ML23198A359

Issuance of Amendment Nos. 322 and 212 Analytical Methodology to the Core Operating Limits Report for a Full Spectrum Loss of Coolant Accident (EPID L-2022-LLA-0129) - Nonproprietary

ML23198A359
Person / Time
Site:	Beaver Valley
Issue date:	10/02/2023
From:	Sujata Goetz NRC/NRR/DORL/LPL1
To:	Blair B Energy Harbor Nuclear Corp
References
EPID L-2022-LLA-0129
Download: ML23198A359 (1)

Text

OFFICIAL USE ONLY PROPRIETARY INFORMATION

October 2, 2023

Barry N. Blair Site Vice President Energy Harbor Nuclear Corp.

Beaver Valley Power Station Mail Stop P-BV-SSB P.O. Box 4, Route 168 Shippingport, PA 15077-0004

SUBJECT:

BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 322 AND 212 RE: ANALYTICAL METHODOLOGY TO THE CORE OPERATING LIMITS REPORT FOR A FULL SPECTRUM LOSS OF COOLANT ACCIDENT (EPID L-2022-LLA-0129)

Dear Mr. Blair:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment Nos. 322 and 212 to Renewed Facility Operating License Nos. DPR-66 and NPF-73 for the Beaver Valley Power Station, Unit Nos. 1 and 2, respectively. This amendment is in response to your application dated August 31, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22249A257).

This amendment revises the Beaver Valley technical specifications (TS) 5.6.3, Core Operating Limits Report (COLR) by adding the Westinghouse Electric Company LLC (Westinghouse)

Topical Report WCAP-16996-P-A, Rev.1, Realistic LOCA [loss-of-coolant accident] Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA (FSLOCA)

Methodology), to the list of approved analytical methods used to determine the core operating limits and adds a note to the loss-of-coolant accident methods listed in TS 5.6.3.b to restrict their future use. The proposed amendment also removes Zircalloy from the list of fuel rod cladding in TS 4.2.1, Fuel Assemblies.

Enclosure 3 to this letter contains sensitive unclassified non-safeguards information. When separated from Enclosure 3, this document is DECONTROLLED.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION

B. Blair - 2 -

The NRC staff has determined that the related safety evaluation contains proprietary information pursuant to Title 10 of the Code of Federal Regulations Section 2.390, Public inspections, exemptions, requests for withholding, The proprietary information is indicated by text enclosed within double brackets. The proprietary version of the safety evaluation is provided as Enclosure 3.

Accordingly, the NRC staff has also prepared a non-proprietary version of the safety evaluation, which is provided as Enclosure 4.

A Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Sujata Goetz, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Docket Nos. 50-334 and 50-412

Enclosures:

1. Amendment No. 322 to DPR-66

2. Amendment No. 212 to NPF-73

3. Safety Evaluation (Proprietary)

4. Safety Evaluation (Non-Proprietary)

cc: Listserv without Enclosure 3

OFFICIAL USE ONLY PROPRIETARY INFORMATION ENERGY HARBOR NUCLEAR CORP.

ENERGY HARBOR NUCLEAR GENERATION LLC

DOCKET NO. 50-334

BEAVER VALLEY POWER STATION, UNIT NO. 1

AMENDMENT TO RENEWED FACILITY OPERATING LICENSE

Amendment No. 322 Renewed License No. DPR-66

1. The U.S. Nuclear Regulatory Commission (the Commission) has found th at:

A. The application for amendment by Energy Harbor Nuclear Corp.* acting on its own behalf and as agen t for Energy Harbor Nuclear Generation LLC (the licensees), dated August 31, 2022, complies with the standards and req uirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commi ssions rules and regulations set forth in 10 CFR Chapte r 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 am endment can be conducted without endangering the health and safety of the publi c, and (ii) that such activities will be conducted in compliance with the Commissi ons regulations;

D. The issuance of this amendment will not be inimical to the common defen se 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 sa tisfied.

• Energy Harbor Nuclear Corp. is authorized to act as agent fo r Energy Harbor Nuclear Generation LLC and has exclu sive responsibility and control over the physical construc tion, operation, and maintenance of the facility.

Enclosure 1

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2. Accordingly, the license is amended by changes to the TS as indicated in the attachment to this license amendme nt, and paragraph 2.C.(2) of Renewed Facilit y Operating License No. DPR-66 is hereby amended to read as follows:

(2) Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 322, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Hipólito J. González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications

Date of Issuance: October 2, 2023 ENERGY HARBOR NUCLEAR CORP.

ENERGY HARBOR NUCLEAR GENERATION LLC

DOCKET NO. 50-412

BEAVER VALLEY POWER STATION, UNIT 2

AMENDMENT TO RENEWED FACILITY OPERATING LICENSE

Amendment No. 212 Renewed License No. NPF-73

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

A. The application for amendment by Energy Harbor Nuclear Corp.* acting on its own behalf and as agent for Energy Harbor Nuclear Generation LLC (the licensees), dated August 31, 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.

• Energy Harbor Nuclear Corp. is authorized to act as agent for Energy Harbor Nuclear Generation LLC and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility.

Enclosure 2

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2. Accordingly, the license is amended by changes to the Technical Specif ications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-73 is hereby amended to read as follo ws:

(2) Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 212, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. Energy Harbor Nuclear Corp. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

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

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Hipólito J. González, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications

Date of Issuance: October 2, 2023 ATTACHMENT TO LICENSE AMENDMENT NOS. 322 AND 212

BEAVER VALLEY POWER STATION, UNITS 1 AND 2

RENEWED FACILITY OPERATING LICENSE NOS. DPR-66 AND NPF-73

DOCKET NOS. 50-334 AND 50-412

Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.

Renewed Facility Operating License No. DPR-66

Remove Insert Page 3 Page 3

Renewed Facility Operating License No. NPF-73

Remove Insert Page 4 Page 4

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.

Appendix A, Technical Specifications

Remove Insert 4.0-1 4.0-1

5.6-2 5.6-2 5.6-3 5.6-3 5.6-4 5.6-4 5.6-5 5.6-5 (3) Energy Harbor Nuclear Corp., pursuant to the Act and 10 CFR Parts 30, 40 and 70, 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) Energy Harbor Nuclear Corp., 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;

(5) Energy Harbor Nuclear Corp., 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 operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1:

Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; 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

Energy Harbor Nuclear Corp. is authorized to operate the facility at a steady state reactor core power level of 2900 megawatts thermal.

