ML20301A452

Non-Proprietary, Issuance of Amendment Nos. 300 and 300 to Revise Technical Specifications 6.2.C, Core Operating Limits Report

ML20301A452
Person / Time
Site:	Surry
Issue date:	10/28/2020
From:	Vaughn Thomas Plant Licensing Branch II
To:	Stoddard D Virginia Electric & Power Co (VEPCO)
Thomas V
References
EPID L-2019-LLA-0243
Download: ML20301A452 (29)

Text

OFFICIAL USE ONLY PROPRIETARY INFORMATION October 28, 2020 Mr. Daniel G. Stoddard Senior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

SURRY POWER STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENT NOS. 300 AND 300 TO REVISE TECHNICAL SPECIFICATIONS 6.2.C, CORE OPERATING LIMITS REPORT (EPID L-2019-LLA-0243)

Dear Mr. Stoddard:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 300 to Renewed Facility Operating License No. DPR-32 and Amendment No. 300 to Renewed Facility Operating License No. DPR-37 for the Surry Power Station (Surry), Unit Nos. 1 and 2, respectively. The amendments revise the Technical Specifications (TS) in response to your application dated October 30, 2019.

The amendments revise TS 6.2.C, Core Operating Limits Report, to add the Westinghouse Topical Report WCAP-16996-P-A, Revision 1, "Realistic LOCA [Loss of Coolant Accident]

Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FSLOCA' EM) to the list of methodologies approved for reference in the Core Operating Limits Report (COLR).

OFFICIAL USE ONLY PROPRIETARY INFORMATION to this letter contains proprietary information. When separated from Enclosure 3, this document is DECONTROLLED.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 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 with double brackets. The proprietary version of the safety evaluation is provided as. Accordingly, the NRC staff has also prepared a non-proprietary version of the safety evaluation, which is provided in Enclosure 4.

The Commissions monthly Federal Register notice will include the notice of issuance.

Sincerely,

/RA/

Vaughn V. Thomas, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-280 and 50-281

Enclosures:

1. Amendment No. 300 to DPR-32

2. Amendment No. 300 to DPR-37

3. Safety Evaluation (Proprietary)

4. Safety Evaluation (Non-Proprietary) cc w/o Enclosure 3: Listserv

VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 300 Renewed License No. DPR-32

1.

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

A.

The application for amendment by Virginia Electric and Power Company (the licensee) dated October 30, 2019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specification as indicated in the attachment to this license amendment of the Renewed Facility Operating License No. DPR-32 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications Contained in Appendix A, as revised through Amendment No. 300 are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

The 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/

Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. DPR-32 and the Technical Specifications Date of Issuance: October 28, 2020

VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 300 Renewed License No. DPR-37

1.

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

A.

The application for amendment by Virginia Electric and Power Company (the licensee) dated October 30, 2019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specification as indicated in the attachment to this license amendment of the Renewed Facility Operating License No. DPR-37 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications Contained in Appendix A, as revised through Amendment No. 300 are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

The 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/

Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. DPR-37 and the Technical Specifications Date of Issuance: October 28, 2020

ATTACHMENT TO SURRY POWER STATION, UNIT NOS. 1 AND 2 LICENSE AMENDMENT NO. 300 RENEWED FACILITY OPERATING LICENSE NO. DPR-32 DOCKET NO. 50-280 AND LICENSE AMENDMENT NO. 300 RENEWED FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NO. 50-281 Replace the following pages of the Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contained marginal lines indicating the areas of change.

Renewed Facility Operating License No. DPR-32 REMOVE INSERT 3

3 Renewed Facility Operating License No. DPR-37 REMOVE INSERT 3

3 TSs REMOVE INSERT Figure 3.12-2 6.2-2 6.2-2 Surry - Unit 1 Renewed License No. DPR-32 Amendment No. 300

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2587 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 300 are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.

D. Records The licensee shall keep facility operating records in accordance with the requirements of the Technical Specifications.

E. Deleted by Amendment 65 F. Deleted by Amendment 71 G. Deleted by Amendment 227 H. Deleted by Amendment 227 I. Fire Protection The licensee shall implement and maintain in effect the provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report and as approved in the SER dated September 19, 1979, (and Supplements dated May 29, 1980, October 9, 1980, December 18, 1980, February 13, 1981, December 4, 1981, April 27, 1982, November 18, 1982, January 17, 1984, February 25, 1988, and Surry - Unit 2 Renewed License No. DPR-37 Amendment No. 300 E. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such by product and special nuclear materials as may be produced by the operation of the facility.

