ML20325A095

Issuance of Amendment to Revise Technical Specifications to Allow Usage of a Small Break Loss of Coolant Accident Methodology (Non-Proprietary)

ML20325A095
Person / Time
Site:	North Anna
Issue date:	03/19/2021
From:	Geoffrey Miller Plant Licensing Branch II
To:	Stoddard D Virginia Electric & Power Co (VEPCO)
Miller G, NRR/DORL/LPL2-1, 415-2020
Shared Package
ML20325A088	List: ML20325A095
References
EPID L-2018-LLA-0195, EPID L-2018-LLA-0215
Download: ML20325A095 (65)

Text

OFFICIAL USE ONLY - PROPRIETARY INFORMATION March 19, 2021 Mr. Daniel G. Stoddard Senior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Blvd.

Glen Allen, VA 23060-6711

SUBJECT:

NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2, SURRY POWER STATION UNIT NOS. 1 AND 2 ISSUANCE OF AMENDMENT TO REVISE TECHNICAL SPECIFICATIONS TO ALLOW USAGE OF A SMALL BREAK LOSS OF COOLANT ACCIDENT METHODOLOGY (EPIDS L-2018-LLA-0195 AND L-2018-LLA-0215)

Dear Mr. Stoddard:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment Nos. 287 and 270 to Renewed Facility Operating License Nos. NPF-4 and NPF-7 for the North Anna Power Station (North Anna), Unit Nos. 1 and 2, respectively, and Amendment Nos. 303 and 303 to Renewed Facility Operating License Nos. DPR-32 and DPR-37, for the Surry Power Station (Surry), Unit Nos. 1 and 2, respectively. These amendments are in response to your application dated July 12, 2018, for North Anna, and July 31, 2018, for Surry, respectively. The requests were supplemented by letters dated July 9, 2019, May 28, 2020, and October 22, 2020.

The amendments revise North Anna and Surry Technical Specifications to include Framatome Topical Report EMF-2328(P)(A), Revision 0, PWR [pressurized water reactor] Small Break LOCA [loss-of-coolant accident] Evaluation Model, S-RELAP5 Based, as supplemented by the North Anna plant-specific application report ANP-3467P, Revision 0, North Anna Fuel-Vendor Independent Small Break LOCA Analysis Licensing Report, and Surry plant-specific report ANP-3676P, Surry Fuel-Vendor Independent Small Break LOCA Analysis Licensing Report, in the list of methodologies approved for reference in the Core Operating Limits Report (COLR).

The COLR for North Anna is in TS 5.6.5.b and the COLR for Surry is in TS 6.2.C. The amendments also remove obsolete COLR references that supported use of a fuel product that is not planned for future use.

Enclosure 5 to this letter contains Proprietary information. When separated from Enclosure 5, this document is DECONTROLLED OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

G. Edward Miller, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-338, 50-339, 50-280, and 50-281

Enclosures:

1. Amendment No. 287 to NPF-4

2. Amendment No. 270 to NPF-7

3. Amendment No. 303 to DPR-32

4. Amendment No. 303 to DPR-37

5. Proprietary Safety Evaluation

6. Non-Proprietary Safety Evaluation cc: Listserv OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-338 NORTH ANNA POWER STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 287 Renewed License No. NPF-4

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

A. The application for amendment by Virginia Electric and Power Company et al.,

(the licensee) dated July 12, 2018, as supplemented by letters dated July 9, 2019, May 28, 2020, and October 22, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as 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 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION

2. Accordingly, the license is amended by changes to paragraph 2.C (2) of Renewed Facility Operating License No. NPF-4, as indicated in the attachment to this license amendment, and is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A, as revised through Amendment No. 287, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

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

Attachment:

Changes to Renewed Facility Operating License No. NPF-4 and Technical Specifications Date of Issuance: March 19, 2021 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-339 NORTH ANNA POWER STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 270 Renewed License No. NPF-7

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

A. The application for amendment by Virginia Electric and Power Company et al.,

(the licensee) dated July 12, 2018, as supplemented by letters dated July 9, 2019, May 28, 2020, and October 22, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as 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 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION

2. Accordingly, the license is amended by changes to paragraph 2.C (2) of Renewed Facility Operating License No. NPF-7, as indicated in the attachment to this license amendment, and is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 270, are hereby incorporated in the renewed license.

