ML21034A169

Issuance of Amendment Nos. 143 and 50 Regarding Implementation of Full Spectrumtm Loss-of-Coolant Accident Analysis (LOCA) and New LOCA-Specific Tritium Producing Burnable Absorber Rod Stress Analysis Methodology

ML21034A169
Person / Time
Site:	Watts Bar
Issue date:	02/26/2021
From:	Kimberly Green Plant Licensing Branch II
To:	Jim Barstow Tennessee Valley Authority
Green K
References
EPID L-2020-LLA-0005
Download: ML21034A169 (43)

Text

OFFICIAL USE ONLY PROPRIETARY INFORMATION February 26, 2021 Mr. James Barstow Vice President, Nuclear Regulatory Affairs and Support Services Tennessee Valley Authority 1101 Market Street, LP 4A-C Chattanooga, TN 37402-2801

SUBJECT:

WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 143 AND 50 REGARDING IMPLEMENTATION OF FULL SPECTRUM' LOSS-OF-COOLANT ACCIDENT ANALYSIS (LOCA) AND NEW LOCA-SPECIFIC TRITIUM PRODUCING BURNABLE ABSORBER ROD STRESS ANALYSIS METHODOLOGY (EPID L-2020-LLA-0005)

Dear Mr. Barstow:

The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 143 to Facility Operating License No. NPF-90 and Amendment No. 50 to Facility Operating License No. NPF-96 for the Watts Bar Nuclear Plant, Units 1 and 2, respectively. These amendments are in response to your application dated January 17, 2020, which was corrected by letter dated January 26, 2021, and supplemented by letter dated August 27, 2020.

The amendments revise WBN Units 1 and 2, Technical Specification (TS) 5.9.5, Core Operating Limits Report, to replace the loss-of-coolant accident (LOCA) analysis evaluation model references with reference to the FULL SPECTRUM' Loss-of-Coolant Accident (FSLOCA') Evaluation Model analysis. The amendments also revise the WBN Unit 2, Operating License (OL) condition 2.C(4) to reflect the implementation of the FSLOCA Evaluation Model methodology, and the WBN Unit 1 TS 4.2.1, Fuel Assemblies, to delete discussion of Zircalloy fuel rods. Additionally, the amendments approve the use the new LOCA-specific tritium producing burnable absorber rod (TPBAR) stress analysis methodology to evaluate the integrity of the TPBARs for the conditions expected during a large break LOCA and provide a recovery of margin in the post-LOCA criticality evaluation in the presence of assumed TPBAR failures.

Enclosure 4 to this letter contains proprietary information. When separated from Enclosure 4, this document is decontrolled.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION J. Barstow A copy of our related safety evaluation is also enclosed. Notice of issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Kimberly J. Green, Senior Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-390 and 50-391

Enclosures:

1. Amendment No. 143 to NPF-90

2. Amendment No. 50 to NPF-96

3. Safety Evaluation (Non-Proprietary)

4. Safety Evaluation (Proprietary) cc: Listserv OFFICIAL USE ONLY PROPRIETARY INFORMATION

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-390 WATTS BAR NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 143 License No. NPF-90

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

A. The application for amendment by Tennessee Valley Authority (TVA, the licensee) dated January 17, 2020, corrected by letter dated January 26, 2021, and supplemented by letter dated August 27, 2020, 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.

Enclosure 1

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 Facility Operating License No. NPF-90 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. 143 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of the date of its issuance, and shall be implemented before power operations with a core design that credits TPBAR integrity in the post-LOCA criticality analysis, not to exceed December 31, 2021.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Undine S. Shoop, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License and Technical Specifications Date of Issuance: February 26, 2021

ATTACHMENT TO AMENDMENT NO. 143 WATTS BAR NUCLEAR PLANT, UNIT 1 FACILITY OPERATING LICENSE NO. NPF-90 DOCKET NO. 50-390 Replace page 3 of Facility Operating License No. NPF-90 with the attached revised page 3.

Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

Remove Pages Insert Pages 4.0-1 4.0-1 5.0-29 5.0-29

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-391 WATTS BAR NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 50 License No. NPF-96

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

A. The application for amendment by Tennessee Valley Authority (TVA, the licensee) dated January 17, 2020, corrected by letter dated January 26, 2021, and supplemented by letter dated August 27, 2020, 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.

Enclosure 2

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 Facility Operating License No. NPF-96 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. 50 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of the date of its issuance, and shall be implemented prior to startup from the outage where the Watts Bar, Unit 2 steam generators (SGs) are replaced with SGs equivalent to those in Watts Bar, Unit 1, not to exceed January 15, 2024.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Undine S. Shoop, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License and Technical Specifications Date of Issuance: February 26, 2021

ATTACHMENT TO AMENDMENT NO. 50 WATTS BAR NUCLEAR PLANT, UNIT 2 FACILITY OPERATING LICENSE NO. NPF-96 DOCKET NO. 50-391 Replace page 3 of Facility Operating License No. NPF-96 with the attached revised page 3.

Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

Remove Pages Insert Pages 5.0-31 5.0-31

C. The 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 TVA is authorized to operate the facility at 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. 50 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) TVA shall implement permanent modifications to prevent overtopping of the embankments of the Fort Loudon Dam due to the Probable Maximum Flood by June 30, 2018.

(4) PAD4TCD may be used to establish core operating limits until the WBN Unit 2 steam generators are replaced with steam generators equivalent to the existing steam generators at WBN Unit 1. FULL SPECTRUM LOCA Methodology shall be implemented when the WBN Unit 2 steam generators are replaced with steam generators equivalent to the existing steam generators at WBN Unit 1.

(5) By December 31, 2019, the licensee shall report to the NRC that the actions to resolve the issues identified in Bulletin 2012-01, Design Vulnerability in Electrical Power System, have been implemented.

(6) The licensee shall maintain in effect the provisions of the physical security plan, security personnel training and qualification plan, and safeguards contingency plan, and all amendments made pursuant to the authority of 10 CFR 50.90 and 50.54(p).

(7) TVA shall fully implement and maintain in effect all provisions of the Commission approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The TVA approved CSP was discussed in NUREG-0847, Supplement 28, as amended by changes approved in License Amendment No. 7.

(8) TVA shall implement and maintain in effect all provisions of the approved fire protection program as described in the Fire Protection Report for the facility, as described in NUREG-0847, Supplement 29, subject to the following provision:

Unit 2 Facility Operating License No. NPF-96 Amendment No. 50

Reporting Requirements 5.9 5.9 Reporting Requirements 5.9.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

1. WCAP-9272-P-A, WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY, July 1985 (W Proprietary). (Methodology for Specifications 3.1.4 - Moderator Temperature Coefficient, 3.1.6 -

Shutdown Bank Insertion Limit, 3.1.7 - Control Bank Insertion Limits, 3.2.1 - Heat Flux Hot Channel Factor, 3.2.2 - Nuclear Enthalpy Rise Hot Channel Factor, 3.2.3 - Axial Flux Difference, and 3.9.1 - Boron Concentration).

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

3. WCAP-10216-P-A, Revision 1A, RELAXATION OF CONSTANT AXIAL OFFSET CONTROL F(Q) SURVEILLANCE TECHNICAL SPECIFICATION, February 1994 (W Proprietary). (Methodology for Specification 3.2.3 - Axial Flux Difference (Relaxed Axial Offset Control).)

4. WCAP-12610-P-A, VANTAGE + FUEL ASSEMBLY REFERENCE CORE REPORT, April 1995. (W Proprietary). (Methodology for Specification 3.2.1 - Heat Flux Hot Channel Factor).

(continued)

Watts Bar - Unit 2 5.0-31 Amendment 49, 50

ENCLOSURE 3 NON-PROPRIETARY SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 143 TO FACILITY OPERATING LICENSE NO. NPF-90 AND AMENDMENT NO. 50 TO FACILITY OPERATING LICENSE NO. NPF-96 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-390 AND 50-391 Proprietary information pursuant to Section 2.390 of Title 10 of the Code of Federal Regulations has been redacted from this document.

Redacted information is identified by blank space enclosed within (( double brackets )).