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power Levels not in excess of 2587 megawatts (thermal)

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 300 are hereby incorporated in this renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.

D. Records The licensee shall keep facility operating records in accordance with the Requirements of the Technical Specifications.

E. Deleted by Amendment 54 F. Deleted by Amendment 59 and Amendment 65 G. Deleted by Amendment 227 H. Deleted by Amendment 227

TS 6.2-2 The analytical methods used to determine the core operating limits identified above shall be those previously reviewed and approved by the NRC, and identified below.

The CORE OPERATING LIMITS REPORT will contain the complete identification for each of the TS referenced topical reports used to prepare the CORE OPERATING LIMITS REPORT (i.e., report number, title, revision, date, and any supplements). The core operating limits shall be determined so that applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided for information for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

REFERENCES 1.

VEP-FRD-42-A, Reload Nuclear Design Methodology 2.

WCAP-16009-P-A, Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),

(Westinghouse Proprietary).

3.

WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, (W Proprietary) 4.

WCAP-10079-P-A, NOTRUMP, A Nodal Transient Small Break and General Network Code, (W Proprietary) 5.

WCAP-12610-P-A, VANTAGE+ Fuel Assembly Report, (Westinghouse Proprietary) 6.

VEP-NE-2-A, Statistical DNBR Evaluation Methodology 7.

WCAP-16996-P-A, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology). (Westinghouse Proprietary) 8.

DOM-NAF-2-A, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code, including Appendix B, Qualification of the Westinghouse WRB-l CHF Correlation in the Dominion VIPRE-D Computer Code, and Appendix D, Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code 9.

WCAP-8745-P-A, Design Bases for Thermal Overpower Delta-T and Thermal Overtemperature Delta-T Trip Function

10. WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, Optimized ZIRLO, (Westinghouse Proprietary)

Amendment Nos. 300 and 300

ENCLOSURE 4 (NON-PROPRIETARY)

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO THE FULL SPECTRUM LARGE BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS METHODOLOGY FACILITY OPERATING LICENSE NOS. DPR-32 AND DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY DOMINION ENERGY VIRGINIA SURRY POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-280 AND 50-281 Proprietary information has been redacted from this document pursuant to Section 2.390 of Title 10 of the Code of Federal Regulations.

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

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO THE FULL SPECTRUM LARGE BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS METHODOLOGY AMENDMENT NO. 300 TO RENEWED FACILITY OPERATING LICENSE NOS. DPR-32 AMENDMENT NO. 300 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY DOMINION ENERGY VIRGINIA SURRY POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-280 AND 50-281

1.0 INTRODUCTION

By letter dated October 30, 2019 (Reference 1), Virginia Electric and Power Company (Dominion Energy Virginia, the licensee) submitted a license amendment request (LAR) to revise Technical Specifications (TS) 6.2.C, Core Operating Limit Report, for the Surry Power Station (Surry), Unit Nos. 1 and 2.

The amendments would revise TS 6.2.C to add Westinghouse Topical Report WCAP-16996-P-A, Revision 1, "Realistic LOCA [Loss of Coolant Accident] Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology) (Reference 2) to the list of methodologies approved for reference in the Core Operating Limits Report (COLR) and to delete Figure 3.12-2, Hot Channel Factor Normalized Operating Envelope. The added reference identifies the analytical methods used to determine core operating limits for the large break loss of coolant accident (LBLOCA) event described in the Surry Updated Final Safety Analysis Report (UFSAR), Section 14.5.1, Major Reactor Coolant System Pipe Ruptures.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION

2.0 REGULATORY EVALUATION

2.1 Description of the Licensees Proposed Change The licensee proposed adding the Westinghouse Topical Report, WCAP-16996-P-A, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes, (Reference 2) to the Reference list in TS 6.2.C. This change would allow the licensee to use the FULL SPECTRUM LOCA Evaluation Methodology (FSLOCA' EM) described in the Topical Report to determine core operating limits for the LBLOCA described in UFSAR, Section 14.5.1. The current LBLOCA analysis method, WCAP-16009-P-A, Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),

(Reference 3) would be retained in order to allow each unit to transition to the new method between cycles. In addition, the licensee proposed to remove Figure 3.12-2 and a legacy reference that is no longer required, VEP-NE-3-A, "Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code," from the Surry TS 6.2.C.