The licensee shall operate the facility in accordance with the Technical Specifications.

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

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

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

Attachment:

Changes to Renewed Facility Operating License No. NPF-7 and Technical Specifications Date of Issuance: March 19, 2021 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION ATTACHMENT TO NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2 LICENSE AMENDMENT NO. 287 RENEWED FACILITY OPERATING LICENSE NO. NPF-4 DOCKET NO. 50-338 AND LICENSE AMENDMENT NO. 270 RENEWED FACILITY OPERATING LICENSE NO. NPF-7 DOCKET NO. 50-339 Replace the following pages of the Renewed Facility Operating Licenses with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert NPF-4, page 3 NPF-4, page 3 NPF-7, page 3 NPF-7, page 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 5.6-3 5.6-3 5.6-4 5.6-4 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 303 Renewed License No. DPR-32

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

A. The application for amendment by Virginia Electric and Power Company (the licensee) dated July 31, 2018, as supplemented by letters dated July 9, 2019, May 28, 2020, and October 22, 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 3 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION

2. Accordingly, the license is amended by changes to the Technical Specification as indicated in the attachment to this license amendment, and paragraph 3.B 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. 303 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: March 19, 2021 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 303 Renewed License No. DPR-37

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

A. The application for amendment by Virginia Electric and Power Company (the licensee) dated July 31, 2018, as supplemented by letters dated July 9, 2019, May 28, 2020, and October 22, 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 4 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION

2. Accordingly, the license is amended by changes to the Technical Specification as indicated in the attachment to this license amendment, and paragraph 3.B 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. 303 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: March 19, 2021 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION ATTACHMENT TO SURRY POWER STATION, UNITS NOS. 1 AND 2 LICENSE AMENDMENT NO. 303 RENEWED FACILITY OPERATING LICENSE NO. DPR-32 DOCKET NO. 50-280 AND LICENSE AMENDMENT NO. 303 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.

Remove Pages Insert Pages DPR-32, page 3 DPR-32, page 3 DPR-37, page 3 DPR-37, page 3 TS TS 6.2-2 6.2-2 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

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. EMF-2328(P)(A), PWR Small Break LOCA Evaluation Model S-RELAP5 Based, as supplemented by ANP-3676P, Surry Fuel-Vendor Independent Small Break LOCA Analysis, as approved by NRC Safety Evaluation Report dated March 19, 2021.

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

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

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

(Westinghouse Proprietary)

7. 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

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

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

Amendment Nos. 303, 303

OFFICIAL USE ONLY PROPRIETARY INFORMATION SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 287 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-4 AMENDMENT NO. 270 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-7 AMENDMENT NO. 303 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-32 AMENDMENT NO. 303 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2 SURRY POWER STATION UNIT NOS. 1 AND 2 DOCKET NOS. 50-280, 50-281, 50-338, AND 50-339 Information Sensitivity:

The following document contains proprietary information. The proprietary information is withheld from public disclosure pursuant to 10 CFR 2.390. Proprietary information within the document is marked between double square brackets.

((This sentence is an example of the proprietary designation.))

1.0 INTRODUCTION

By applications dated July 12, 2018, for North Anna Power Station, Units 1 and 2 (North Anna or NAPS) (Reference 1) and July 31, 2018, for Surry Power Station, Units 1 and 2 (Surry or SPS)

(Reference 2), respectively, Virginia Electric and Power Company (Dominion Energy Virginia) requested changes to technical specifications (TSs) governing the analytical methods that may be used to determine core operating limits for each plant:

For North Anna, the proposed change would add the Framatome small-break loss-of-coolant accident (SBLOCA) evaluation model described in EMF-2328(P)(A), PWR

[pressurized water reactor] Small Break LOCA Evaluation Model S-RELAP5 Based (Reference 3) and (Reference 4), as supplemented by the NAPS plant-specific methodology report ANP-3467P, Revision 0, North Anna Fuel-Vendor Independent Enclosure 6 OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Small Break LOCA Analysis Licensing Report, (Reference 5) to the list of analytical methods used to determine core operating limits in North Anna TS 5.6.5.b.