OFFICIAL USE ONLY PROPRIETARY INFORMATION SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 143 AND 50 TO FACILITY OPERATING LICENSE NOS. NPF-90 AND NPF-96 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-390 AND 50-391

1.0 INTRODUCTION

By letter dated January 17, 2020 (Reference 1), which was corrected by letter dated January 26, 2021 (Reference 3), as supplemented by letter dated August 27, 2020 (Reference 2), the Tennessee Valley Authority (TVA, the licensee), submitted a license amendment request (LAR) to revise the Watts Bar Nuclear Plant (Watts Bar), Units 1 and 2, Technical Specifications (TSs).

The requested changes would revise Watts Bar, Units 1 and 2, TS 5.9.5, Core Operating Limits Report, to replace the loss-of-coolant accident (LOCA) analysis evaluation model references with reference to the FULL SPECTRUM' Loss-of-Coolant Accident (FSLOCA') Evaluation Model analysis. The LAR also requested revision to the Watts Bar, Unit 2, Operating License (OL) condition 2.C(4) to reflect the implementation of the FSLOCA Evaluation Model methodology, and the Watts Bar, Unit 1, TS 4.2.1, Fuel Assemblies, to delete discussion of Zircalloy fuel rods. Additionally, the LAR requested approval of the use the new LOCA-specific tritium producing burnable absorber rod (TPBAR) stress analysis methodology to evaluate the integrity of the TPBARs for the conditions expected during a large break LOCA and provide a recovery of margin in the post-LOCA criticality evaluation in the presence of assumed TPBAR failures.

The supplement dated August 27, 2020, and correction dated January 26, 2021, provided additional information that clarified and corrected the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register on June 2, 2020 (85 FR 33745).

The LAR references several analyses or documents that were not provided as part of the application or as listed in the references section. In order to confirm that the analyses and references support the requested licensing action, the NRC staff performed a virtual audit of several documents. The NRC staffs audit report summarizes the staffs audit activities (Reference 4). Based on the review of audit documents, the staff issued a request for additional OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION information (Reference 5). By letter dated August 27, 2020, the licensee responded to the requests for additional information (RAIs) (Reference 2).

This safety evaluation contains proprietary information, which is marked with double brackets and bold font such as (( )).

2.0 REGULATORY EVALUATION

2.1 Requested Changes The licensee proposed the following changes to operating licenses and TSs for Watts Bar, Units 1 and 2 to reflect the use of the WCAP-16996-P-A, Revision 1 methodology:

Watts Bar, Unit 1, TS 4.2.1, Fuel Assemblies, currently states:

The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircalloy, ZIRLO, or Optimized ZIRLO' clad fuel rods Watts Bar, Unit 1, TS 4.2.1, Fuel Assemblies, is requested to change to:

The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of ZIRLO or Optimized ZIRLO' clad fuel rods Watts Bar, Units 1 and 2, TS 5.9.5, Core Operating Limits Report (COLR), currently states, in part:

2a. WCAP-12945-P-A, Volume I (Revision 2) and Volumes 2 through 5 (Revision 1), Code Qualification Document for Best-Estimate Loss of Coolant Analysis, March 1998 (W Proprietary). (Methodology for Specification 3.2.1 - Heat Flux Hot Channel Factor, and 3.2.2 - Nuclear Enthalpy Rise Hot Channel Factor).

b. WCAP-10054-P-A, Small Break ECCS Evaluation Model Using NOTRUMP Code, August 1985. Addendum 2, Rev. 1: Addendum to the Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model, July 1997. (W Proprietary). (Methodology for Specifications 3.2.1 - Heat Flux Hot Channel Factor, and 3.2.2 - Nuclear Enthalpy Rise Hot Channel Factor).

Watts Bar, Units 1 and 2, TS 5.9.5, Core Operating Limits Report (COLR), is requested to state, in part:

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

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Watts Bar, Unit 2, Operating License Condition 2.C(4) currently states:

PAD4TCD may be used to establish core operating limits until the WBN Unit 2 steam generators are replaced with steam generators equivalent to the existing steam generators at WBN Unit 1.

Watts Bar, Unit 2, Operating License Condition 2.C(4) is requested to change to:

PAD4TCD may be used to establish core operating limits until the WBN Unit 2 steam generators are replaced with steam generators equivalent to the existing steam generators at WBN Unit 1. FULL SPECTRUM LOCA Methodology shall be implemented when the WBN Unit 2 steam generators are replaced with steam generators equivalent to the existing steam generators at WBN Unit 1.

2.2 Regulations and Guidance The NRC staff considered the following regulatory requirements during its review of licensees LAR.

Section 50.36, Technical specifications, of Title 10 of the Code of Federal Regulations (10 CFR) establishes the regulatory requirements related to the content of TSs.

Paragraph 50.36(a)(1) requires an application for an operating license to include proposed TSs.

A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the TSs.

Pursuant to 10 CFR 50.36, TSs for operating reactors are required, in part, to include items in the following five specific categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) Surveillance Requirements; (4) design features; and (5) administrative controls. The regulation at 10 CFR 50.36(c)(5) states that 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.

Paragraph (a)(1)(i) of 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, requires, in part, the use of an acceptable emergency core cooling system (ECCS) evaluation model (EM), and for that model to make comparisons to applicable experimental data, account for uncertainty, and show with a high level of probability that the acceptance criteria in paragraph (b) are not exceeded.

The following paragraphs of 10 CFR 50.46(b) require that:

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

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

(3) Maximum hydrogen generation. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

(4) Coolable geometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling.

(5) Long-term cooling. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

Appendix A, General Design Criteria [GDC] for Nuclear Power Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, GDC 35, Emergency core cooling, requires that a system to provide abundant emergency core cooling shall be provided.

The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

The NRC staff considered the following guidance documents to provide additional guidance on acceptable approaches to demonstrate that the above regulatory requirements are met.

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, establishes the review and acceptance criteria for NRC staff to use when evaluating a licensees emergency core cooling system (ECCS) performance analysis for compliance with 10 CFR 50.46 and relevant GDC (Reference 6).

NRC Regulatory Guide (RG) 1.157, Best-Estimate Calculations of Emergency Core Cooling System Performance, describes models, correlations, data, model evaluation procedures, and methods that are acceptable to the NRC staff for meeting the requirements for a realistic or best-estimate calculation of ECCS performance during a LOCA and for estimating the uncertainty in that calculation (Reference 7).

NRC Regulatory Guide 1.203, Transient and Accident Analysis Methods, describes a process that the NRC staff considers acceptable for use in developing and assessing evaluation models that may be used to analyze transient and accident behavior that is within the design basis of a nuclear power plant (Reference 8).

Generic Letter 88-16, Removal of Cycle-Specific, Parameter Limits from Technical Specifications, provides that it is acceptable for licensees to control reactor physics parameter limits by specifying an NRC-approved calculation methodology (Reference 9).

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION

3.0 TECHNICAL EVALUATION

The NRC staff evaluated the licensee's LAR to determine whether the proposed changes would continue to meet the regulations and guidance provided in Section 2.2 of this safety evaluation.

The NRC staff reviewed the licensees proposed changes to verify that all limitations and conditions in applicable NRC-approved methods are met, the licensee appropriately applied the LOCA EM to Watts Bar, Units 1 and 2, and the acceptance criteria of 10 CFR 50.46(b)(1) through (4) are satisfied.

3.1 Background

TVA requested approval to use the FSLOCA EM to evaluate the peak cladding temperatures for large-break and small-break LOCAs. In addition, TVA requested approval to use the new TPBAR stress analysis methodology to provide a recovery of margin in the post-LOCA criticality evaluation in the presence of assumed TPBAR failures. TVA proposes to use the new LOCA-specific TPBAR stress analysis methodology to evaluate the integrity of the TPBARs for the conditions expected during a large-break LOCA (LBLOCA).