(2) Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 322, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Auxiliary River Water System

(Deleted by Amendment No. 8)

Amendm ent No. 322 Beaver Valley Unit 1 Renewed Operating License DPR-66

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C. This renewed operating license shall be deemed to contain an d is subject to the conditions specified in the following Commission regulations set fort h in 10 CFR Chapter 1 and is subject t o 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

Energy Harbor Nuclear Corp. is authorized to operate the facilit y at a steady state reactor core power level of 2900 megawatts thermal.

(2) Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 212, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. Energy Harbor Nuclear Corp. shall operate t he facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Amendment No. 212 Beaver Valley Unit 2 Renewed Operating License NPF-73 Design Features 4.0

4.0 DESIGN FEATURES

4.1 Site Location

The Beaver Valley Power Station is located in Shippingport Borough, Beaver County, Pennsylvania, on the south bank of the Ohio River. The site is approximately 1 mile southeast of Midland, Pennsylvania, 5 miles east of East Liverpool, Ohio, and approximately 25 miles northwest of Pittsburgh, Pennsylvania. The Unit 1 exclusion area boundary has a minimum radius of 2000 feet from the center of containment. The Unit 2 exclusion area boundary has a minimum radius of 2000 feet around the Unit No. 1 containment building.

4.2 Reactor Core

4.2.1 Fuel Assemblies

The reactor shall contain 157 fuel assemblies. Each assembly shall consist of a matrix of ZIRLO or Optimized ZIRLO' clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO 2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used.

Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Rod Assemblies

The reactor core shall contain 48 control rod assemblies. The control material shall be silver indium cadmium as approved by the NRC.

4.3 Fuel Storage

4.3.1 Criticality

4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment as specified in LCO 3.7.14, "Spent Fuel Pool Storage,"

b. Unit 1 Keff d 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.12 of th e UFSAR,

Beaver Valley Units 1 and 2 4.0 - 1 Amendments /

Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.3 CORE OPERATING LIMITS REPORT (COLR) (continued)

LCO 3.1.5.1, "Unit 1 Shutdown Bank Insertion Limits" LCO 3.1.5.2, Unit 2 Shutdown Bank Insertion Limits LCO 3.1.6.1, "Unit 1 Control Bank Insertion Limits" LCO 3.1.6.2, Unit 2 Control Bank Insertion Limits LCO 3.2.1, "Heat Flux Hot Channel Factor (F Q(Z))"

LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor ( NF' )" H

LCO 3.2.3, "Axial Flux Difference (AFD)"

LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation" - Overtemperature and Overpower ' T Allowable Value parameter values LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits" LCO 3.9.1, "Boron Concentration"

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology,"

WCAP-8745-P-A, "Design Bases for the Thermal Overtemperature ' T and Thermal Overpower ' T Trip Functions,"

WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), November 2016,

WCAP-12945-P-A, Volumes 1 through 5, "Code Qualification Document for Best Estimate LOCA Analysis," [Shall not be used to determine core operating limits after December 2024]

(For Unit 1 only) WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)," [Shall not be used to determine core operating limits after December 2024]

WCAP-10216-P-A, "Relaxation of Constant Axial Offset Control/

FQ Surveillance Technical Specification,"

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

WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report,"

WCAP-15025-P-A, "Modified WRB-2 Correlation, WRB-2M, for Predicating Critical Heat Flux in 17x17 Rod Bundles with Modified LPD Mixing Vane Grids,"

Beaver Valley Units 1 and 2 5.6 - 2 Amendments  

Reporting Requirements 5.6 5.6 Reporting Requirements

5.6.3 CORE OPERATING LIMITS REPORT (COLR) (continued)

WCAP-13749-P-A, Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement, March 1997 (Westinghouse Proprietary),

WCAP-16045-P-A, Qualification of the Two-Dimensional Transport Code PARAGON, WCAP-16045-P-A, Addendum 1-A, Qualification of the NEXUS Nuclear Data Methodology,

WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLO',

WCAP-17661-P-A, Improved RAOC and CAOC FQ Surveillance Technical Specifications.

As described in reference documents listed above, when an initial assumed power level of 102% of RATED THERMAL POWER is specified in a previously approved method, 100.6% of RATED THERMAL POWER may be used when input for reactor thermal power measurement of feedwater flow is by the leading edge flow meter (LEFM) or feedwater venturi normalized to a prior LEFM flow measurement.

Caldon, Inc. Engineering Report-80P, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM TM System"

Caldon, Inc. Engineering Report-160P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM TM System"

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing, Overpressure Protection System (OPPS) enable temperature, and PORV lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

LCO 3.4.3, "RCS Pressure and Temperature (P/T) Limits," and LCO 3.4.12, "Overpressure Protection System (OPPS)"

Beaver Valley Units 1 and 2 5.6 - 3 Amendments  

Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.4 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued)

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, May 2004.

WCAP-18124-NPA, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, July 2012, may be used as an alternative to Section 2.2 of WCAP-14040-A, Revision 4.

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

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

5.6.6 Steam Generator (SG) Tube Inspection Report 5.6.6.1 Unit 1 SG Tube Inspection Report

A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.5.1, Unit 1 SG Program. The report shall include:

a. The scope of inspections performed on each SG;

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

c. For each degradation mechanism found:

1. The nondestructive examination techniques utilized;

2. The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;

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

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

Beaver Valley Units 1 and 2 5.6 - 4 Amendments  

Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.6 Steam Generator (SG) Tube Inspection Report (continued)

5.6.6.1 Unit 1 SG Tube Inspection Report (continued)

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

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

f. The results of any SG secondary side inspections.