For Surry, the proposed change would add the Framatome small-break LOCA evaluation model described in EMF-2328(P)(A) (References 3 and 4), as supplemented by the SPS plant-specific methodology report ANP-3676P, Surry Fuel-Vendor Independent Small Break LOCA Analysis Licensing Report, (Reference 6) to the list of analytical methods used to determine core operating limits in Surry TS 6.2.C.

Considering the similar technical content and proximate submittal dates, to promote efficiency, the U. S. Nuclear Regulatory Commission (NRC) staff opted to address both license amendment requests in the present safety evaluation.

The NRC staff required additional information to complete its review of the requested license amendments, and transmitted 22 requests for additional information (RAIs) to the licensee by letter dated February 8, 2019 (Reference 7). The licensee responded to the RAIs by letter dated July 9, 2019 (Reference 8). On October 21, 2019 (Reference 9), the NRC staff issued a second round of 10 RAIs to address topics that had not been resolved by the licensees initial RAI response (Reference 8). The licensee responded to the second round of RAIs in a submittal dated May 28, 2020 (Reference 10). The licensee further submitted additional supplementary information supporting the proposed TS changes on October 22, 2020 (Reference 11). The supplements provided additional information and clarified the application but did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on October 9, 2018 (83 FR 50697) for Surry and December 4, 2018 (83 FR 62614) for North Anna.

2.0 REGULATORY EVALUATION

2.1 Background

The current SBLOCA analysis methodology for North Anna and Surry is the NRC-approved Westinghouse evaluation model (EM) described in WCAP-10054-P-A/WCAP-10081-A, Westinghouse Small Break ECCS [emergency core cooling system] Evaluation Model Using the NOTRUMP Code (Reference 12), and WCAP-10079-P-A / WCAP-10080-A, NOTRUMP, A Nodal Transient Small Break and General Network Code (Reference 13). The license amendments for North Anna and Surry would replace this Westinghouse evaluation model with a fuel-vendor independent (FVI) SBLOCA methodology composed of Framatomes generically approved EMF-2328 methodology (References 3 and 4), as supplemented for each plant by a plant-specific methodology report included in the submittal.

The plant-specific methodology reports, ANP-3467P (for North Anna) and ANP-3676P (for Surry), supplement the generic EMF-2328 methodology by:

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION proposing to extend application of the methodology to fuel assemblies that, irrespective of manufacturer, satisfy a set of performance-based criteria,1 and describing analysis completed for each plant using ((

)) for a 17x17 assembly for North Anna and 15x15 assembly for Surry, and non-fuel-related plant-specific details.

Analysis of LOCA events is generally performed on behalf of a licensee by the vendor supplying fuel to the reactor. Although the system-wide thermal-hydraulics experienced during a small-break LOCA event would not be expected to be influenced substantially by fuel design changes of recent experience, aspects of the fuel rod design affecting thermal-mechanical behavior (e.g.,

cladding swelling and rupture) can significantly influence this event. Because LOCA analyses typically rely upon vendor-proprietary design information, simulation methods, and validation tests and experiments, evaluation models developed by one vendor are not typically approved for, and may not be fully compatible with, applications to fuel designs developed by other vendors.

Hence, the NRC staff's review of the proposed license amendments for North Anna and Surry focused upon the unique fuel-vendor-independent aspect of the proposed methodology.

2.2 Proposed Technical Specification Changes To support implementation of the FVI SBLOCA methodology, the licensee has proposed to revise the Core Operating Limits Report (COLR) section of the TS for North Anna (TS 5.6.5.b) and Surry (TS 6.2.C) to add EMF-2328(P)(A) (References 3 and 4), as supplemented by the associated plant-specific FVI methodology report (i.e., ANP-3467P or ANP-3676P), and as approved by the NRC staff's safety evaluation.2 Additionally, the licensee proposed to delete existing references in these TS to the current Westinghouse SBLOCA methodologies that would no longer be used.

For North Anna, the licensee has proposed to make the following changes to TS 5.6.5.b:

Delete the following references:

3. WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code.

4. WCAP-10079-P-A, NOTRUMP, A Nodal Transient Small Break and General Network Code.

15. BAW-10168P-A, RSG LOCA - BWNT Loss-of-Coolant Accident Evaluation Models for Recirculation Steam Generator Plants, Volume II only (SBLOCA models).