The presence of TPBARs in Watts Bar, Units 1 and 2 results in positive reactivity insertion following a LOCA in the event of cladding rupture at high temperatures. During a post-LOCA core uncovery, the overheating of fuel rods causes heating of the TPBARs located in adjacent control rod guide tubes. The heating of the TPBARs can result in rupture of the TPBAR cladding due to the increase in internal pressure. As a result of potential TPBAR cladding rupture, Li-6 (Lithium-6) will leak out near the rupture location and the potential for subsequent leaching of Li-6 in the long term. Lithium-6 is a neutron absorbing isotope and the loss of Li-6 results in positive reactivity addition. These TPBAR structural integrity analyses were used to demonstrate that the TPBARs remain intact following a LBLOCA. This demonstration is needed, because if a TPBAR loses its structural integrity, some of the lithium contained within can leach out, causing an increase in the reactivity in the core, and potentially adding heat.

These analyses are necessary to confirm that the amount of energy analyzed during the transient, using FSLOCA, is valid and realistic.

The LOCA-specific TPBAR stress analysis methodology relies on conditions resulting from LBLOCA simulations generated according to the FSLOCA evaluation model. Application of new TPBAR LOCA-specific TPBAR stress analysis methodology requires survival ((

)) survival does not need to be demonstrated.

The following sections describe the LOCA-specific TPBAR stress analysis methodology, TPBAR cladding stress evaluation following an LBLOCA, TPBAR cladding acceptance criteria and methodology conservatisms, and stress analysis results.

3.2 LOCA-Specific TPBAR Stress Analysis Methodology The LOCA-specific TPBAR cladding stress analysis methodology is designed to determine the potential for TPBAR cladding mechanical rupture under LBLOCA temperature and differential pressure conditions. The LBLOCA for Watts Bar, Units 1 and 2 is analyzed using the WCOBRA/TRAC-TF2 thermal hydraulic code using the FSLOCA methodology to predict the response of fuel rods for cladding temperatures and oxidation following a postulated LOCA.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The TPBAR cladding temperatures are used in the determination of whether the TPBAR is expected to rupture following the postulated LOCA.

3.2.1 Modeling of WBN vessel in WCOBRA/TRAC-TF2 Figure 1 of TVAs August 27, 2020, letter (Reference 2) provides vertical view of the Watts Bar, Units 1 and 2, vessel (3D) noding. Positive flow is in the direction indicated by the arrow.

WCOBRA/TRAC-TF2 assumes the existence of a vertical flow path between vertically stacked channels, unless specified otherwise by input. ((

))

Core Model Both heated and unheated conductors are modeled in WCOBRA/TRAC-TF2. Unheated conductors are lower core plate, reactor vessel wall, core barrel, upper core plate guide tubes, and, and support columns and heated conductors are fuel rod, other conductors with detailed axial and radial noding. ((

))

Loop Model Figures 3 through 5 of TVAs August 27, 2020, letter show noding diagrams for loop layout and the steam generators and other major components in the primary system outside the reactor vessel.

Emergency Core Cooling and Safety Injection Model The emergency core cooling system (ECCS) for Watts Bar, Units 1 and 2 consists of four accumulator tanks, two centrifugal charging pumps, two high head safety injection (HHSI) pumps, and two low head safety injection (LHSI) pumps with each pump connected to injection lines, which are headered with the accumulator injection lines, with the exception of the charging pumps, which are connected directly to the cold legs. The loss of one train of safety injection is considered as the limiting single failure assumption.

Overview of Methodology Stress field on the TPBAR is evaluated using a methodology derived using guidance from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC)

(Reference 11), including calculations for the primary membrane stresses (i.e., internal pressure) and the primary bending stresses. Two non-limiting stress components such as cladding stresses from thermal shock during reflood quench stage of LBLOCA, and the stress arising from the end plug to cladding tube weld joint geometric discontinuity along with internal pressure were assessed using ANSYS analysis. The results of the ANSYS analysis show that the analytical model used in the TPBAR design process produces a highly conservative estimate of the stress concentration (by a factor of ~2) at the weld joint. However, the licensee OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION has shown that both thermal stresses from reflood quench and the stress concentration at the weld joints have been determined to be non-limiting.

Survival of TPBARs during blowdown phase of LBLOCA During the early few seconds of blowdown following a LBLOCA, the fuel rod heating is rapid and is primarily due to the transfer of fuel pellet stored energy to the cladding after heat transfer to the coolant is interrupted. ((

)) Reference 12, NDP-98-181, Revision 1, Tritium Production Core (TPC) Topical Report, which was approved by NRC staff, examined the response of the TPBAR to an LBLOCA as:

The TPBAR generates minimal heat during a LOCA and is heated primarily by radiation from the fuel rods to the fuel assembly guide thimble and radiation from the thimble across the gap to the TPBAR. Convection of the steam and entrained liquid, on the outer thimble surface, provides cooling comparable to that experienced by the fuel rods.

The LOCTA_JR code was used to predict the thermal behavior of TPBARs during a LOCA (References 12, 13, and 14). TPBARs are modeled with the thimble tube heatup model which solves the one-dimensional transient conduction and radiation equations. LOCTA_JR uses the cladding temperature of the surrounding fuel rods and the core steam and entrained liquid convective heat transfer coefficients and temperatures as boundary conditions for the calculation of the TPBAR response within the fuel assembly guide thimble. The LOCTA_JR LOCA analysis shows the fuel assembly thimble temperature leads the TPBAR cladding temperature with respect to time (Figure 6 of Reference 2). During the heating phase the thimble temperature heats sooner and faster than the TPBAR and during the cooling period the thimble temperature cools sooner and faster.

((

))

((

))

((

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))

The licensee performed TPBAR structural analysis in which it conservatively assumed that the

((

))

In view of the above, the NRC staff has determined that the structural integrity of TPBAR cladding is maintained during an LBLOCA.

3.2.2 TPBAR Cladding Stress Analysis Acceptance Criteria The TPBAR cladding acceptance criteria for stress analysis consists of two modes: rapid burst due to over-pressurization and thermal creep rupture, which is a time-dependent material deformation mechanism that leads to ductility exhaustion. The acceptance criterion for the rapid burst failure mode was developed using data from unirradiated cold-worked SS-316 available in the open literature and from tests performed by PNNL on unirradiated TPBAR coated cladding (References 3 and 4). Mechanical testing on material from the Fast Flux Test Facility program found that irradiation of cold-worked SS-316 similar to TPBAR cladding material leads to an increase in both the yield stress and the ultimate tensile strength at pressurized water reactor (PWR) operating temperature conditions (References 2 and 4).

Burst Experiments and Creep Rupture Experiments In the burst tests, short cladding specimens were fitted with one solid end plug and one open end plug connected to a gas pressurization system; full-length specimens were pre-pressurized and sealed prior to testing. Short specimens were heated to burst and full-length specimens were heated to a constant temperature and pressure until burst. The temperatures and pressures in the tests were based on LOCA analyses performed for nuclear reactor systems expected to host TPBARs. PNNL conducted five sets of burst tests. The details of the tests are summarized in TTP-3-721, Revision 1, High Temperature Fracture Models for Assessment of TPBAR Cladding Survivability During LOCA, one of the documents audited by the NRC staff (Reference 4), and briefly described in the response to SFNB RAI 2(a) (Reference 2). The first PNNL test (Test #1a) was conducted on a preliminary TPBAR design for proof of concept and used short specimens constructed from 20-percent cold-worked 316 stainless steel (SS) cladding charged with helium gas at various pressures and inductively heated to the point of bursting. Test #1b used 4-foot long specimens and was conducted at four different backfill pressures with helium from 1000 pounds per square inch absolute (psia) to 3000 psia and heated till burst. Tests #2 and #3 involved full-length specimens of 20-percent cold-worked stainless steel 316 cladding containing prototype hardware (getters, pellets, liners) in order to test the response of the prototype TPBARs under LOCA temperatures. Tests #4 and #5 were performed at PNNL. The test articles used short lengths (approximately seven inches) of OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION coated cladding with a hollow stainless steel filler rod to limit the gas volume. Each burst test was found acceptable if the test specimen temperature was increased until the specimen burst.

PNNL conducted 98 tests used for the development of a creep rupture model. Each specimen was pressurized at room temperature in atmospheric conditions to a pre-determined level, then the temperature was ramped (at a rate of 20°C/s) to the test temperature. After the temperature was reached, the pressure and temperature were held constant until rupture. The specimen was subjected to steady creep rate followed by increasing strain rates, tertiary creep or flow.