5.6.6.2 Unit 2 SG Tube Inspection Report

1. A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.5.2, Unit 2 SG Program. The report shall include:

a. The scope of inspections performed on each SG;

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

c. For each degradation mechanism found:

1. The nondestructive examination techniques utilized;

2. The location orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;

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

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

5. The repair methods utilized and the number of tubes repaired by each repair method.

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;

Beaver Valley Units 1 and 2 5.6 - 5 Amendments  

OFFICIAL USE ONLY PROPRIETARY INFORMATION

NON-PROPRIETARY SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR

REGULATION RELATED TO AMENDMENT NOS. 322 AND 212 TO RENEWED

FACILITY OPERATING LICENSE NOS. DPR-66 AND NPF-73

ENERGY HARBOR NUCLEAR GENERATION LLC

BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2

DOCKET NOS. 50-334 AND 50-412

1.0 INTRODUCTION

By letter dated August 31, 2022, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22249A257), Energy Harbor Nuclear Corp. (the licensee) submitted a license amendment request (LAR) to the renewed facility operating licenses numbered DPR-66 and NPF-73 for the Beaver Valley Power Station, Unit Nos. 1 and 2, respectively.

The amendment requests to revise Beaver Valley technical specifications (TSs) 5.6.3, Core Operating Limits Report (COLR) by adding the Westinghouse Electric Company LLC, Topical Report WCAP-16996-P-A, Rev.1, Realistic LOCA [loss of coolant accident] Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), (ML17277A132) to the list of approved analytical methods used to determine the core operating limits and by adding a note to the loss of coolant accident methods listed to restrict their future use. The proposed amendment also removes Zircalloy from the list of fuel rod cladding in TS 4.2.1, Fuel Assemblies.

1.1 TSs Changes

1.1.1 Current Technical Specification Requirements

In the LAR, Attachment 1, Section 2.2 Current Technical Specification Requirements, the licensee states:

TS 5.6.3 requires, in part, core operating limits be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and [it] contains references to the approved analytical methods that are used to determine the core operating limits. The current methods listed in TS 5.6.3.b for LOCA analyses are WCAP-12945-P-A, Volumes 1 through 5, Code Qualification Document for

Enclosure 4

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Best Estimate LOCA Analysis; WCAP-16009-P-A, Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM).

In the LAR, Section 2.4 Description of the Proposed Change the licensee states:

TS 4.2.1 currently states, in part, that each fuel assembly shall consist of a matrix of Zircalloy, ZIRLO, or Optimized ZIRLO' clad fuel rods.

1.1.2 Proposed Changes

In the LAR, Section 2.4 Description of the Proposed Change the licensee states:

TS 5.6.3.b currently includes approved analytical methods that would no longer be used to support BVPS [Beaver Valley Power Station] reload cores. The proposed revision would add a note restricting the use of the following legacy analytical methods listed in TS 5.6.3.b as follows:

WCAP-12945-P-A, Volumes 1 through 5, Code Qualification Document for Best Estimate LOCA Analysis, [Shall not be used to determine core operating limits after December 2024]

(For Unit 1 only) WCAP-16009-P-A, Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM), [Shall not be used to determine core operating limits after December 2024]

The proposed revision would add WCAP-16996-P-A to the list of approved methods as follows:

WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology),

November 2016

In addition, the licensee proposed to remove Zircalloy from the list of fuel rod cladding from the reactor core.

2.0 REGULATORY EVALUATION

The NRC staff considered the following regulations and guidance during its review of the proposed changes.

1. Section 182a of the Atomic Energy Act of 1954, as amended, requires applicants for nuclear power plant operating licenses to include TSs as a part of the license. The NRCs regulatory requirements related to the content of TSs are set forth in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.36, which requires that the TSs include items in eight specific categories: (1) safety limits, limiting safety system setting s and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveilla nce requirements (SR); (4) design featur es; (5) administrative controls; (6) de commissioning; (7) initial notification; and (8) written reports.

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2. 10 CFR 50.36(b) states, The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to § 50.34.

3. 10 CFR 50.36(c)(4) states, Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1), (2), and (3) of this section.

4. 10 CFR 50.36(c)(5) states, Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. Each licensee shall submit any reports to the Commission pursuant to approved technical specifications as specified in § 50.4.

5. The following paragraphs of 10 CFR 50.46(b)(1)-(5), requires, in part, that:

(b)(1): Peak cladding temperature - the calculated maximum fuel element cladding temperature shall not exceed 2200 degrees °F [Fahrenheit].

(b)(2): Maximum cladding oxidation - the calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

(b)(3): Maximum hydrogen generation - the calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

(b)(4): Coolable geometry - calculated changes in core geometry shall be such that the core remains amenable to cooling.

(b)(5): Long-term cooling - after any calculated successful initial operation of the ECCS [emergency core cooling system], the calculated core temperature shall be maintained at an acceptably low value.

6. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, Section 15.6.5, Revision 3, Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary, dated March 2007 (ML070550016).

7. NRC Regulatory Guide (RG) 1.157 (Task RS 701-4), Best-Estimate Calculations of Emergency Core Cooling System Performance, dated May 1989 (ML003739584).

8. NRC RG 1.203, Transient and Accident Analysis Methods, dated December 2005 (ML053500170).

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9. NRC Generic Letter 88-16, Removal of Cycle-Specific Parameter Limit s from Technical Specifications, (Generic Letter 88-16) dated October 4, 1988 (ML031200485).

3.0 TECHNICAL EVALUATION

3.1 Nuclear Systems Evaluation

The NRC staff evaluated the licensees August 31, 2022, LAR to determine whether the proposed changes are in compliance with the regulations and consistent with the guidance in Section 2.0 of this safety evaluation. The NRC staff reviewed the proposed changes and verified that (1) the new LOCA evaluation methodology (EM) is an NRC-approved code, (2) that all limitations and conditions for FSLOCA EM will be met, (3) the licensee will be appropriately applying the LOCA EM to Beaver Valley, Units 1 and 2, and (4) that the results meet the acceptance criteria in 10 CFR 50.46(b)(1) through (5).

3.2 Removal of Zircalloy cladding

The licensee proposed to remove Zircalloy from the list of fuel rod cladding from TS 4.2.1.