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

Regulations The regulations in 10 CFR 50.36, Technical specifications, require that TSs include items in the following categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports.

The regulations in 10 CFR 50.36(c)(5), Administrative controls, provide provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. This applies to the list of references to approved methods to be used to determine the core operating limits contained in the COLR.

The regulations in 10 CFR 50.46(b) require, in part, that, during LOCA events, the following criteria are met:

(1)

For peak cladding temperature, the calculated maximum fuel element cladding temperature shall not exceed 2200 °F [degrees Fahrenheit].

(2)

For maximum cladding oxidation, the calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

(3)

For 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

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

(4)

For coolable geometry, the calculated changes in core geometry shall be such that the core remains amenable to cooling.

Guidance Documents The NRC issued construction permits for Surry Units 1 and 2 before May 21, 1971; consequently, Surry Units 1 and 2 were not subject to the requirements in Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix A, General Design Criteria [GDC] for Nuclear Power Plants (see SECY-92-223, Resolution of Deviations Identified during the Systematic Evaluation Program, dated September 18, 1992 (Reference 13). Surry Units 1 and 2 meet the intent of the GDC published in 1967 (draft GDC).

The NRC staff used the following 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," March 2007 (Reference 4).

NRC Regulatory Guide (RG) 1.157, "Best-Estimate Calculations of Emergency Core Cooling System Performance," May 1989. (Reference 5)

NRC Regulatory Guide 1.203, "Transient and Accident Analysis Methods,"

December 2005. (Reference 6)

NRC Generic Letter 88-16, "Removal of Cycle Specific Parameter Limits from Technical Specifications," dated October 4, 1988. (Reference 7)

NRC Information Notice 97-09, Main Steam Safety Valve (MSSV) Setpoints and Performance Issues Associated with Long MSSV Inlet Piping.

3.0 TECHNICALEVALUATION 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 2.0 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 Evaluation Model (EM) to Surry and associated acceptance criteria of 10 CFR 50.46(b)(1) through (4) are satisfied.

3.1 Description of FSLOCA' Methodology As described in WCAP-16996-P-A, Revision 1, the purpose of the Full Spectrum LOCA (FSLOCA') EM is to build on the ASTRUM EM, described in WCAP-16009-P-A (Reference 3),

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION by extending the applicability of the WCOBRA/TRAC Code to include the treatment of small break LOCA (SBLOCA) and intermediate break LOCA (IBLOCA) scenarios. The term Full Spectrum specifies that the new EM is intended to resolve the full spectrum of LOCA 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' methodology include 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 2 times the pipe area.

The licensee is currently using the ASTRUM EM methodology described in WCAP-16009-P-A (Reference 3) to perform its LBLOCA licensing analyses. The licensee proposes to use the WCAP-16996-P-A (Reference 2) methodology in order to fulfill a prior commitment to the NRC to update its licensing basis to account for thermal conductivity degradation (TCD). FSLOCA' EM is an analysis methodology for LOCAs that has been reviewed and approved by the NRC, and further discussion of the methodology and its application to LOCAs can be found in WCAP-16996-P-A (Reference 2). The FSLOCA' methodology divides the break spectrum into two parts: Region I (small break LOCA), and Region II (large break LOCA). The licensee proposes to use this methodology only for the Region II analyses.

3.2 Analysis The analyses for Surry, Units 1 and 2 were performed by the licensee in accordance with the NRC-approved methodology in WCAP-16996-P-A. The analyses were performed assuming both loss of outside power (LOOP) and offsite power available (OPA). The FSLOCA' EM, as approved by the NRC, is designed to perform analyses for both Regions I and II. Nonetheless, while the FSLOCA' methodology divides the break spectrum into the two regions, independent analyses are performed to determine results within each region. Therefore, given that the FSLOCA' analyses for Region I and Region II are separable, and do not influence each other, NRC staff finds it acceptable that the licensee is only performing analyses for Region II.

The major plant parameter and analysis assumptions used in the Surry, Units 1 and 2 FSLOCA' EM are provided in Tables 1 through 4 of Attachment 4 to the LAR. Tables 5 and 6 of Attachment 4 to the LAR provide results and sequence of events and Table 7 summarizes the Region II LOOP and OPA uncertainty values used and the Decay Heat Uncertainty Multiplier analysis.

The NRC staffs review concluded that the input assumptions such as Core Parameters, Reactor Coolant System Parameters, and Containment Parameters and the uncertainty values used in the analysis were reasonable and acceptable based on consistency with the Surry plant configuration and current licensing basis.