For each test, the pressure was converted to stress, and with the temperature and time to failure, the results were evaluated using the Larson-Miller parameter and Life Fraction Rule (Reference 15) where the life fraction (LF) was set to unity corresponding to the failure time, and each test gives a unique C-value or C-parameter. The C-value is a material constant to be determined from experimental data. Evaluation of C-parameter is detailed in one of the audit documents (Document Number 4 of Reference 4). The value of C-parameter was from the test data and was adjusted for each test so that the calculated is unity at the time of failure. This resulted in C-parameter values that ranged from a minimum to maximum value from which an average value resulted. A best estimate of C-parameter with a standard deviation was developed for which a range of failure times for upper bound was obtained from tests performed.

The burst rupture and creep rupture curves are the allowable stress limits for TPBAR behavior during a LOCA event. They represent the stress limits for burst rupture and thermal creep rupture mechanisms. The cladding stresses are calculated using classic elasticity theory and the development of stress intensities derived from ASME BPVC methods. These stress intensities are then compared to actual burst and creep burst curves at TPBAR temperature and axial location to generate the smallest factor of safety. In the case of TPBAR cladding burst failure, the approach compares the calculated stress intensities against the allowable burst stress limit to determine the minimum factor of safety (FS). The FS for TPBAR cladding burst during LOCA conditions is calculated from the ratio lower bound burst allowable stress to stress intensity. The structural integrity of the TPBAR is evaluated by comparing the factor of safety to a value of unity. Failure is expected for FS less than 1.0.

From the curves that represent discrete times to failure at times ranging from 60 to 1200 seconds the life fraction rule is applied to compute the accumulation of damage caused by creep mechanisms for conditions when the stress intensity exceeds the creep rupture curves. The accumulated creep damage (D) is then used to assess the potential for failure by the creep rupture mechanism using the following acceptance criteria.

D < 1.0 creep rupture not expected D 1.0 creep rupture expected The creep damage allowable limit is applied only to the primary membrane plus bending stress intensities as these stress components are the extended for long time applied stresses during the event. The most conservative stress intensities are then compared to actual experimental data corresponding to TPBAR temperature and axial location to generate the smallest factor of safety in the TPBAR.

The NRC staff reviewed all the documents including the RAI response (Reference 2) that describe the tests and experiments for burst rupture and creep rupture and the resulting curves OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION that are indications of allowable stress limits for TPBAR cladding behavior during the LOCA. As a result, the staff has determined that the safety factor and acceptance criteria derived from the test data are acceptable for determining the TPBAR cladding integrity during a LOCA event.

3.2.3 TPBAR Stress Analysis Methodology Conservatisms Table 4.2.2-1 of TVAs August 27, 2020, letter (Reference 2) provides conservative elements and assumptions used in LOCA-specific TPBAR stress analysis methodology. Conservatism has been built into rod temperatures for TPBAR temperatures and internal pressure, TPBAR internal void volume, cladding tolerance stack up, cladding corrosion allowance, Tritium released from the pellets available for gas pressurization, and burst criterion. The NRC staff has reviewed the details of TPBAR stress analysis and has determined that there is sufficient conservatism built into the parameters used in the calculations.

3.2.4 Uncertainty Analysis Uncertainty analysis for TPBAR structural integrity analysis is similar to the approach implemented in the FSLOCA EM for LBLOCA analysis. The statistical analysis is performed to construct tolerance limits for the figures of merit related to the TPBAR structural integrity. The statistical analysis is performed to construct tolerance limits for the figures of merit related to the TPBAR structural integrity: rupture due to primary membrane and bending stresses, and rupture due to creep damage. Typical LOCA analysis consists of WCT-TF2 simulations ((

))

In the TPBAR structural integrity analysis, a sample of postulated LOCAs is simulated in WCT-TF2 to determine the resulting fuel rod cladding temperatures. The TPBAR cladding temperature is then used to calculate margins to failure thresholds. Section 30 of WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology) (Reference 16), describes the statistical process which is applied to TPBAR figures of merit (FOMs) and then compares the tolerance limits to the analysis acceptance criteria (failure thresholds). ((

)) (Reference 2). The licensees results of the statistical analysis demonstrate a lower tolerance limit for primary membrane and bending stress safety factor and an upper tolerance limit for creep damage ratio which bound 95 percent of postulated LOCAs with 95 percent confidence (i.e., 95/95).

The NRC staff has evaluated the uncertainty analysis for TPBAR structural integrity analysis and has determined that the licensee has implemented an appropriate uncertainty method.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.2.5 Westinghouse Post-LOCA Criticality According to the Westinghouse methodology (Reference 3), post-LOCA conditions in the reactor core are evaluated to address the potential to challenge the margin to criticality. The challenge to margin to criticality arises from (1) initial boron concentration that remains in the primary system following LOCA, (2) decrease in sump boron concentration as the sump water becomes the ECCS injection source, (3) alignment of hot leg circulation that prevents the occurrence of post-LOCA criticality, and (4) the loss of xenon negative reactivity. Conservative assumptions and inputs to post-LOCA criticality methodology are: minimum refueling water boron concentration, minimum cold leg accumulator boron concentration, minimum RCS boron concentration, minimum containment sump boron concentration, time of hot leg alignment, xenon activity for the hot leg circulation case, zero xenon reactivity for long term, cold conditions, and most reactive time in life.

The new post-LOCA criticality methodology assumes no TPBAR failure during the time-in-cycle that the failure contributes to criticality evaluation. Positive reactivity additions associated with TBPAR failure have been dispositioned. For each cycle of operation, the licensee will show that the reload methodology will confirm post-LOCA subcriticality is maintained.

The NRC staff reviewed the four scenarios that were considered and the conservative assumptions and inputs to the post-LOCA criticality analysis and determined that the core including the TPBARs will be maintained subcritical after an LBLOCA.

3.2.6 TPBAR Structural Integrity Analysis Results The LOCA-specific TPBAR structural analysis results are summarized in the LAR (Reference 3). The TPBAR integrity analysis shows retention of significant margin to the acceptance criteria for both offsite power available (OPA) and loss of offsite power (LOOP).

The licensee determined that creep damage is negligible for both OPA and LOOP conditions.

Increased tritium mass is a dominant contributor to more limiting results, as it produces a higher TPBAR internal pressure, and thus more limiting stress conditions. Primary membrane and bending stress safety factor results have been plotted ((

)). Application of the methodology requires TPBAR survival

(( )) TPBAR failure is of no consequence and its survival need not be demonstrated as long as post-LOCA criticality margins remains acceptable in the presence of ruptured TPBARs.

Figure 4.3.2-5 in the LAR shows that the calculated burst stress required to rupture TPBAR at LOCA temperature, the stress intensity and the ratio of the two as the calculated safety factor at the same node yielding the lowest safety factor. ((

)) The effect of the reduction in TPBAR average temperature, a surrogate for internal pressure and starting the initial few seconds after the break, results in a general reduction in stress intensity over the same period. Conversely, the burst stress follows the opposite trend with respect to TPBAR peak cladding temperature, i.e., a reduction in stress required to cause rupture.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION In summary, the licensee stated that the 95/95 tolerance limits for (1) the primary membrane and bending stress safety factor, and (2) the creep damage ratio maintain significant margin to the burst stress failure criteria, and as such, there is very high confidence that the TPBARs will not rupture following a postulated LBLOCA.

Sensitivity studies for PWR parameters derived from a subset uncertainty analysis covering various designs and fuel were examined to determine the sensitivity of the analysis results to the error correction. The error correction was found to be different for different transient phases, blowdown versus reflood. Based on the results from the PWR sensitivity studies, the correction of the error is estimated to result in a fuel cladding temperature increase for the time period relevant to TPBAR structural integrity, which is assumed to also lead to a TPBAR cladding temperature increase. TPBAR structural integrity calculations were performed with an assumed increase in TPBAR temperature throughout the transient. The updated TPBAR structural integrity analysis results in acceptable 95/95 primary membrane and bending stress safety factor and 95/95 cumulative creep damage ratio as indicated in Table 4.3.2-1 of the LAR.

The NRC staff reviewed the results of TPBAR structural integrity analysis, including the sensitivity studies, and has determined that TPBARs will not rupture with 95/95 probability and confidence.

