements in 10 CFR Part 50, Appendix G, Table 1, items 1.a and 2.a.

Appendix G to 10 CFR Part 50, Table 1, item 1.b, requires that when the pressure is greater than (>) 20 percent of the preservice system hydrostatic test pressure, the minimum temperature on the P-T curve must be the highest reference temperature of the material in the closure flange region that is highly stressed by the bolt preload plus 90 °F. However, footnote 6 to table 1 states that Lower temperatures are permissible if they can be justified by showing that the margins of safety of the controlling region are equivalent to those required for the beltline when it is controlling. This requirement creates a bend (a knee) in the non-beltline P-T curve at 560 psig because the closure flange is part of the non-beltline region. The NRC staff verified that the P-T curves in Figures 3-1 and 3-2 of the proposed PTLR satisfy the requirements in 10 CFR Part 50, Appendix G, table 1, item 1.b.

Appendix G to 10 CFR Part 50, Table 1, item 2.b, requires that when the pressure is greater than (>) 20 percent of the preservice system hydrostatic test pressure, the minimum temperature on the P-T curve must be the highest reference temperature of the material in the closure flange region that is highly stressed by the bolt preload plus 120 °F. Section 6.1 of WCAP-16982 states that the limiting unirradiated RTNDT of 12 °F (rounded up to 15 °F) is associated with the flange of the Salem, Unit No. 2, RPV. The licensee used an uncertainty of 18 °F for temperature and 61 psi for pressure. The licensee stated that the minimum allowable temperature of this region is 153 °F (15 °F +18 °F +120 °F) for normal operation at pressures greater than 560 psig with margins for instrument uncertainties as shown in Tables 6-2 and 6-3 and Figures 6-1 and 6-2 of WCAP-16982. The NRC staff verified that the minimum allowable temperature of 153 °F for a pressure greater than 560 psi is consistent with Table 1, Item 2.b of Appendix G to 10 CFR, Part 50, and therefore is acceptable.

Appendix G to 10 CFR Part 50, Table 1, item 2.c requires that when the core is critical and the operating pressure is less than or equal to () 20 percent of the preservice system hydrostatic test pressure, the minimum temperature must be the larger of the minimum permissible temperature for the inservice system hydrostatic pressure test or the closure flange RTNDT plus 40 °F. The licensee calculated a minimum temperature of 289 °F as shown in Table 6-2 of WCAP-16982. The NRC staff noted that 289 °F is higher than the closure flange RTNDT plus 40 °F; therefore, the minimum temperature of 289 °F is acceptable. This is because a higher temperature in the subject temperature regime minimizes the potential of pressurized thermal shock on the RPV shell metal and is preferrable than a lower temperature. The NRC staff verified that the P-T curves in Figures 3-1 and 3-2 of the proposed PTLR satisfy the requirements in 10 CFR Part 50, Appendix G, table 1, item 2.c.

Appendix G to 10 CFR Part 50, table 1, item 2.d requires that when the core is critical and the reactor internal pressure is greater than (>) 20 percent of the preservice system hydrostatic test pressure, the minimum temperature must be the larger of the minimum permissible temperature for the inservice system hydrostatic pressure test or closure flange RTNDT plus 160 °F. The licensee calculated a minimum temperature of 289 °F as shown in Table 6-2 of WCAP-16982.

The NRC staff noted that 289 °F is higher than the closure flange RTNDT plus 160 °F and therefore, is acceptable. The NRC staff verified that the P-T curves in Figures 3-1 and 3-2 of the proposed PTLR satisfy the requirements in 10 CFR Part 50, Appendix G, table 1, item 2.d.

The NRC staff finds that the proposed P-T curves in Figures 3-1 and 3-2 of the proposed PTLR satisfy the minimum temperature requirements of 10 CFR Part 50, Appendix G, Table 1 and, therefore, are acceptable. The NRC staff further finds that the P-T curves in Figures 3-1 and 3-2 of the proposed PTLR satisfy the requirements of GDC 14, 15, 30, and 31; 10 CFR Part 50, Appendix G; and the ASME Code,Section XI, Appendix G.