The following table summarizes the peak cladding temperature (PCT), maximum local oxidization (MLO), and core wide oxidization (CWO) results from the Surry analyses. The limiting PCT result is 1844 degrees °F for the Region II analysis, however, an error correction from the gamma energy redistribution multiplier is estimated to increase the Region II analysis

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION PCT by 31°F (see Section 3.4 of this safety evaluation, Limitation and Condition 2 evaluation for further discussion) with a total analysis PCT result of 1875 °F.

Results Region II Value (OPA)

Region II Value (LOOP) 95/95 PCT 1844°F + 31°F = 1875°P 1817°F + 31°F = 1848°F 95/95 MLO 6.2%

6.5%

95/95 CWO 0.37%

0.43%

The NRC staff reviewed the analysis and the submittals provided by the licensee and determined that the Region II analysis for Surry, Units 1 and 2 was performed by the licensee in accordance with the NRC-approved methodology, and accordingly, that the proposed change maintains sufficient safety margins and the relevant criteria of 10 CFR 50.46(b) are satisfied.

3.3 TS Evaluation 3.3.1 The proposed change replaces the current Reference 7 for TS 6.2, VEP-NE-3-A, "Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code," with a new methodology, WCAP-16996-P-A, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes, (Reference 2). The licensee states that VEP-NE-3-A is a legacy method that is no longer used at Surry, and the WCAP-16996-P-A methodology would become the new method for analyzing its LBLOCA analysis. The NRC staff considers the proposed TS change to be acceptable and notes that it would continue to provide administrative controls consistent with 10 CFR 50.36(c)(5).

The proposed amendment would allow the Region II part of WCAP-16996-P-A methodology to be utilized to support future core reloads at Surry. This proposed change assures the core operating limits have been calculated in accordance with NRC-approved methodologies. The current WCAP-16009-P-A LBLOCA (Reference 3) analysis method would be retained in order to allow both Units 1 and 2 for Surry to transition to the new method (Reference 2) between cycles rather than having to update the method for both units at the same time. The NRC staff considers the proposed changes to revise the list of methodologies to reflect those necessary for future cycle specific operating limits in TS 6.2.C to be acceptable because the change would continue to provide administrative controls consistent with 10 CFR 50.36(c)(5).

3.3.2 The licensee proposed to remove Figure 3.12-2 from the Surry TS. This figure displays the Hot Channel Factor K(z) curve. The licensee stated that Figure 3.12-2 should have been removed as part of a prior license amendment. The NRC staff considers the proposed deletion of an unused figure in the TS to be acceptable because the proposed change is administrative in nature, superseded by NRC-approved methods, and would continue to meet 10 CFR 50.36(c)(5).

3.4 Limitations and Conditions The safety evaluation for WCAP-16996-P-A, Revision 1 (Reference 8), contains 15 limitations and conditions that must be met in order for a licensee to implement the NRC-approved FSLOCA' EM.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION A summary of each limitation and condition and how it has been met, as stated by the licensee in its application dated October 30, 2019, and the associated NRC staff findings are provided below. The NRC staff confirmed the statements provided by the licensee in the LAR (Reference

1) by reviewing the analysis documentation associated with the review. The NRC staff also performed an audit of the review and the details of the audit are provided in the audit plan and the audit report summary (Reference 9) and (Reference 10), respectively.

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

The analysis for Surry, 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).

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 for Limitation and Condition 1.

Limitation and Condition 2 - Applicability with Regard to Type of PWR Plants Applicability of FSLOCA' EM is defined in terms of PWR-type plants so that analysis is approved for Westinghouse-designed 3-loop and 4-loop PWRs [Pressurized Water Reactors] with cold-side injection only. Plant-specific licensing actions referencing FSLOCA' analyses should include a statement summarizing the extent to which the FSLOCA' methods and modeling were followed, and justification for any departures.

Surry, Units 1 and 2 are Westinghouse-designed 3-loop PWRs with cold-side injection, so they eligible to use the FSLOCA' EM. The analysis for Surry, Units 1 and 2 utilizes the NRC-approved FSLOCA' methodology, with the following exceptions.