Throughout the industry, ZIRLO and Optimized ZIRLO have replaced Zircalloy as the alloy of choice in PWRs and as a result, Zircalloy is not supported by the FSLOCA EM analysis, nor will it be utilized in future core designs. Beaver Valley fuel will not use Zircalloy cladding. The removal of Zircalloy cladding from the list of fuel rod cladding in TS 4.2.1, is an administrative change that has no technical impact on the ability to meet COLR limits, and as a result, the safety of plant operations remains unaffected by the proposed change. Based upon the information above and its review, the NRC staff finds that the removal of Zircalloy from the licensees list of cladding in their TS is acceptable.

3.3 Description of FULL SPECTRUM'LOCA (FSLOCA) Methodology

As described in Section 1, Introduction of the WCAP-16996-P-A, Revision 1, the purpose of the FSLOCA EM is to build on the previous evaluation methods by extending the applicability of the computer code, WCOBRA/TRAC to include small break loss of coolant accident (SBLOCA) and intermediate break LOCA (IBLOCA). The term Full Spectrum specifies that the new evaluation method is intended to resolve all the scenarios that result from a postulated break in the cold leg of a pressurized water reactor (PWR). The break sizes considered in the Westinghouse FSLOCA EM includes any break size in which break flow is beyond the capacity of the normal charging pumps, up to and including a double ended guillotine rupture with a break flow area equal to two times the pipe area.

In the LAR, Section 3.0, Technical Evaluation, the licensee stated that two separate analyses with the FSLOCA EM were performed for Beaver Valley Units 1 and 2 due to plant design differences. The Beaver Valley, Unit 1 reactor vessel has thermal shields, and the Beaver Valley, Unit 2 reactor vessel has neutron pads. In addition, the steam generator designs of the two units are different.

In the LAR, Enclosure C, Attachment 1 of LTR-LIS-21-67, Suggested Technical Evaluation Section of the Beaver Valley Power Station Unit 1 and 2 LAR Input (Non-Proprietary),

Section 1.0, Introduction states that The major plant parameter and analysis assumptions used in the Beaver Valley Unit 1 analysis with the FSLOCA EM are provided in Tables 1a

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through 6a. The major plant parameter and analysis assumptions used in the Beaver Valley Unit 2 analysis with the FSLOCA EM are provided in Tables 1b through 6b.

3.4 Region 1 Analysis (small break LOCA)

3.4.1 Description of the Transient

As described in the Suggested Technical Evaluation Section of the Beaver Valley, Unit 1 and Unit 2 LAR Input, Section 3.1, Description of Representative Transient, a small break LOCA undergoes the following phases, as summarized below:

1. Blowdown: Following the reactor trip on the low pressurizer pressure setpoint, the safety injection (SI) is initiated on the low pressurizer pressure SI setpoint.

RCS [Reactor Coolant System] eventually reaches saturation, ending the rapid depressurization slightly above the steam generator secondary side pressure.

2. Natural Circulation: [In this phase] the RCS pressure remains slightly above the secondary side pressure. The core remains covered, and the fuel cladding temperatures remain at the saturation temperature level.

3. Loop Seal Clearance: The liquid levels in the crossover leg become depressed all the way to the bottom elevations of the piping, allowing the steam trapped to vent to the break.

4. Boil-Off: In this phase the RCS pressure remains high enough such that safety injection flow cannot make up for the primary system fluid inventory lost through the break, leading to core uncovery and a fuel rod cladding temperature heatup.

5. Core Recovery: The RCS pressure continues to decrease, and once it reaches that of the accumulator gas pressure, the ECCS water from the accumulators replenishes the reactor vessel and recovers the core mixture level. The transient is considered over as the break flow is compensated by the injected flow.

3.4.2 NRC Staff Evaluation

In the LAR, Enclosure C, Suggested Technical Evaluation Section of the Beaver Valley Power Station Unit 1 and Unit 2 LAR Input, Section 3.2, Analysis Results states that the Beaver Valley Units 1 and 2 Region I analyses were performed in accordance with the NRC-approved methodology. NRC staff reviewed Table 1a Plant Operating Range Analyzed and Key Parameters for Beaver Valley Unit 1 and Table 1b Plant Operating Range Analyzed and Key Parameters for Beaver Valley Unit 2 and confirmed that the most limiting ECCS single failure of one ECCS train is assumed in both units analyses. For both units, control rod drop is modeled for breaks less than one square foot assuming a two second signal delay time and a 2.7 second rod drop time.

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The Suggested Technical Evaluation contains the results of the Region I uncertainty analysis in Table 7a Beaver Valley Unit 1 Analysis Results with the FSLOCA EM and Table 7b Beaver Valley Unit 2 Analysis Results with the FSLOCA EM. The NRC staff reviewed the analysis-of-record peak clad temperature (PCT) provided in Tables 7a and 7b. The NRC staff also conducted a regulatory audit (ML23124A388) to support its review of the LAR. Based upon its review, the NRC staff determined that the analyses assumed conservative input assumptions and, therefore, are acceptable.

3.5 Region 2 Analysis (large break LOCA)

3.5.1 Description of the Transient

As described in Attachment 1, Suggested Technical Evaluation Section of the Beaver Valley Power Station Unit 1 and Unit 2 LAR Input, Section 4.1, Description of Representative Transient, a large break LOCA undergoes the following phases:

1. Blowdown - Critical Heat Flux (CHF) Phase

In this phase, the break flow is subcooled, the discharge rate of coolant from the break is high, the core flow reverses, the fuel rods go through departure from nucleate boiling (DNB), and the reactor is shut down due to the core voiding.

2. Blowdown - Upward Core Flow Phase

Heat transfer is increased as the two-phase mixture is pushed into the core.

The break discharge rate is reduced because the fluid becomes saturated at the break. This phase ends as the lower plenum mass is depleted, the fluid in the loops become two-phase, and the RCP head degrades.

3. Blowdown - Downward Core Flow Phase

During this period, due to steam upflow in the downcomer, a portion of the injected ECCS water is bypassed around the downcomer and sent out through the break. As the system pressure continues to decrease, the break flow and consequently the downward core flow are reduced. The system pressure approaches the containment pressure at the end of this last period of the blowdown phase.

During this phase, the core begins to heat up as the system approaches containment pressure, and the phase ends when the reactor vessel begins to refill with ECCS water.

4. Refill Phase

The core continues to heat up as the lower plenum refills with ECCS water.