3.2.7 Summary and Conclusion for the LOCA-Specific TPBAR Stress Analysis Methodology The licensee has performed a TPBAR structural integrity analysis following a LOCA. The LOCA-specific TPBAR stress analysis methodology relies on conditions resulting from LBLOCA simulations according to the FSLOCA methodology. The NRC staff reviewed all the aspects of the TPBAR stress analysis for structural integrity such as the Watts Bar units core model, cladding stress analysis acceptance criteria, conservatism built into the integrity analysis, post-LOCA criticality analysis, and uncertainty analysis. The NRC staff also has determined that the licensees assumption that the TPBARs remain intact post-LOCA, which is relied upon in the FSLOCA analysis, is acceptable.

3.3 Limitations and Conditions The safety evaluation for WCAP-16996-P-A, Revision 1 (Reference 17) contains 15 limitations and conditions that must be met in order for a licensee to be permitted to implement the NRC-approved FSLOCA EM.

A summary of each limitation and condition and how it has been met as stated by the licensee in its corrected application dated January 26, 2021, and the associated NRC staff findings are provided below.

3.3.1 Limitation and Condition Number 1 - Applicability with Regard to LOCA Transient Phases Condition The FSLOCA EM is not approved to demonstrate compliance with 10 CFR 50.46 acceptance criterion (b)(5) related to the long-term cooling.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION Licensees Compliance The analysis for Watts Bar, Units 1 and 2 with the FSLOCA EM is only being used to demonstrate compliance with 10 CFR 50.46 (b)(1) through (b)(4).

Staff Evaluation 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.

3.3.2 Limitation and Condition Number 2 - Applicability with Regard to Type of PWR Plant Condition 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.

Licensees Compliance Watts Bar, Units 1 and 2 are Westinghouse-designed 4-loop PWRs with cold-side injection, so they are within the NRC-approved methodology. The analysis for Watts Bar, Units 1 and 2 utilizes the NRC-approved FSLOCA methodology, except for the changes which were previously transmitted to the NRC pursuant to 10 CFR 50.46 in LTR-NRC-18-30 (Reference 18).

After the licensee completed the analysis, two errors were discovered in the WCOBRA/TRAC-TF2 code which were found to have negligible impact on the results of the FSLOCA EM. The errors were transmitted to the NRC in LTR-NRC-18-30.

The treatment for the uncertainty in the gamma energy redistribution and the equation for the assumed increase in hot rod and hot assembly relative power are described in WCAP-16996-P-A. The power increase in the hot rod and hot assembly due to energy redistribution in the application of the FSLOCA EM to Watts Bar, Units 1 and 2 was calculated incorrectly. This error resulted in a 0 to 5 percent deficiency in 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 found that the error correction has only a limited impact on the power modeled for a single assembly in the core. The error correction results in negligible impact on the system thermal-hydraulic response during the postulated LOCA.

For Region I, the primary impact of the error correction is on the rate of cladding heatup above the two-phase mixture level in the core during the boiloff phase. The licensee evaluated the peak clad temperature (PCT) impact using run-specific PCT versus linear heat rate relationships and the run-specific hot rod and hot assembly linear heat rate increase resulting from the error correction. The results led to a final PCT of 978 °F for the Region I analysis.

For Region II, the licensee examined parametric PWR sensitivity studies, derived from a subset of uncertainty analysis simulations covering various design features and fuel arrays, to OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION determine the sensitivity of the analysis results to the error correction. The licensee found the PCT impact from the error correction 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 correction of the error is estimated to increase the Region II analysis PCT by 20 °F, leading to an analysis result of 1477 °F for the Region II analysis assuming loss-of-offsite power and 1464 °F for the Region II analysis assuming offsite power available.

The analysis results including the error correction continue to maintain compliance with the 10 CFR 50.46 acceptance criteria.

Staff Evaluation Given that Watts Bar, Units 1 and 2 are Westinghouse-designed 4-loop PWRs with cold-side injection, the NRC staff finds that the FSLOCA EM is applicable to them. In addition, the staff finds that the licensee has appropriately applied the FSLOCA EM with the changes described in this safety evaluation. Therefore, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 2.

3.3.3 Limitation and Condition Number 3 - Applicability for Containment Pressure Modeling Summary For Region II, the containment pressure calculation will be executed in a manner consistent with the approved methodology (i.e., the COCO or LOTIC2 model will be based on appropriate plant-specific design parameters and conditions, and engineered safety features which can reduce pressure are modeled). This includes utilizing a plant-specific initial containment temperature, and only taking credit for containment coatings which are qualified and outside of the break zone-of-influence.

Licensees Compliance The containment pressure calculation for the Watts Bar, Units 1 and 2 analysis was performed consistent with the NRC-approved methodology. Appropriate design parameters and conditions were modeled, as were the engineered safety features which can reduce the containment pressure. A plant-specific initial temperature associated with normal full-power operating conditions was modeled, and no coatings were credited on any of the containment structures.

Staff Evaluation The NRC staff confirmed that the licensee used the NRC-approved methodology for the Region II containment pressure calculation. During the regulatory audit (Reference 4), the NRC staff reviewed calculation notes of the containment pressure calculation and confirmed the use of appropriate conditions. Therefore, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 3.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.3.4 Limitation and Condition Number 4 - Decay Heat Modeling Summary The decay heat uncertainty multiplier will be ((

)). The analysis simulations for the FSLOCA EM will not be executed for longer than 10,000 seconds following reactor trip unless 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 will be provided in the analysis submittal in units of sigma and absolute units.

Licensees Compliance Consistent with the NRC-approved methodology, the decay heat uncertainty multiplier will be

(( )) for Watts Bar, Units 1 and 2. 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 7 [of Attachment 1 to Enclosures 3 and 4 of the LAR].

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

3.3.5 Limitation and Condition Number 5 - Fuel Burnup Limits Summary The maximum assembly and rod length-average burnup is limited to ((

)), respectively.

Licensees Compliance The maximum analyzed assembly and rod length-average burnup is less than or equal to

(( )), respectively, for Watts Bar, Units 1 and 2.

Staff Evaluation Based on the above, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 5.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.3.6 Limitation and Condition Number 6 - WCOBRA/TRAC-TF2 Interface with PAD 5.0 Summary The fuel performance data for analyses with the FSLOCA EM should be based on the PAD5 code (at present), which includes the effect of thermal conductivity degradation. The nominal fuel pellet average temperatures and rod internal pressures should be the maximum values, and the generation of all the PAD5 fuel performance data should adhere to the NRC-approved PAD5 methodology.

Licensees Compliance PAD5 fuel performance data is utilized in the Watts Bar, Units 1 and 2 analysis with the FSLOCA EM. The analyzed fuel pellet average temperatures bound the maximum values calculated in accordance with Section 7.5.1 of Reference 21, and the analyzed rod internal pressures were calculated in accordance with Section 7.5.2 of Reference 21.

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

3.3.7 Limitation and Condition Number 7 - Interfacial Drag Uncertainty in Region I Analysis Summary The YDRAG uncertainty parameter should be ((

)).

Licensees Compliance Consistent with the NRC-approved methodology, the YDRAG uncertainty parameter was ((

)) for the Watts Bar, Units 1 and 2, Region I analysis.

Staff Evaluation The NRC staff finds that the licensee appropriately used the specified interfacial drag uncertainty parameter as noted above and, therefore, meets the requirement of Limitation and Condition 7.

1 Reference 2 of the licensees Enclosure 3: Westinghouse Performance Analysis and Design Model (PAD5), WCAP-17642-P-A, Revision 1, November 2017.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.3.8 Limitation and Condition Number 8 - Biased Uncertainty Contributors in Region I Analyses Summary The ((

)).

Licensees Compliance Consistent with the NRC-approved methodology, the ((

)) for the Watts Bar, Units 1 and 2, Region I analysis.

Staff Evaluation The NRC staff finds that the licensee appropriately used the specified biased uncertainty parameters as noted above and, therefore, meets the requirement of Limitation and Condition 8.