3.4.4 Evaluation of Adjusted Reference Temperature RG 1.99, Revision. 2, defines the ART 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 CF and a fluence factor. The CF is calculated based on the amount of copper (Cu) and nickel (Ni) in the RPV shell material. The CF may be calculated from the tables of RG 1.99, Revision. 2, or from surveillance data. The fluence factor is calculated based on the neutron fluence at the postulated flaw depths of 1/4T and 3/4T locations. 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 copper and nickel contents, the neutron fluence and the calculational procedures.

The licensee calculated the limiting ARTs of 209 °F and 150 °F at the 1/4T and 3/4T locations, respectively, for 50 EFPY at lower shell longitudinal weld seams 3-442A & C as shown in table 6-1 of WCAP-16982. The NRC staff notes that the licensee did not use surveillance data from other nuclear plants (e.g., Diablo Canyon) to calculate the ARTs as shown in WCAP-16982.

For the current submittal, the licensee used Diablo Canyon surveillance data to calculate the ART for weld 3-442A & C at Salem, Unit No. 2, which is allowed per RG 1.99, Revision 2. As such, the licensee calculated the ARTs of 180.5 °F and 123.8 °F at 1/4T and 3/4T locations, respectively, for 50 EFPY as shown in Table 6-2 of WCAP-18571.

However, because the previous calculated ARTs for weld 3-442A and C (209 °F and 150 °F) in WCAP-16982-NP are the highest of all ARTs of reactor vessel plates and welds, the licensee has decided to use the ART values of weld 3-442A & C (209 °F and 150 °F) that were used to develop the P-T limit curves for 50 EFPY as shown in Figures 6-1 and 6-2 of WCAP-16982-NP as the limiting ART to establish the proposed P-T limit curves.

Initial Reference Temperature of Nil-Ductility Transition The initial RTNDT values for the Salem, Unit No. 2, RPV plate and weld materials are presented in Table 3-1 of WCAP-18571. The licensee stated that the initial RTNDT values for the RPV plate material are based on the actual measurement. As discussed in Table 3-1 of WCAP-18571, the licensee assigned a generic initial RTNDT value of -56 °F for Linde 0091 and 1092 weld flux that were used in the fabrication of the Salem, Unit No. 2, RPV shell weld in accordance with 10 CFR 50.61. The licensee identified the limiting material weld 3-442A & C as a Linde 1092 weld material as shown in table 3-1 of WCAP-18571. The NRC staff verified that to calculate the ART, the licensee used the appropriate generic initial RTNDT for weld 3-442A & C as specified in 10 CFR 50.61 for Linde 0091 and 1092 weld flux.

Chemistry Factor Table 5-3 of WCAP-18571 presents the chemistry factors for all RPV materials using Positions 1.1 and 2.1 of RG 1.99, Revision 2 to develop the proposed P-T limit curves.

Table 4-3 of the proposed PTLR (and table 5-1 of WCAP-18571) presents the calculated chemistry factors based on surveillance data for the RPV materials that were included in the Salem, Unit No. 2, surveillance capsule program.

The licensee stated that Salem, Unit No. 2, welds 3-442A, B & C have the same Heat #21935/12008 as that of Diablo Canyon, Unit 2, welds as discussed in section 4.6 of the proposed PTLR. The licensee evaluated the credibility of Diablo Canyons weld data with respect to the Salem, Unit No. 2, welds 3-442A, B and C because the RPV welds of these two plants have the same heat number. As shown in Appendix A of WCAP-18571, based on five screening criteria, the licensee stated that the surveillance data from Diablo Canyon, Unit 2, welds with Heat #21935/12008 are credible and, therefore, are applicable to the Salem, Unit No. 2, welds 3-442A, B and C. However, the NRC staff noted that the licensee did not use the calculated ARTs based on the Diablo Canyon surveillance data to develop the proposed P-T limits. Instead, the licensee used the ARTs in the P-T limits for the 50 EFPY in WCAP-16982-NP to develop the proposed P-T limits. The NRC staff finds that the ARTs in the P-T limits for the 50 EFPY in WCAP-16982-NP are higher (thus conservative) than the ART values calculated using the Diablo Canyon surveillance data; therefore, the NRC staff finds that the ARTs for the proposed P-T limit curves are acceptable.