As ECCS water enters the core, the fuel rods in the lower core region begin to quench and liquid entrainment begins, resulting in increased fuel rod heat transfer.

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5. Reflood Phase:

During the early reflood phase, the accumulators begin to empty, and nitrogen is discharged into the RCS.

3.5.2 NRC Staff Evaluation

The NRC staff review of Tables 1a and 1b provided in the Suggested Technical Evaluation of the LAR indicated that the most conservative ECCS single failure of one train is assumed in analyses of both Units. The NRC staff reviewed the analysis-of-record PCT provided in Tables 7a and 7b and confirmed that Beaver Valley, Units 1 and 2, Region II analyses were also performed by the licensee using the NRC-approved FSLOCA EM. The NRC staff also conducted a regulatory audit (ML23124A388) to support its review of the LAR. Based upon its review, the NRC staff found that the analyses assumed conservative input assumptions and, therefore, are acceptable.

3.6 Compliance with 10 CFR 50.46

3.6.1 Compliance with 10 CFR 50.46(b)(1) through (b)(3)

The NRC staff reviewed the results in Tables 7a (Unit 1) and Table 7b (Unit 2) in the Suggested Technical Evaluation of the LAR, for the peak clad temperature, maximum local oxidation, and core-wide oxidation and determined there is an acceptable margin relative to the regulatory requirements in 10 CFR 50.46(b)(1), (b)(2), and (b)(3), as discussed below.

The data provided in Table 7a and 7b indicates that the peak cladding temperature reached is 2001 °F for Region II, and thus does not exceed 2200 °F and is, therefore, in compliance with 10 CFR 50.46(b)(1).

Tables 7a and 7b indicate that the highest maximum cladding oxidization level reached at Beaver Valley is 12.6 percent at Unit 1, which is below the 17 percent required. Therefore, Beaver Valley Unit 1 and 2 is in compliance with 10 CFR 50.46(b)(2).

Table 7a and 7b indicate that the highest oxidation level is in Region II of Unit 2, at 92 percent, which is less than the 1 percent required. Therefore, Beaver Valley Unit 1 and 2 is in compliance with 10 CFR 50.46(b)(3).

Based upon its review of the above, the NRC staff finds that the licensee is in compliance with 10 CFR 50.46(b)(1) through (b)(3).

3.6.2 Compliance with10 CFR 50.46(b)(4) Coolable geometry

The requirement in 10 CFR 50.46(b)(4) is that core geometry can be cooled. This criterion is met if the peak clad temperature does not exceed 2200 degrees, as required by (b)(1) and the maximum local oxidation does not exceed 17 percent as required by (b)(2).

The NRC staff reviewed Table 7a and 7b in the Suggested Technical Review and determined that coolable core geometry is maintained because the inboard grid deformation due to a combined LOCA and seismic loads is not expected and is not calculated to occur for Beaver Valley, Unit 1 and Unit 2. This is consistent with Section 32.1 of the NRC-approved FSLOCA

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EM, Volume III, which states that the effects of a LOCA and seismic loads on the core geometry do not need to be considered unless fuel assembly grid deformation extends beyond the core periphery.

Based upon its review, as discussed above and in section 3.6.1 of this safety evaluation, the NRC staff finds that the licensee is in compliance with 10 CFR 50.46(b)(4) because coolable core geometry is maintained.

3.6.3 Compliance with 10 CFR 50.46(b)(5) Long term cooling

For compliance with 10 CFR 50.46(b)(5), the licensee did not utilize the FSLOCA EM because it is not NRC-approved for the long-term analyses. Therefore, in regard to continued compliance with 10 CFR 50.46(b)(5), the NRC staff finds that current licensing basis is maintained and is not affected by the implementation of the FSLOCA EM because the licensee did not use the FSLOCA EM for long-term analysis.

3.6.4 Compliance with 10 CFR 50.46 Conclusion

Based on the considerations above, the NRC staff finds the licensee has satisfied the requirements of 10 CFR 50.46(b)(1) through (b)(5).

3.7 Compliance with FSLOCA EM Limitations and Conditions

The safety evaluation for WCAP-16996-P-A, Revision 1 (ADAMS Package Accession No. ML17207A124) contains 15 limitations and conditions that must be met to implement the NRC-approved FSLOCA EM.

A summary of each limitation and condition and how it has been met by the licensee is provided in Enclosure A, Attachment 1 of LTR-LIS-21-67, Suggested Technical Evaluation Section of the Beaver Valley Power Station Unit 1 and Unit 2 LAR Input (Proprietary) application dated August 31, 2022, (Suggested Technical Evaluation), Section 2.3 Compliance with FSLOCA EM Limitations and Conditions.

3.7.1 Limitation and Condition 1 - FSLOCA EM Applicability with Regard to LOCA Transient Phases

The licensee stated that the FSLOCA EM is not approved to demonstrate compliance with 10 CFR 50.46(b)(5) for long term cooling. Furthermore, the licensee stated that the Beaver Valley, Units 1 and 2 analyses performed with the FSLOCA EM are only being used to demonstrate compliance with 10 CFR 50.46 (b)(1) through (b)(4).

Given that the licensee is not using the FSLOCA EM to demonstrate compliance with 10 CFR 50.46(b)(5), the NRC staff finds that the licensee has met the requirements of Limitation and Condition 1.

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3.7.2 Limitation and Condition 2 - FSLOCA EM Applicability with Regard to Type of PWR Plants

The licensee provided, in part, the following summary of Limitation and Condition 2:

Summary

The FSLOCA EM is approved for the analysis of Westinghouse-designed 3-loop and 4-loop PWRs with cold-side injection. Analyses should be executed consistent with the approved method, or any deviations from the approved method should be described and justified.

Compliance

The licensee stated that the Beaver Valley Unit 1 and Unit 2 are Westinghouse-designed 3-loop PWRs with cold-side injection, so they are within the NRC-approved methodology. The staff review indicates that the analyses for Beaver Valley Unit 1 and Unit 2 utilized the NRC-approved FSLOCA methodology, with the following three exceptions identified in the LAR: (1) the changes which were previously transmitted to the NRC pursuant to 10 CFR 50.46 in Westinghouse Letter dated July 18, 2018, LTR-NRC-18-30 [ML19288A174] were incorporated into the analyses, (2) the modeled fuel average temperatures bound the PAD5 data, and (3) the blowdown energy release assumption was modified to use a plant specific bounding value.