3.3.9 Limitation and Condition Number 9 - Effect on Bias in Applications for Region I Summary For PWR designs which are not Westinghouse 3-loop PWRs, a sensitivity study will be executed to confirm that the ((

)) for the plant design being analyzed. This sensitivity study should be executed once, and then referenced in all applications to that particular plant class.

Licensees Compliance Watts Bar, Units 1 and 2 are Westinghouse-designed 4-loop PWRs. The requested sensitivity study was performed for a 4-loop Westinghouse-designed PWR and is discussed in Reference 19.

Staff Evaluation The NRC staff reviewed the attachment to Reference 19. This document describes the sensitivity studies done on the selected parameters and demonstrates that ((

)).

Therefore, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 9.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.3.10 Limitation and Condition Number 10 - Boundary Between Region I and Region II Breaks Summary For PWR designs which are not Westinghouse 3-loop PWRs, a sensitivity study will be executed to: (1) demonstrate that no unexplained behavior occurs in the predicted safety criteria across the region boundary, and (2) ensure that the ((

)) must cover the equivalent 2 to 4-inch break range using RCS-volume scaling relative to the demonstration plant. This sensitivity study should be executed once, and then referenced in all applications to that particular plant class.

Additionally, the minimum sampled break area for the analysis of Region II should be 1 square foot (ft2).

Licensees Compliance Watts Bar, Units 1 and 2 are Westinghouse-designed 4-loop PWRs. The requested sensitivity study was performed for a 4-loop Westinghouse-designed PWR and is discussed in Reference 19.

The minimum sampled break area for the Watts Bar, Units 1 and 2, Region II analysis was 1 ft2.

Staff Evaluation The NRC staff reviewed the attachment to Reference 19. This document describes the sensitivities performed to demonstrate that the boundary between Region I and Region II breaks is appropriate for a 4-loop Westinghouse-designed plant.

The licensee completed the sensitivity in accordance with this guidance. In addition, the Region II analysis considers a minimum break area of 1.0 ft2 consistent with the requirement in the limitation and condition.

Therefore, the NRC staff finds that the requirements of Limitation and Condition 10 are met as the licensee used an acceptable sensitivity study described in Reference 19 to determine the appropriate break size range for Region I and boundary between Region I and Region II.

3.3.11 Limitation and Condition Number 11 -- (( )) in Uncertainty Analysis for Region II and Documentation of Reanalysis Results for Region I and Region II Summary There are various aspects of this Limitation and Condition, which are summarized below:

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

)) and the Region I and Region II analysis seeds OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION 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 documented, for the intended purpose of demonstrating compliance with the applicable acceptance criteria, then the changes and associated rationale for the changes will be provided in the analysis submittal. Additionally, the preliminary values for PCT, maximum local oxidation (MLO), and core-wide oxidation (CWO) 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 preliminary values is not required.

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

Licensees Compliance This Limitation and Condition was met for the Watts Bar, Units 1 and 2, analysis as follows:

1. The (( )) the Region I and Region II analysis seeds, and the 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 analyses seeds were not changed once they were 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 analyses are provided for Watts Bar, Units 1 and 2 in Table 1 [of Attachment 1 to Enclosures 3 and 4 of the LAR.].

Staff Evaluation During the regulatory audit (Reference 4), the NRC staff reviewed the documentation containing the analysis seeds and inputs. Given that the licensee has declared and documented the appropriate inputs and did not change these values once declared and documented, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 11.

3.3.12 Limitation and Condition Number 12 - Steam Generator Heat Removal during Small Break LOCAs Summary 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.

Licensees Compliance A bounding plant-specific dynamic pressure loss from the steam generator secondary-side to the main steam safety valves was modeled in the Watts Bar, Units 1 and 2, analysis.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION Staff Evaluation The NRC staff confirmed the use of a constant value to account for the dynamic pressure loss (Reference 2). The licensee stated that the initial opening pressure of the first stage main steam safety valve (MSSV) was modeled as 10 pounds per square inch (psi) higher than the plant-specific first stage MSSV set pressure, plus uncertainty. The licensee further state that at the beginning of the natural circulation period of the limiting transient, the flow through each MSSV is approximately 25 percent of rated flow and diminishes after that, a pressure loss of approximately 1.25 psi. The 10 psi additional modeled pressure is acceptable to account for the pressure loss. Given the use of a bounding dynamic pressure loss, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 12.

3.3.13 Limitation and Condition Number 13 - Upper Head Spray Nozzle Loss Coefficient Summary In plant-specific models for analysis with the FSLOCA EM: (1) the ((

)) and (2) the (( )).

Licensees Compliance The ((

)) in the analysis for Watts Bar, Units 1 and 2. The (( )) in the analysis.

Staff Evaluation Based on the above, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 13.

3.3.14 Limitation and Condition Number 14 - Correlation for Oxidation Summary 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.

Licensees Compliance For the Watts Bar, Units 1 and 2, analysis, the Baker-Just correlation was used to convert the LOCA transient time-at-temperature to an ECR. The resulting LOCA transient ECR was then summed with the pre-existing corrosion for comparison against the 10 CFR 50.46 local oxidation acceptance criterion of 17 percent.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION Staff Evaluation The NRC staff finds that by using the Baker-Just correlation, converting to an ECR, and accounting for pre-existing corrosion, the licensee has met the requirements of Limitation and Condition 14.

3.3.15 Limitation and Condition Number 15 - LOOP versus Offsite Power Available Treatment in Uncertainty Analysis for Region II 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 (( )).

Licensees Compliance The Region II uncertainty analysis for Watts Bar, Units 1 and 2 was 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 (see Section 5.0 [of to Enclosures 3 and 4 of the LAR]).

The ((

)).

Staff Evaluation Given that the licensee has performed the Region II analysis for both LOOP and OPA ((

)) and that the results from both analyses are in compliance with the acceptance criteria in 10 CFR 50.46(b)(1) through (b)(4), the NRC staff finds that the licensee has met the requirements of Limitation and Condition 15.

3.4 Results and Compliance with 10 CFR 50.46 As discussed in Section 3.3 of this safety evaluation, the licensee has acceptably implemented WCAP-16996-P-A, Revision 1, which is approved for use by the NRC staff and is an acceptable EM in accordance with the requirements of 10 CFR 50.46(a)(1)(i). Also as discussed in Section 3.2 of this safety evaluation, the licensee used acceptable means to analyze the post-LOCA TPBAR structural integrity.

The licensee presented the results for PCT, MLO, and CWO in Table 4 in Enclosures 3 and 4 to the LAR for Watts Bar Nuclear, Units 1 and 2, respectively. Details on the analysis results were provided in Attachment 1 to the Enclosures 3 and 4 to the LAR.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION To demonstrate compliance with 10 CFR 50.46(b)(1) through (b)(4), and thus GDC 35, the following criteria must be met:

1. PCT;

2. Maximum cladding oxidation;

3. Maximum hydrogen generation; and

4. Coolable geometry.

Each of the above four 10 CFR 50.46 criteria is discussed below.

Note that the FSLOCA EM does not address 10 CFR 50.46(b)(5), Long-term cooling.

Long-term cooling is dependent on the demonstration of the continued delivery of cooling water to the core. The licensee stated that the actions that are currently in place to maintain long-term cooling are not impacted by the application of the NRC-approved FSLOCA EM. During the NRC staffs review of the licensees requests, the staff did not identify any information that would call into question the plants ability to provide long-term cooling.

Peak Cladding Temperature The requirement of 10 CFR 50.46(b)(1) states that [t]he calculated maximum fuel element cladding temperature shall not exceed 2200 °F. The licensee stated that the analysis for PCT corresponds to a bounding estimate of the 95th percentile PCT at the 95-percent confidence level, and given that the resulting PCT is less than 2,200 °F, the analyses with the FSLOCA EM confirm that 10 CFR 50.46 acceptance criterion (b)(1) is satisfied. The licensee presented the results in Table 4 of Enclosures 3 and 4 to the LAR for Watts Bar Nuclear, Units 1 and 2.

Given that the maximum calculated PCT is below the 2,200 °F PCT limit, the NRC staff finds that the acceptance criterion of 10 CFR 50.46(b)(1) is met.