Margin Tables 6-2 and 6-3 of WCAP-18571 present the margin term of the 1/4T and 3/4T locations for each of the RPV shell material, respectively. The NRC staff noted that Equation 4 of Position C.1.1 of RG 1.99, Revision 2, specifies the margin calculation. The margin consists of Sigma I and Sigma delta. Sigma I is applicable to the standard deviation of the initial RTNDT based on either a generic value or measured value. For welds, 10 CFR 50.61 permits the use of a generic value of 17 °F for Sigma I. The Sigma delta is applicable to the standard deviation of shift in RTNDT ( RTNDT). For welds, Sigma delta is 28 °F as specified in RG 1.99, Revision 2.

The NRC staff verified that the licensee used the appropriate margin to calculate the ART of 209 °F and 150 °F for Salem, Unit No. 2, weld 3-442A & C in accordance with RG 1.99, Revision 2, for the proposed P-T limit curves.

3.4.5 Pressurized Thermal Shock The regulations in 10 CFR 50.61 establish screening criteria for PWR vessel embrittlement, which is measured by the maximum RTNDT in the limiting beltline material at the end of the plant license, to prevent a PTS event in PWRs. The screening criteria are represented as RTPTS values. The screening criteria are based on the RTPTS values for (1) beltline axial welds, forgings or plates, and (2) beltline circumferential weld seams. The regulations in 10 CFR 50.61 set the RTPTS screening criterion for plates, forgings, and axial welds to be less than 270 °F, and the screening criterion for circumferentially oriented welds to be less than 300 °F. The NRC revised 10 CFR 50.61 in 1991 and 1995 to change the procedure for calculating radiation embrittlement such that the procedure for calculating the RTPTS values is to be consistent with the methods given in RG 1.99, Revision 2.

The licensee used the method in 10 CFR 50.61 to calculate the RTPTS values for the Salem, Unit No. 2, RPV materials as presented in table B-1 of WCAP-18571. The NRC staff independently calculated the RTPTS for the Sale, Unit No. 2, RPV materials and verified that the RTPTS values in Table B-1 of WCAP-18571-NP satisfy the requirements of 10 CFR 50.61 because the RTPTS values of Salem, Unit No. 2,, RPV shell materials are below the RTPTS screening criteria values of 270 °F for plates, forgings, and axial weld materials, and below 300 °F for circumferential weld materials at 50 EFPY.

3.4.6 Upper Shelf Energy The upper shelf Charpy energy (USE) that represents fracture toughness of RPV shell materials decreases because of the neutron fluence irradiation, as the RPV operates for an extended period.Section IV.A.1 of 10 CFR Part 50, Appendix G, requires the USE of the RPV shell material to be above 50 ft-lb value and requires licensees to submit an analysis at least 3 years prior to the time that the Charpy USE of the RPV shell materials is predicted to drop below 50 ft-lb.

The NRC staff notes that two methods can be used to estimate the change in USE, depending on the availability of credible surveillance capsule data as defined in RG 1.99, Revision 2. For RPV beltline materials that are not in the surveillance program or are not credible, the Charpy USE is assumed to decrease as a function of fluence and copper and nickel content, as indicated in Position C.1.1 of RG 1.99, Revision 2. When two or more credible surveillance data become available, the credible surveillance data may be used to determine the Charpy USE of the RPV beltline material based on Position C.2.1 of RG 1.99, Revision 2.

The licensee used the 1/4T fluence values to project USE values to determine if the Salem, Unit No. 2, beltline and extended beltline materials remain above the 50 ft-lb limit at 50 EFPY as shown in tables 5-1 and 5-2 of WCAP-16982-NP. Table 5-1 of WCAP-16982-NP presents the predicted end of license renewal (i.e., at 50 EFPY) USE value for the RPV shell materials based on Position C.1.1 of RG 1.99, Revision 2. Table 5-2 of WCAP-16982-NP presents the predicted end of license renewal USE value for the surveillance material (i.e., Intermediate Shell Plate B4712-2) based on Position C.2.2 of RG 1.99, Revision 2. The licensee also plotted the percentage of drop in USE vs. neutron fluence for plate B4712-2 as shown in figure 5-1 of WCAP-16982-NP. The licensee stated that all of the beltline and extended beltline materials in the Salem, Unit No. 2, RPV are projected to remain above the USE screening criterion value of 50 ft-lb at 50 EFPY.