Since the safety evaluation was issued on the FSLOCA' methodology (WCAP-16996-P-A, Revision 1), several changes and corrections have been made to the FSLOCA' EM. In a letter to the NRC (Reference 11), Westinghouse reported the impact of errors in the emergency core cooling system (ECCS) evaluation models used by Westinghouse Electric Company. A description of the error corrections to the Westinghouse FSLOCA' EM is provided in Reference 11. The NRC staff reviewed the errors and the resolution as documented in Reference 11 and identified concluded that only the following applies to Region II analyses of FSLOCA' EM:

Conservation of Non-Condensable Gas: Westinghouse identified that there existed an imbalance in the non-condensable gas mass which could occur in the WCOBRA/TRAC-TF2 code because it does not have the capability to implement the vapor property functions for temperatures below 32°F under certain conditions.

In the LAR (Reference 1), the licensee stated that this error was corrected in the loop components for the WCOBRA/TRAC-TF2 code. The resolution of this issue represents a Non-Discretionary Change in the Evaluation Model as described in Section 4.1.2 of WCAP-13451, Westinghouse Methodology for Implementation

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION of10 CFR 50.46. The error had minimal impact on LOCA transient calculations, leading to an estimated peak cladding temperature impact of 0°F. The NRC staff finds that changes to the conservation of non-condensable gas are appropriate and acceptable.

After completion of the analysis for Surry, Units 1 and 2, Westinghouse subsequently discovered three errors in the WCOBRA/TRAC-TF2 code. The first error was regarding the calculation of radiation heat transfer to liquid. The second error was regarding vapor temperature resetting, where the vapor temperature could incorrectly be reset to the saturation temperature for heat transfer calculations. The licensee in the LAR (Reference 1) stated that these two errors were found to have a negligible impact on analysis results.

The third error impacted the gamma energy redistribution multiplier on the hot rod and hot assembly power used for the Surry, Units 1 and 2, analyses. This error resulted in up to about a 5 percent deficiency in the hot rod and hot assembly rod linear heat rates on a run-specific basis, depending on the as-sampled value for the multiplier uncertainty.

The licensee stated in the LAR (Reference 1) that the error impacting the gamma energy redistribution multiplier had only a limited impact on the power modeled for a single assembly in the core and the error correction had a negligible impact on the system thermal-hydraulic response during the postulated LOCA for the Region II analysis. The PCT impact from the error correction was found by the licensee to be different for the transient phases (i.e., blowdown versus reflood) based on parametric PWR sensitivity studies. The correction of the error was estimated to increase the Region II analysis PCT by 31 °F, leading to a final PCT analysis result of 1875 °F for Region II. The NRC staff reviewed the Region II analysis and confirmed that the analysis results, with the inclusion of the error correction, continue to demonstrate compliance with 10 CFR 50.46(b)(1) PCT limit of 2200° F.

In summary, the NRC staff finds that the licensee has appropriately applied the FSLOCA' EM with the changes described above. Therefore, the NRC staff finds that the licensee has met the requirements for Limitation and Condition 2.

Limitation and Condition 3 - Applicability for Containment Pressure Modeling 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.

The NRC staff reviewed the information provided in the Attachment 4 of the LAR (Reference 1) and confirmed that the containment pressure calculation for the Surry Units 1 and 2 analysis was performed consistent with the NRC-approved methodology.

[The containment pressure is calculated for each LOCA transient in the analysis using the COCO code. The COCO containment code is integrated into the WC0BRA/TRAC-TF2 thermal-hydraulic code.] The licensee stated that appropriate design parameters and conditions were modeled, which can reduce the containment

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION pressure. A minimum initial temperature associated with normal full-power operating conditions was modeled, and only containment coatings which are qualified and outside of the break zone-of-influence were credited.

The NRC staff finds that the licensee used NRC-approved methodology (i.e., the COCO or LOTIC2 models) for the Region II containment pressure calculation with the appropriated design parameters and conditions. Therefore, the NRC staff finds that the licensee has met the requirements for Limitation and Condition 3.

Limitation and Condition 4 - Decay Heat Modeling The decay heat uncertainty multiplier ((

)). The analysis simulations for the FSLOCA EM cannot be applied for transient time longer than 10,000 seconds following shutdown unless the decay heat model is shown to be acceptable for the analyzed core conditions.

The sampled value of the decay heat uncertainty multiplier as applied for the limited runs in Region I and Region II analysis results will be provided in the license amendment submittal in units of (sigma) and absolute units.

The licensee in the LAR (Reference 1) stated that the decay heat uncertainty multiplier

((

)) for the Surry Units 1 and 2 Region II analysis. 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 II analysis results have been provided in units of and approximate absolute units in Table 7 of Attachment 4 to the LAR (Reference 1).