Maximum Cladding Oxidation The requirements of 10 CFR 50.46(b)(2) state, in part, that [t]he calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

The licensee stated that the analysis for MLO corresponds to a bounding estimate of the 95th percentile MLO at the 95-percent confidence level. Since the resulting MLO is less than 17 percent when converting the time-at-temperature to an equivalent cladding reacted using the Baker-Just correlation and adding the pre-transient corrosion, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(2) is satisfied. The licensee presented the results in Table 4 of Enclosures 3 and 4 to the LAR for Watts Bar, Units 1 and 2.

Given that the resulting MLO is below the 17 percent limit, the NRC staff finds that the acceptance criterion of 10 CFR 50.46(b)(2) is met.

Maximum Hydrogen Generation The requirement of 10 CFR 50.46(b)(3) states that [t]he 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 OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

The licensee stated that the analysis for CWO corresponds to a bounding estimate of the 95th percentile CWO at the 95-percent confidence level. Since the resulting CWO is less than 1 percent, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(3) is satisfied. The licensee presented the results in Table 4 of Enclosures 3 and 4 to the LAR for Watts Bar, Units 1 and 2.

Given that the resulting CWO is below the 1 percent limit, the NRC staff finds that the acceptance criterion of 10 CFR 50.46(b)(3) is met.

Coolable Geometry The requirement of 10 CFR 50.46(b)(4) states that [c]alculated changes in core geometry shall be such that the core remains amenable to cooling. The licensee stated that this criterion is met by demonstrating compliance with criteria 10 CFR 50.46(b)(1), (b)(2), and (b)(3), and by ensuring that fuel assembly grid deformation due to combined LOCA and seismic loads is specifically addressed.

Section 32.1 of the NRC-approved FSLOCA EM (Reference 16) documents that the effects of LOCA and seismic loads on the core geometry do not need to be considered unless fuel assembly grid deformation extends beyond the core periphery (i.e., deformation in a fuel assembly with no sides adjacent to the core baffle plates).

The licensee stated that the FSLOCA EM analysis does not affect the existing AOR related to combined LOCA and seismic loads (Reference 2). Section 4.2.1.3.5 of the Watts Bar UFSAR (Reference 20) states:

Only a small (outer) portion of the core experienced significant grid impact forces.

The maximum grid impact forces are required to be less than the allowable grid crush strength. A calculation of the maximum LOCA and seismic grid impact forces, combined using the square root sum of the squares method (in accordance with NUREG 0800, Section 4.2, Appendix A), demonstrated that the maximum value is less than the allowable grid strength for both the homogeneous core (RFA-2 with IFMs) and the mixed core (RFA-2 with IFMs and V+/P+ without IFMs).

The NRC staff finds that since the current licensing basis for combined LOCA and seismic loads is not affected by the FSLOCA EM analysis, the conclusion that no fuel assembly grid deformation extends beyond the core periphery remains valid.

Given that the criteria in 10 CFR 50.46(b)(1), (b)(2), and (b)(3) are met and that fuel assembly grid deformation due to combined LOCA and seismic loads is specifically addressed, the NRC staff finds that the acceptance criterion of 10 CFR 50.46(b)(4) is met.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.5 Comparison to Results from Analyses of Record A comparison between the results from the Analysis of Record (AOR) and the proposed FSLOCA EM was conducted.

The small break LOCA (SBLOCA) AOR results for Watts Bar, Units 1 and 2 originate from analyses performed using the NRC-approved methodology, WCAP-10054-P-A, Small Break ECCS Evaluation Model Using NOTRUMP Code, (Reference 21), and Addendum 2, Revision 1: Addendum to the Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model, (Reference 22). The results of the SBLOCA analysis can be found in the Watts Bar Nuclear Plant Updated Final Safety Analysis Report Chapter 15, Table 15.3-2 (Reference 23). The maximum PCT is reduced in the FSLOCA EM due to more realistic treatment of phenomena such as decay heat modeling.

The SBLOCA MLO results are substantially higher for the analyses with the FSLOCA EM than in the AORs. This can be primarily attributed to the AOR results which only consider the oxidation that accrues during the LOCA transient, whereas the results from the FSLOCA EM include the steady-state corrosion. The CWO results for both the AORs and the analyses with the FSLOCA EM indicate that there will be little to no CWO during the SBLOCA transient.

The NRC finds that the differences between the results of the SBLOCA AOR and FSLOCA EM are reasonable and expected.

The LBLOCA AOR for Watts Bar, Unit 1 was calculated using WCAP-12945-P-A, Volume I (Revision 2) and Volumes 2 through 5 (Revision 1), Code Qualification Document for Best-Estimate Loss of Coolant Analysis (Reference 24). The LBLOCA AOR for Watts Bar, Unit 2 was calculated using WCAP-16009-P-A, Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)

(Reference 25). There is a reduction in PCT from the AORs to the FSLOCA EM. This reduction can be attributed to such differences in the methodologies such as improvements to the statistical analysis method, improvements to the fuel temperature calculation, improvements to the axial power shape methodology, and improvements to the swelling, burst, and blockage models.

The MLO results for the LBLOCA AOR are higher for Unit 1 and lower for Unit 2 when compared to the FSLOCA EM results. The AOR results only considered the oxidation accrued during the LOCA transient, whereas the results from the FSLOCA EM included the steady-state corrosion. The increase in the MLO for Watt Bar, Unit 2 is primarily attributed to the contribution of the steady-state corrosion. The Watt Bar, Unit 1, AOR MLO result is based on a LOCA transient with a cladding temperature in excess of the PCT result and for which the time-at-temperature has been artificially increased (both artifacts of the AOR evaluation model). The excessive transient cladding temperature and artificial increase in time-at-temperature from the AOR evaluation model leads to a conservatively high result.

The LBLCOA CWO results for the Watts Bar, Unit 2, AOR and FSLOCA EM are relatively low.

For the Watts Bar, Unit 1, AOR, the CWO result is based on the same transient as the local oxidation result with a conservatively high cladding temperature. As such, the reduction in CWO for the analysis with the FSLOCA EM in comparison with the Watts Bar, Unit 1, AOR is OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION largely attributed to the use of a transient with lower temperature, as predicted by differences in the evaluation models.

The NRC staff finds that the differences in results (PCT, MLO, and CWO) between the AOR and FSLOCA for the LBLOCA are reasonable and are as expected.

3.6 Evaluation of Proposed Technical Specifications Changes In the Watts Bar, Unit 1, TS 4.2.1, the word Zircalloy is requested to be removed. The FSLOCA EM methodology considers ZIRLO cladding. Additional analysis would be needed for Zircalloy fuel rods. This proposed TS change is consistent with the FSLOCA EM methodology and is therefore acceptable.

The Watts Bar, Units 1 and 2, TS 5.9.5 list the LOCA evaluation model references. The licensee proposes to remove the references for the CQD and NOTRUMP EMs and insert the reference to the FSLOCA EM methodology. This proposed change is acceptable because the change would be consistent with using the FSLOCA EM.

Watts Bar, Unit 2 contains an operating license condition. The operating license condition sets the direction to use PAD4TCD to determine the operating limits until the Watts Bar, Unit 2 steam generators are replaced. The change to this operating license condition adds the direction to implement the FSLOCA methodology once the steam generators are replaced. This is an acceptable change to the operating license condition because it maintains the intent of the current requirement that an appropriate EM will be used when the steam generators are replaced for Unit 2.

3.7 Technical Conclusion The licensee requested to modify TS 5.9.5b, Core Operating Limits Report (COLR), to replace the existing NRC-approved LOCA methodologies with the NRC-approved LOCA methodology contained in WCAP-16996-P-A, Revision 1. The NRC staff concludes that the proposed TS change is acceptable as it changes from one set of NRC-approved methods to another NRC-approved method. The NRC staff's review has determined that the licensee appropriately applied the FSLOCA EM to Watts Bar, Units 1 and 2, and finds that the resulting analyses meet the criteria in 10 CFR 50.46 and GDC 35. In addition, the proposed revised TS 5.9.5, Core Operating Limits Report, continues to meet 10 CFR 50.36(c)(5) by providing controls necessary to assure operation of the facility in a safe manner.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Tennessee State official was notified of the proposed issuance of the amendment on August 14, 2020. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20, or change an inspection or SR. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission previously issued a proposed finding that the amendment involves no significant hazards consideration published in the Federal Register on June 2, 2020 (85 FR 33745), 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 Commission's 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.