The NRC staff verified that plate B4712-2, which is plotted in figure 5-1 is consistent with figure 2 of RG 1.99, Revision 2. The NRC staff determined that the USE for plate B4712-2 at the end of license renewal will be above 50 ft-lb after the drop of the USE value from the initial USE.

In addition, the NRC staff performed an independent calculation and verified that the USE values for the beltline and extended beltline materials in tables 5-1 and 5-2 of WCAP-16982-NP remain above the 50 ft-lb limit at 50 EFPY. Therefore, the NRC staff finds that Salem, Unit No. 2, satisfies the USE limit of 10 CFR Part 50, Appendix G.

3.4.7 Evaluation of TS Changes - Consistency with GL 96-03 and TSTF-419-A The NRC staff reviewed the licensee's proposed TS revisions for implementation of the proposed PTLR to determine whether the revisions address the criteria of GL 96-03 and TSTF-419-A.

1. The NRC staff verified that the licensee's revision to TS Section 1.1, "Definitions," to include the new definition of the PTLR, "Pressure and Temperature Limits Report (PTLR)," is consistent with TSTF-419-A. Therefore, it is acceptable.

2. In TS Section 3/4.4.6, "Pressure/Temperature Limits," the LCO, action statements, and SRs are revised to replace all specified P-T limit figures, heatup and cooldown rates, and minimum temperature criteria with a reference to the applicable limits in the PTLR.

TS Section 3/4.4.6 specifically requires RCS operation within the limits specified in the PTLR. The NRC staff verified that this is consistent with GL 96-03 and TSTF-419-A.

Therefore, it is acceptable.

3. A new TS requirement has been added to the Salem, Unit No. 2, TS administrative controls. Specifically, Section 6.9.1.11, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," has been added. This TS section contains the TS administrative controls governing the content, methodology, and NRC reporting requirements for updates to the PTLR. The NRC staff identified that the new TS Section 6.9.1.10 does the following:

(a) It identifies that the P-T limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing, including the heatup and cooldown rates, shall be established in the PTLR.

(b) It identifies the TS sections (TS Section 3/4.4.9, LCO, action statements, and SRs) that require operation in accordance with the limits in the PTLR, per GL 96-03 and TSTF-419-A.

(c) It specifies that the analytical methods used to determine the P-T limits shall be those previously reviewed and approved by the NRC.

(d) It specifies that the PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

The NRC staff verified that all the above PTLR administrative controls to be included in the new TS Section 6.9.1.11 are consistent with GL 96-03 and TSTF-419-A. Therefore, the NRC staff determined that the addition of TS Section 6.9.1.11 is acceptable.

3.5 Technical Conclusion Based on information submitted by the licensee, the NRC staff determined that: (1) the proposed P-T curves have used the limiting ART values which are derived in accordance with the methodology of RG 1.99, Revision 2; (2) the proposed P-T curves were determined based on the NRC-approved methodology in WCAP-14040-A; SRP 5.3.2; and the ASME Code,Section XI, Nonmandatory Appendix G; (3) the proposed P-T limit curves are valid for 50 EFPY; and (4) the proposed Salem, Unit No. 2, revised TS satisfies 10 CFR Part 50 Appendices G and H, 10 CFR 50.60, 10 CFR 50.61, GL 96-03, and TSTF-419-A.

Therefore, the NRC Staff concludes that the Salem Unit No. 2 TS, as amended by the proposed changes will continue to meet the requirements of 10 CFR 50.36.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the New Jersey State official was notified of the proposed issuance of the amendment on April 20, 2023. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves 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 amendment involves no significant hazards consideration, as published in the Federal Register (87 FR 67508; November 8, 2022), and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

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

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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: J. Tsao F. Forsaty M. Hamm Date of Issuance: May 9, 2023

ML23096A184 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DNRL/NVIB/BC NAME JKim KEntz DWidrevitz (A)

DATE 4/11/2023 4/10/2023 2/24/2023 OFFICE NRR/DSS/SNSB/BC (A)

NRR/DSS/STSB/BC OGC - NLO NAME DWoodyatt VCusumano STurk DATE 3/31/2023 4/13/2023 5/4/2023 OFFICE NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME HGonzález JKim DATE 05/09/2023 05/09/2023