The NRC staff confirmed that that the licensee appropriately modeled the decay heat per the limitation and condition and correctly 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 for Limitation and Condition 4.

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

)) respectively.

The NRC staff reviewed the analysis that was performed by the licensee and confirmed that maximum assembly and rod length-average burnup is less than or equal to ((

)) respectively, for Surry Units 1 and 2,above limit.

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

Limitation and Condition 6 - WCOBRA/TRAC-TF2 Interface with PAD 5.0 The fuel performance data for analyses with the FSLOCA' EM should be based on the PAD5 code (Reference 13), which includes the effect of thermal conductivity degradation. The nominal fuel pellet average temperatures and rod internal pressures

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION should be the maximum values, and the generation of all the PAD5 fuel performance data should adhere to the latest NRC-approved PAD5 methodology.

PAD5 fuel performance data was utilized in the Surry, 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 4, and the analyzed rod internal pressures were calculated in accordance with Section 7.5.2 of Reference 4.

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.

Limitation and Condition 7 - Interfacial Drag Uncertainty in Region I Analyses The YDRAG uncertainty parameter should be ((

)) including the comprehensive list of

((

)) given in Table 29.2.3-1 of Reference 2.

The licensee proposed to use the FSLOCA' methodology only for Region II analyses.

Accordingly, this Limitation and Condition is not applicable, and the licensee did not need to perform a Region I uncertainty analysis in this application of the FSLOCA' EM.

Limitation and Condition 8 - Biased Uncertainty Contributors in Region I Analyses The ((

))

The licensee proposed to use the FSLOCA' methodology only for Region II analyses.

Accordingly, this Limitation and Condition is not applicable, and the licensee did not need to perform a Region I uncertainty analysis in this application of the FSLOCA' EM.

Limitation and Condition 9 - Effect of Bias in Applications for Region I For PWR designs which are not Westinghouse 3-loop PWRs, a confirmatory analysis will be performed to assess the effect associated with the ((

)) for the plant design being analyzed.

Surry Units 1 and 2 are Westinghouse 3-loop PWRs, and the licensee proposed to use the FSLOCA' methodology only for Region II analyses. Therefore, this Limitation and Condition does not apply to the licensees LAR.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Limitation and Condition 10 - Boundary Between Region I and Region II Breaks For PWR designs which are not Westinghouse 3-loop PWRs, a confirmatory evaluation will be performed to demonstrate that the applied break size boundary between Region I and Region II analyses serves the intended goal of ((

)).

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

Surry Units 1 and 2 are Westinghouse 3-loop PWRs, and the licensee proposed to use the FSLOCA' methodology only for Region II analyses. Therefore, the first part of this Limitation and Condition does not apply to the licensees LAR.

Given that the minimum sampled break area for the Surry, Units 1 and 2 Region II analysis is 1 ft2, the NRC staff finds that the licensee has met the applicable requirements of Limitation and Condition 10.

Limitation and Condition 11 - ((

)) in Uncertainty Analyses for Region II and Documentation of Reanalysis Results for Region I and Region II There are various aspects of this Limitation and Condition, which are summarized below:

1) The ((

)) seed, and the analysis inputs and ((

)) to be used for the Region I and Region II uncertainty analyses will be declared and documented prior to performing the uncertainty analyses. The ((

)) will not be adjusted as a result of the outcome.

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, MLO, and CWO which caused the input changes will be provided. These preliminary values are not subject to 10 CFR 50, 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. This Limitation and Condition was met for the Surry, Units 1 and 2 analyses as follows:

1) The ((

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

)) and the 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.

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3) The plant operating ranges which were sampled within the uncertainty analyses are provided for Surry Units 1 and 2, in Table 1 of Attachment 4 of the LAR (Reference 1).

The NRC staffs review of the analysis and the information the licensee provided confirmed that the analysis inputs were not changed once they were declared and documented. 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.

Limitation and Condition 12 - Steam Generator Heat Removal During SBLOCAs In plant-specific applications, a check will be performed to confirm the effects associated with dynamic pressure losses from the steam generator secondary-side to the main steam safety valves are properly considered and adequately accounted for in analysis with the FSLOCA' EM consistent with NRC Information Notice 97-09, Main Steam Safety Valve (MSSV) Setpoints and Performance Issues Associated with Long MSSV Inlet Piping.