7.0 REFERENCES

1. Letter from Polickoski, J. T., Tennessee Valley Authority (TVA) to U.S. Nuclear Regulatory Commission, Application to Implement the FULL SPECTRUM LOCA (FSLOCA) Methodology for Loss-of-Coolant Accident (LOCA) Analysis and New LOCA-Specific Tritium Producing Burnable Absorber Rod Stress Analysis Methodology (WBN-TS-19-04), January 17, 2020, ADAMS Accession No. ML20017A338 (not publicly available).

2. Letter from Barstow, J. to U.S. Nuclear Regulatory Commission, Response to NRC Request for Additional Information Regarding Application to Implement the FULL SPECTRUM LOCA (FSLOCA) Methodology for Loss-of-Coolant Accident (LOCA)

Analysis and New LOCA-specific Tritium Producing Burnable Absorber Rod Stress Analysis Methodology (WBN-TS-19-04) (EPID L-2020-LLA-0005), August 27, 2020, ADAMS Accession No. ML20240A324.

3. Letter from Barstow, J., TVA to U.S. Nuclear Regulatory Commission, Correction of Application to Implement the FULL SPECTRUM'1 LOCA (FSLOCA'1) Methodology for Loss-of-Coolant Accident (LOCA) Analysis and New LOCA-specific Tritium Producing Burnable Absorber Rod Stress Analysis Methodology (WBN-TS-19-04)

(EPID L-2020-LLA-0005), January 26, 2021, ADAMS Accession No. ML21027A143.

4. Letter from Green, K. J., U.S. Nuclear Regulatory Commission to Barstow, J., TVA, Watts Bar Nuclear Plant, Units 1 and 2 - Regulatory Audit Summary Related to Request to Implement FULL SPECTRUM' LOCA Methodology for Loss-of-Coolant Accident Analysis and TPBAR Stress Analysis, December 9, 2020, ADAMS Accession No. ML20322A023.

5. Email from Wentzel, M., U.S. Nuclear Regulatory Commission to Wells, R., TVA,

Subject:

Watts Bar Nuclear Plant, Units 1 and 2 - Request for Additional Information Regarding Request to Implement the FULL SPECTRUM' LOCA Methodology (EPID L-2020-LLA-0005), July 14, 2020, ADAMS Accession No. ML20196L862.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION

6. U.S. Nuclear Regulatory Commission, 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, NUREG-0800, March 2007, ADAMS Accession No. ML070550016.

7. U.S. Nuclear Regulatory Commission, Best-Estimate Calculations of Emergency Core Cooling System Performance, Regulatory Guide 1.157, May 1989, ADAMS Accession No. ML003739584.

8. U.S. Nuclear Regulatory Commission, Transient and Accident Analysis Methods, Regulatory Guide 1.203, February 2005, ADAMS Accession No. ML050230008.

9. U.S. Nuclear Regulatory Commission, Removal of Cycle-Specific, Parameter Limits from Technical Specifications, Generic Letter 88-16, October 4, 1988, ADAMS Accession No. ML031130447.

10. Letter from Gresham, J. A., Westinghouse Electric Company to U.S. Nuclear Regulatory Commission, Submittal of WCAP-17642-P-A/WCAP-17642-NP-A, Revision 1, Westinghouse Performance Analysis and Design Model (PAD5), November 2017, ADAMS Package Accession No. ML17335A334.

11. 2010 ASME Boiler Pressure Vessel Code Section III Division 1 Subsection NG.

12. Westinghouse Electric Company, Tritium Production Core (TPC) Topical Report, NDP-98-181, Revision 1, February 1999, ADAMS Accession No. ML16077A093.

13. U.S. Nuclear Regulatory Commission, Safety Evaluation Report related to the Department of Energys Topical Report on the Tritium Production Core, NUREG-1672, May 1999,

14. Letter from Martin, R. E., U.S. Nuclear Regulatory Commission to Scalice, J. A., TVA, Safety Evaluation of LOCTA_JR Code for Loss-of-Coolant Accident Analysis of Fuel Rods - Watts Bar Nuclear Plant, Unit 1, and Sequoyah Nuclear Plant, Units 1 and 2 (TAC Nos. MA9520, MA9583, MA9584), January 17, 2001, ADAMS Accession No. ML010170152.

15. Johnson, G. D., Evaluation of Mechanical Properties with the Larson-Miller Parameter, HEDL-TME 75-33, April 1975.

16. Letter from Gresham, J. A., Westinghouse Electric Company to U.S. Nuclear Regulatory Commission, Submittal of WCAP-16996-P-A/WCAP-16996-NP-A, Volumes I, II, III and Appendices, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology) (TAC No. ME5244)

(Proprietary/Non-Proprietary), October 2, 2017, ADAMS Package Accession No. ML17277A130.

17. Letter from Morey, D. C., U.S. Nuclear Regulatory Commission to Gresham, J. A.,

Westinghouse Electric Company,

Subject:

Revised Final Safety Evaluation for Westinghouse Electric Company Topical Report WCAP-16996-P/WCAP-16996-NP, OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Volumes I, II, and III, Revision 1, Realistic Loss-of-Coolant Accident Evaluation Methodology Applied to the Full Spectrum of Break Sizes (TAC NO. ME5244),

September 12, 2017, ADAMS Package Accession No. ML17207A124.

18. Letter from Gresham, J. A., Westinghouse Electric Company to Whitman, J.,

U.S. Nuclear Regulatory Commission, U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2017, LTR-NRC-18-30, July 18, 2018, ADAMS Accession No. ML19288A174.

19. Letter from Mercier, E. J., Westinghouse Electric Company to U.S. U.S. Nuclear Regulatory Commission, Information to Satisfy the FULL SPECTRUM LOCA (FSLOCA) Evaluation Methodology Plant Type Limitations and Conditions for 4-loop Westinghouse Pressurized Water Reactors (PWRs) (Proprietary/Non-Proprietary),

LTR-NRC-18-50, July 13, 2018, ADAMS Package Accession No. ML18198A038.

20. Watts Bar Nuclear Plant, Updated Final Safety Analysis Report, Amendment 3, Chapter 4, Reactor, October 2020, ADAMS Accession No. ML20323A311.

21. Letter from Rahe, E. P., Westinghouse Electric Company to Thomas, C. O.,

U.S. Nuclear Regulatory Commission, transmittal of WCAP-10054-P-A, Small Break ECCS Evaluation Model Using NOTRUMP Code, August 1985, ADAMS Accession No. ML100050586 (not publicly available).

22. Westinghouse Electric Company, Addendum to the Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model, WCAP-10054-P-A, Addendum 2, Revision 1, July 1997, ADAMS Accession No. ML19317C533 (not publicly available).

23. Watts Bar Nuclear Plant, Updated Final Safety Analysis Report, Amendment 3, Chapter 15, Accident Analyses, October 2020, ADAMS Accession No. ML20323A316.

24. Westinghouse Electric Company, Code Qualification Document for Best-Estimate Loss of Coolant Analysis, WCAP-12945-P-A, Volume I (Revision 2) and Volumes 2 through 5 (Revision 1), March 1998, ADAMS Package Accession No. ML093070051 (not publicly available).

25. Westinghouse Electric Company, Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),

WCAP-16009-P-A, January 2005, ADAMS Package Accession No. ML080630386 (not publicly available).

Principal Contributors: M. Panicker, NRR D. Woodyatt, NRR Date: February 26, 2021 OFFICIAL USE ONLY PROPRIETARY INFORMATION

ML21034A166 (Package);

ML21034A148 (Proprietary);

ML21034A169 (Non-proprietary)

OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DSS/SFNB/BC NRR/DSS/SNSB/BC NAME KGreen BAbeywickrama RLukes SKrepel DATE 02/02/2021 02/05/2021 10/23/2020 10/01/2020 OFFICE NRR/DSS/STSB OGC - NLO NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM w/comment NAME VCusumano DRoth UShoop KGreen DATE 02/05/2021 02/25/2021 02/26/2021 02/26/2021