The NRC staff finds that the use of the FSLOCA EM is acceptable for Beaver Valley because it is a 3-loop PWR with cold-side injection.

Exceptions

The NRC staff reviewed each of the three exceptions to the FSLOCA EM taken by the licensee and the evaluation is provided below.

Exception 1: Changes previously transmitted to the NRC pursuant to 10 CFR 50.46 in LTR-NRC-18-30.

The NRC staff finds this exception acceptable because the staff was made aware of the changes, and the impact to the analysis was determined by the vendor to be zero to negligible degrees for PCT by LTR-NRC-18-30. LTR-NRC-18-30 was provided to the NRC because the changes affect information previously detailed in the Westinghouse FSLOCA topical report.

Changes include discretionary general code maintenance and non-discretionary changes; however, as stated above, the impact to the FSLOCA analysis was determined by the vendor to be zero to negligible degrees for PCT.

Exception 2: Modeled fuel average temperatures bound the PAD5 data.

The licensee used a plant specific value for modeled fuel average temperature in their FSLOCA analysis instead of the input that was determined from PAD5. The licensee stated that the values used bound the PAD5 data. Based upon its review, the NRC staff finds this exception acceptable because the use of plant specific values bound the generic data.

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Exception 3: Plant specific bounding blowdown energy release assumption.

The licensee stated that they used a plant specific blowdown energy release assumption value in their FSLOCA analysis. Based upon its review, the NRC staff finds this approach acceptable because use of the plant specific value is bounding for Beaver Valley and within the bounds of the established FSLOCA EM.

The LAR stated that after completion of the Beaver Valley FSLOCA analyses, the licensee evaluated the impact of changes and errors to the model, as described in the U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2018, (LTR-NRC-19-6, ML19042A380) dated February 2019, and determined that there is a zero to negligible impact on degrees PCT. Since there is a negligible impact on the analysis and the NRC staff was informed of these changes and errors pursuant to 10 CFR 50.46, the NRC staff finds that the licensee has acceptably addressed the errors as described in LTR-NRC-19-6.

Additionally, the LAR provided information and analysis for the gamma energy redistribution error and the equation for the assumed increase in hot rod and hot assembly relative power.

The licensee evaluated the impact of the equation error to the FSLOCA EM and found that there was a 0%-5% underestimation of the modeled hot rod and hot assembly rod linear heat rates on a run-specific basis, depending on the as-sampled value for the uncertainty. The licensee stated that the error correction has only a limited impact on the power modeled for a single assembly in the core. As such, there is a negligible impact of the error correction on the system thermal-hydraulic response during the postulated LOCA. The NRC staff finds that the effect of the equation error for the assumed increase in hot rod and hot assembly relative power has been acceptably addressed because there is a negligible impact on the FSLOCA EM.

In response to the gamma energy redistribution error, the licensee ran sensitivity studies to determine the impact on the FSLOCA EM. The licensee stated that for Region I, the primary impact of the error correction is on the rate of cladding heat up above the two-phase mixture level in the core during the boiloff phase. The impact of this error for Region I was 1°F PCT for Unit 1 and 5°F PCT for Unit 2. The licensee added the identified PCT error to the FSLOCA EM values for both the Unit 1 and Unit 2 analyses to determine the final Region I PCT and, therefore, the NRC staff finds the gamma energy redistribution error has been acceptably addressed as it relates to the Region I analyses and remains within the 10 CFR 50.46 requirements.

The licensee also ran sensitivity studies to determine the gamma energy redistribution error for the Region II analyses. The sensitivity studies were derived from a subset of uncertainty analysis simulations covering various design features and fuel arrays. The licensee stated that, the PCT impact from the error correction was found to be different for the different transient phases (i.e., blowdown versus reflood) based on the PWR sensitivity studies and existing power distribution sensitivity studies. The licensees sensitivity studies determined that there is an estimated 31 °F PCT error for the Region II analysis both for assuming offsite power available and loss of offsite power. The licensee added the identified PCT error to the FSLOCA EM values for both Unit 1 and Unit 2 analyses to determine the final Region II PCT and, therefore, the NRC staff found the gamma energy redistribution error has been acceptably addressed as it relates to the Region II analyses and remains within the 10 CFR 50.46 requirements.

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Based upon its review, as discussed above, the NRC staff finds the licensee has acceptably addressed all exceptions and errors associated with the FSLOCA methodology and, therefore, has met the requirements of Limitation and Condition 2.

3.7.3 Limitation and Condition 3 - FSLOCA EM Applicability for Containment Pressure Modeling

The NRC staffs review of the LAR indicates that the containment pressure calculation for the Beaver Valley, Unit 1 and 2 analyses were performed by the licensee consistent with the NRC-approved methodology. The licensee stated that the containment pressure is calculated for each LOCA transient in the analyses using the Containment Pressure Analysis Code.

Additionally, the licensee stated that appropriate design parameters and conditions were modeled in the analyses.

The NRC staff review determined that the licensee correctly used the NRC-approved methodology and appropriate input assumptions for the containment pressure calculations.

Therefore, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 3.

3.7.4 Limitation and Condition 4 - Decay Heat Modeling in FSLOCA EM Applications

The licensee analyzed the decay heat uncertainty multiplier by ((

)). The simulation performed by the licensee for the FSLOCA EM was not executed for longer than 10,000 seconds following reactor trip to ensure the decay heat model is appropriately justified. The sampled values of the decay heat uncertainty multiplier for the cases which produced the Region I and Region II analysis results were provided in the Suggested Technical Evaluation, which stated:

Consistent with the NRC-approved methodology, the decay heat uncertainty multiplier was ((

)) for the Beaver Valley Unit 1 and Unit 2 analyses. The analysis simulations were all executed for no longer than 10,000 seconds following reactor trip. The sampled values of the decay heat uncertainty multiplier for the cases which produced the Region I and Region II analysis results have been provided in units of sigma and approximate absolute units in Table 10a for Unit 1 and Table 10b for Unit 2.