The licensee has stated in Attachment 4 (Page 7 of 33) of the LAR (Reference 1) that a bounding plant-specific dynamic pressure loss from the steam generator secondary-side to the main steam safety valves was modeled in the Surry Units 1 and 2 analysis.

Given that the licensee has declared and documented the use of the bounding pressure loss in the LAR, the NRC staff finds that the licensee has met the requirements of Limitation and Condition 12.

Limitation and Condition 13 - Upper Head Spray Nozzle Loss Coefficient In plant-specific applications of the FSLOCA EM, 1) the ((

)) in the PWR model used to perform the design-basis LOCA transient calculations, to capture the proper core two-phase level response should the core uncover. Additionally, the ((

)) in such calculations.

The ((

)) in the analyses for Surry Units 1 and 2. The ((

)) in the analyses.

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

Limitation and Condition 14 - Correlation for Oxidation For analyses with 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

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 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 the 17 percent limit.

• The licensee has stated in Attachment 4 (Page 7 of 33) of the LAR that for the Surry 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 CPR 50.46 local oxidation acceptance criterion of 17 percent.

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.

Limitation and Condition 15 - LOOP versus OPA Treatment in Uncertainty Analyses for Region II 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 statistical analysis must adhere to the Limitation and Condition as specified in the NRC-approved methodology for the FSLOCA' EM. For each set, the calculated statistical results at the 95/95 probability, confidence level should be demonstrated to comply with regulatory limits for PCT, MLO, and CWO. Specifically, the ((

))

The Region II uncertainty analysis for Surry, Units 1 and 2 was performed by the licensee twice; once assuming a LOOP and once assuming OPA. The results from both analyses that were performed are in compliance with the current 10 CFR 50.46 acceptance criteria. The licensee stated in the LAR (Reference 1) that the statistical analysis adhered to the Limitations and Conditions as specified in the NRC-approved methodology for the FSLOCA' EM. The results for this limiting condition analysis are in Table 7 of the LAR (Reference 1).

Given that the licensee has performed the Region II analysis for both LOOP and OPA and that the results from both 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.

3.5 Results and Compliance with 10 CFR 50.46 The licensee presented the results for PCT, MLO, and CWO in Table 5 of attachment 4 to the LAR (Reference 1) for Surry, Units 1 and 2.

To demonstrate compliance with 10 CFR 50.46(b)(1) through (b)(4), the following criteria must be met:

1. The calculated maximum fuel element cladding temperature must not exceed 2200° F.

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2. The calculated total oxidation of the cladding must nowhere exceed 0.17 times the total cladding thickness before oxidation.

3. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam must 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.

4. Calculated changes in core geometry must be such that the core remains amenable to cooling.

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

Note that the FSLOCA' EM is not approved to demonstrate compliance with 10 CFR 50.46 acceptance criterion (b)(5) related to the long-term cooling.

Peak Cladding Temperature The requirement of 10 CFR 50.46(b)(1) states, The 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 documented its results in Table 5 of Attachment 4 the LAR (Reference 1) for Surry, 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 10 CFR 50.46(b)(2) states, in relevant 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 after converting the time-at-temperature to an ECR 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 5 of Attachment 4 of the LAR (Reference

1) for Surry, 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, 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.

The licensee stated that the analysis for CWO corresponds to a bounding estimate of the 95th percentile CWO at the 95-percent confidence level. The analysis confirms that the resulting CWO is less than 1 percent and that the 10 CFR 50.46 acceptance criterion (b)(3) is satisfied.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The licensee presented the results in Table 5 of Attachment 4 of the LAR (Reference 1).

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, Calculated 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 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 to inboard assemblies beyond the core periphery (i.e., deformation in a fuel assembly with no sides adjacent to the core baffle plates). The licensee stated that inboard grid deformation due to the combined LOCA and seismic loads was calculated to not occur for Surry, Units 1 and 2. Given that the criteria in 10 CFR 50.46(b)(1), (b)(2), and (b)(3) are met and the fuel assembly grid deformation due to the combined LOCA and seismic loads is specifically addressed, the NRC staff finds the acceptance criterion of 10 CFR 50.46(b)(4) is met.

4.0 TECHNICAL CONCLUSION The licensee proposed to modify TS 6.2.C to replace the existing NRC-approved LOCA methodology (Reference 3) with the NRC-approved FSLOCA' EM (Reference 2). The NRC staff concludes that the proposed change is acceptable because the new methodology is an NRC-approved method. The NRC staff review has determined that the licensee appropriately applied the FSLOCA' EM to Surry, Units 1 and 2; and finds that the resulting analysis meets 10 CFR 50.46(b)(1) through (4) requirements. In addition, NRC staff finds that removal of unused Figure 3.12-2 and a legacy reference, VEP-NE-3-A, Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code, are acceptable. The NRC staff also finds the proposed changes continue to meet 10 CFR 50.36(c)(5) by providing provisions necessary to assure operation of the facility in a safe manner.

5.0 STATE CONSULTATION

In accordance with the Commissions regulations, the NRC notified an official from the Virginia Division of Radiological Health of the proposed issuance of the amendment. On September 21, 2020, the Virginia official confirmed that the Commonwealth of Virginia had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the 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 previously issued a proposed finding that the amendments involve no significant

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION hazards consideration, published in the Federal Register on December 31, 2019 (84 FR 72384),

and the agency has received no public comments on this finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Under 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

7.0 CONCLUSION

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

8.0 REFERENCES

1.

Virginia Electric and Power Company, letter to U.S. Nuclear Regulatory Commission (NRC), "Surry Power Station, Units 1 and 2 - Proposed License Amendment Request - Addition of Analytical Methodology to the Core Operating Limits Report for a Large Break Loss of Coolant Accident (LBLOCA)," October 30, 2019 (ADAMS Accession No. ML19309D196).

2.

Westinghouse Electric Company (WEC), Letter and Report to NRC, "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)," November 30, 2016 and October 2, 2017 (ADAMS Pkg Accession No. ML17277A130).

3.

Westinghouse Electric Company, letter to NRC, "WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," January 2005 (ADAMS Accession No. ML080650171).

4.

U.S. Nuclear Regulatory Commission, "NUREG-0800 - Chapter 15, Revision 3,

[Standard Review Plan], Section 15.6.5, Loss-of-Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary, March 2007" (ADAMS Accession No. ML070550016).

5.

U.S. Nuclear Regulatory Commission, "Regulatory Guide 1.157 (Task RS 701-4),

Best-Estimate Calculations of Emergency Core Cooling System Performance,"

Revision 0, May 1989 (ADAMS Accession No. ML003739584).

6.

U.S. Nuclear Regulatory Commission, "Regulatory Guide 1.203 Transient and Accident Analysis Methods," Revision 0, December 2005 (ADAMS Accession No. ML053500170).

7.

U.S. Nuclear Regulatory Commission, "NRC Generic Letter 1988-16: Removal of Cycle-Specific Parameter Limits From Technical Specifications," dated October 4, 1988 (ADAMS Accession No. ML031200485).

8.

U.S. Nuclear Regulatory Commission, letter to J. Gresham, WEC, "Revised Final Safety Evaluation for Westinghouse Electric Company Topical [Report]

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION WCAP-16996-P/WCAP-16996-NP, Volumes I, II and III, Revision I, Realistic Loss-of-Coolant Accident [LOCA] Evaluation Methodology [Applied to the Full Spectrum of Break Sizes]," dated September 12, 2017 (ADAMS Pkg Accession No. ML17207A124).

9.

U.S. Nuclear Regulatory Commission, letter to D. Stoddard, Dominion Energy, "Surry Nuclear Power Station, Unit 1 and 2 And North Anna Power Station, Units 1 and 2 - Audit Re: Proposed License Amendment Request for the Addition of Westinghouse Topical Report WCAP-16996-P-A, Revision 1, to the Core Operating Limits Report,"

April 17, 2020 (ADAMS Accession No. ML20099F873).

10.

U.S. Nuclear Regulatory Commission, "Audit Results Summary Report, North Anna Power Station and Surry Power Station, Request to Implement Fuel Vendor Independent Evaluation Model for Small Break Loss-of-Coolant Accident Analysis,"

October 2020 (ADAMS Accession No. ML20272A227).

11.

Westinghouse Electric Company, letter to NRC, "LTR-NRC-18-30 U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2017 (Non-Proprietary)," dated July 18, 2018 (ADAMS Accession No. ML19288A174).

12.

Westinghouse Electric Company, "WCAP-13451, Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting," October 1992 (ADAMS Pkg Accession No. ML20116C922).

13.

Westinghouse Electric Company, "WCAP-17642-NP, Revision 1 Westinghouse Performance Analysis and Design Model (PAD5), dated May 2016 (ADAMS Accession No. ML17207A124).

Principal Contributor:

Fred Forsaty, NRR Date issued: October 28, 2020