The NRC staff finds that the licensee appropriately modeled decay heat and reported the resulting sampled values in units of sigma and absolute units for the limiting cases appropriately and therefore, has met the requirements of Limitation and Condition 4.

3.7.5 Limitation and Condition 5 - Fuel Burnup Limits in FSLOCA EM Applications

The LAR stated that for the FSLOCA EM, the maximum assembly and rod length-average burnup is limited to ((

)) respectively. For Beaver Valley, the maximum analyzed assembly and rod length-average burnup was less than or equal to ((

)).

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The NRC staff found Beaver Valleys burnup limit was below the maximum allowed by the FSLOCA EM. Based on the information provided in the LAR and the staff review of calculations performed, the NRC staff found that the licensee has met the requirements of Limitation and Condition 5.

3.7.6 Limitation and Condition 6 - WCOBRA/TRAC-TF2 Interface with PAD5 in the FSLOCA EM

This limitation and condition requires that the fuel performance data for the FSLOCA EM analyses should be based on the latest version of an NRC-approved fuel performance code.

The licensee stated in its LAR that it is using PAD5, which is the latest available version of NRC-approved fuel performance code. In addition, the licensee stated that the analyzed fuel pellet average temperatures are bounding compared to the upper bound maximum values calculated in accordance with Section 7.5.1 of PAD5 and the analyzed rod internal pressures were calculated in accordance with Section 7.5.2 of PAD5....

The NRC staff finds that the licensee has met the requirements of Limitation and Condition 6 because the licensee used the latest NRC-approved fuel performance code (i.e., PAD5) and appropriate conservative inputs.

3.7.7 Limitation and Condition 7 - Interfacial Drag Uncertainty in FSLOCA EM Region I Analyses

The licensee stated that the FSLOCA methodology requires that YDRAG uncertainty parameter should be ((

)) and that they performed their analyses consistent with this requirement of the NRC-appr oved FSLOCA methodology.

The NRC staff confirmed that the licensee appropriately used the specified interfacial drag uncertainty parameter. Therefore, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 7.

3.7.8 Limitation and Condition 8 - Biased Uncertainty Contributors in FSLOCA EM Region I Analyses

The licensee stated that the FSLOCA methodology requires that ((

)) and that they performed their analyses consistent with this requirement of the NRC-approved FSLOCA methodology.

The NRC staff confirmed that the licensee appropriately used the specified biased uncertainty parameters and, therefore, met the requirement of Limitation and Condition 8.

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3.7.9 Limitation and Condition 9 - Effect of Bias in FSLOCA EM Applications for Region I

The licensee stated in its LAR that for PWRs that are not Westinghouse 3-loop reactors, a sensitivity study should be performed to confirm that the ((

)).

The NRC staff found that Limitation and Condition 9 is not applicable, because Beaver Valley 1 and 2 are Westinghouse-designed 3-loop PWRs, and therefore a sensitivity study is not needed.

3.7.10 Limitation and Condition 10 - Boundary between FSLOCA EM Region I and Region II Breaks

Limitation and Condition 10 requires a sensitivity study to address certain safety issues for reactors that are not Westinghouse 3-loop PWRs. The NRC staff finds that Limitation and Condition 10 is not applicable because Beaver Valley, Unit 1 and Unit 2 are Westinghouse-designed 3-loop PWRs, and therefore, no sensitivity study was required.

3.7.11 Limitation and Condition 11 - (( )) in FSLOCA EM Uncertainty Analyses for Region II and Documentation of Reanalysis Results for Region I and Region II

The licensee summarized Limitation and Condition 11 as follows:

1. The (( )), the Region I and Region II analysis see ds, and the analysis inputs will be de clared and documented prior to performing the Region I and Region II uncertaint y analysis. The (( )) and the Region I and Region II analysis seeds will not be changed throughout the remainder of the analysis once they have been declared and documented.

2. If the analysis inputs are changed after they have been declared and docum ented, for the intended purpose of demonstrating compliance with the applicable a cceptance criteria, then the changes and associated rationale for the changes will be provided in the analysis submittal. Additionally, the preliminary values for PCT [peak clad temperature], MLO [maximum local oxidation], and CWO [core-wide oxid ation] which caused the input changes will be provided. These preliminary values are not subject to Appendix B verification, and archival of the supporting information for these prelimina ry values is not required.

3. Plant operating ranges which are sampled within the uncert ainty analysis will be provided in the analysis submittal for both regions.

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The licensee described how Limitation and Condition 11 was met for the Beaver Valley Unit 1 and Unit 2 analyses as follows:

1. The (( )) Region I and Region II analysis seeds, and th e analysis inputs were declared and documented prior to performing the Region I and Region II uncertainty analyses.

The (( )) and the Region I and Region II analysis see ds were not changed once they wer e declared and documented.

2. The analysis inputs were not changed once they were declared and documented.

3. The plant operating ranges which were sampled within the uncertainty a nalysis are provided for Beaver Valley Unit 1 in Table 1a and for Unit 2 in Table 1b.

Based on the NRC staffs review of the information provided by the licensee, the NRC staff finds that Limitation and Condition 11 has been met.

3.7.12 Limitation and Condition 12 - Steam Generator Heat Removal During SBLOCAs

The licensee stated that Limitation and Condition 12 requires that the plant-specific dynamic pressure loss from the steam generator secondary-side to the main steam safety valves must be adequately accounted for in analysis with the FSLOCA EM. The licensee stated that a bounding plant-specific dynamic pressure loss from the steam generator secondary-side to the main steam safety valves was modeled in the Beaver Valley Unit 1 and Unit 2 analyses.

Given the licensees use of a bounding dynamic pressure loss, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 12.

3.7.13 Limitation and Condition 13 - Upper Head Spray Nozzle Loss Coefficient

The LAR summarized Limitation and Condition 13:

In plant specific models for analysis with the FSLOCA EM: 1) the ((

)) and 2) the ((

)).

To meet the requirements of Limitation and Condition 13, the licensee stated that they modeled the upper head spray nozzle loss coefficient as required by FSLOCA EM and that the ((

)).

The NRC staff confirmed that the licensee modeled the upper head spray nozzle loss coefficient as required by the FSLOCA EM and, therefore, has met the requirements of Limitation and Condition 13.

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3.7.14 Limitation and Condition 14 - Correlation for Oxidation

The licensee summarized Limitation and Condition 14:

For analyses with the FSLOCA EM to demonstrate compliance against the current 10 CFR 50.46 oxidation criterion, the transient time-at-temperature will be converted to an equivalent cladding reacted (ECR) using either the Baker-Just or the Cathcart-Pawel correlation. In either case, the pre-transient corrosion will be

[summed with] the LOCA transient oxidation. If the Cathcart-Pawel correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 13 percent limit. If the Baker-Just correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 17 percent limit.

The licensee stated that for the Beaver Valley, Unit 1 and Unit 2 analyses, the Baker-Just correlation was used in each transient calculation to convert the LOCA transient time-at-temperature to an ECR. The resulting LOCA transient ECR was then added to the pre-existing corrosion for comparison against the 10 CFR 50.46 local oxidation acceptance criterion of 17%.

The NRC staff finds that by using the Baker-Just correlation, converting to an equivalent cladding reacted, and accounting for pre-existing corrosion, the licensee met the requirements of Limitation and Condition 14.

3.7.15 Limitation and Condition 15 - Loss of Offsite Power (LOOP) versus Offsite Power Available (OPA) Treatment in Uncertainty Analysis for Region II

The licensee summarized Limitation and Condition 15 as follows:

Summary

The Region II analysis will be executed twice; once assuming LOOP and once assuming OPA. The results from both analysis executions should be shown to be in compliance with the 10 CFR 50.46 acceptance criteria.

The (( )).

Compliance

The Region II uncertainty analyses for Beaver Valley Unit 1 and Unit 2 were performed twice; once assuming a LOOP and once assuming OPA. The results from both analyses that were performed are in compliance with the 10 CFR 50.46 acceptance criteria.

The ((

)).

The NRC staff reviewed the results of the Region II analysis for LOOP and OPA provided by the licensee and found that they met the requirements of Limitation and Condition 15 and, therefore, are in compliance with the acceptance criteria in 10 CFR 50.46(b)(1) through (b)(4).

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3.8 Technical Conclusion

The licensee proposed to modify TS 5.6.3.b, to r eplace the existing NRC-approved LOCA methodologies with the NRC-approved LOCA methodology contained in WCAP-16996-P-A, Revision 1, FSLOCA EM. The licensee also proposed to remove Zircalloy from TS 4.2.1 because it will no longer be used in future core designs.

Based upon the discussion above, the NRC staff concludes that the proposed changes to TS 5.6.3.b are acceptable because it replaces one set of NRC-approved methods for another NRC-approved methods. The NRC staff confirmed that the licensee appropriately applied the FSLOCA EM to Beaver Valley, Units 1 and 2, and finds that the resulting analysis meets the criteria in 10 CFR 50.46 (b)(1) through (b)(5). In addition, the NRC staff concludes that the removal of Zircalloy from TS 4.2.1 is acceptable as it will no longer be used in reactor cores.

3.9 Evaluation of changes to the Beaver Valley Technical Specifications

The LAR proposes certain changes to Beaver Valley TS sections 4.0, Design Features which is controlled by 10 CFR 50.36(c)(4) that states, Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1)(2) and (3) of this section.

TS 4.2.1 currently states, in part, that each fuel assembly shall consist of a matrix of Zircalloy, ZIRLO, or Optimized ZIRLO' clad fuel rods. The proposed change removes Zircalloy from its TS since, according to the licensee, it will no longer be used in future core designs.

The NRC staffs analysis, as discussed in Section 3.2 of this safety evaluation, concludes that the removal of specific Zircalloy cladding from the list of fuel rod cladding is an administrative change that has no technical impact on the ability to meet COLR limits, and as a result, the safety of plant operations remains unaffected by the proposed change. Therefore, the NRC staff finds that the proposed change is acceptable, and in compliance with the 10 CFR 50.36(c)(4) regulation.

The LAR also proposes certain changes to TS section 5.0, Administrative controls which is controlled by 10 CFR 50.36(c)(5) and states, Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. Each licensee shall submit any reports to the Commission pursuant to approved technical specifications as specified in § 50.4.

The NRC staffs technical review in section 3.0 of this safety evaluation determined that the licensee appropriately applied the FSLOCA EM to Beaver Valley, Units 1 and 2, and found that the resulting analysis meets the criteria in 10 CFR 50.46(b)(1) through (4). Therefore, the NRC staff finds that the proposed changes are acceptable, and in compliance with the 10 CFR 50.36(c)(5) regulation.

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3.10 Conforming Changes to Operating License Pages

Amendments 322 and 212 in this letter approved the insertion of certain text in in TS page 5.6-2, which caused the text at the bottom of that page to shift onto page 5.6-3, which in turn caused text to shift into page 5.6-4, which in turn shifted onto page 5.6-5. The NRC staff determined that shifting of the text into the next consecutive page is an acceptable editorial change that does not change the content or interpretation of the remaining conditions and issued the amendments to reflect this change.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Commonwealth of Pennsylvania official was notified of the proposed issuance of the amendments on August 1, 2023. The official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

These amendments change require ments 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 these 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 these amendments involve no significant hazards consideration, as published in the Federal Register dated December 6, 2022 (87 FR 74665), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

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

6.0 CONCLUSION

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

Principal Contributors: D. Woodyatt R. Grover S. Goetz F. Forsaty

OFFICIAL USE ONLY PROPRIETARY INFORMATION

Package: ML23198A358 Proprietary: ML23198A361 Nonproprietary: ML23198A359 NRR-058 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DSS/SNSB NAME SGoetz KZeleznock PSahd DATE 07/17/2023 07/25/2023 / 09/08/2023 05/26/2023 OFFICE NRR/DSS/STSB/BC OGC - NLO NRR/DSS NAME VCusumano MWoods HGonzález DATE 05/26/2023 09/28/2023 10/02/2023 OFFICE NRR/DORL/LPL1/PM NAME SGoetz DATE 10/02/2023