1 As described further below in Section 3.1.2 of this safety evaluation, the licensee narrowed the proposed scope of the license amendment request to cover (1) applicable Framatome fuel designs and (2) the Westinghouse fuel designs in current use at each plant (subject to minor component changes made in accordance with the vendors approved fuel change process).

2 In response to RAI 1, the licensee modified the proposed TS revision in its original submittal to include reference to the approving NRC staff safety evaluation.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Add the following reference:

3. EMF-2328(P)(A), PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, as supplemented by ANP-3467P, North Anna Fuel-Vendor Independent Small Break LOCA Analysis, as approved by NRC Safety Evaluation Report dated March 19, 2021.

For Surry, the licensee has proposed to make the following changes to TS 6.2.C:

Delete the following references:

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)

Add the following reference:

3. EMF-2328(P)(A), PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, as supplemented by ANP-3676P, Surry Fuel-Vendor Independent Small Break LOCA Analysis, as approved by NRC Safety Evaluation Report dated March 19, 2021.

2.3 Applicable Regulatory Requirements For Surry, the NRC issued the construction permits before May 21, 1971; consequently, Surry was 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.

For North Anna, the NRC issued the construction permits before May 21, 1971; consequently, North Anna was 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. The North Anna Updated Final Safety Analysis Report, Chapter 3, discusses the plants conformance to the GDC.

The NRC staff considered the following regulatory requirements during its review of the proposed changes in both the North Anna and SPS application:

10 CFR 50.36, Technical specifications, insofar as it requires technical specifications that include administrative controls relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, insofar as it requires that each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding3 (See (Reference 16) and (Reference 17) must be provided with an ECCS that must be designed so that its calculated cooling performance following a postulated loss-of-coolant accident conforms to the criteria set forth in 10 CFR 50.46(b),

including peak cladding temperature, maximum cladding oxidation, maximum hydrogen generation, and coolable geometry. Note that the 10 CFR 50.46(b)(5) criterion, long-term core cooling, is not addressed in this application. Specific paragraphs of 10 CFR 50.46 include:

o 10 CFR 50.46(a)(1)(i), insofar as it requires the use of an acceptable evaluation model, and requires the calculation of ECCS cooling performance for a number of postulated LOCAs of break sizes, locations and other properties sufficient to provide assurance that the most severe postulated LOCA has been calculated.

o 10 CFR 50.46(a)(1)(ii), insofar as it provides that an evaluation model may conform to the required and acceptable features of ECCS evaluation models provided in Appendix K to 10 CFR 50.

o 10 CFR 50.46(c)(2), insofar as it defines an ECCS evaluation model.

o 10 CFR 50.46(a)(3), insofar as it establishes reporting requirements for changes to, and errors in, ECCS evaluation models and in the applications thereof, sets forth a threshold for significant changes, and requires more frequent reporting of the estimated effects of significant changes.

10 CFR 50.59, Changes, Tests, and Experiments, insofar as it allows licensees to make changes to an NRC-licensed facility without obtaining a license amendment, provided a change to the TS is not required and none of the criteria contained in 10 CFR 50.59(c)(2) are met.

10 CFR 50, Appendix A, General Design Criterion 35, Emergency core cooling, insofar as it requires a system to provide abundant emergency core cooling to transfer heat from the reactor core following any LOCA at a rate such that fuel and clad damage that could interfere with continued effective core cooling is prevented and clad metal-water reaction is limited to negligible amounts.

10 CFR 50, Appendix K, insofar as it conservatively establishes required and acceptable features for ECCS evaluation models, as an alternative to realistic models that explicitly account for uncertainty, consistent with 10 CFR 50.46.

3 Note that the Optimized ZIRLO-clad fuel in current use at both North Anna and SPS does not satisfy the applicability conditions for cladding material stated in 10 CFR 50.46. However, via exemption, the NRC staff has approved the licensees application of the requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50 to both plants (References 16,and 17).

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION 2.4 Applicable Regulatory Guidance In addition to the requirements identified above, the staff used NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [light-water reactor] Edition, Chapter 15.6.5, Loss-of-Coolant Resulting from Spectrum of Postulated Piping Breaks within the Reactor Coolant Pressure Boundary (Reference 18) as guidance for reviewing the application.

Regulatory Guide (RG) 1.203, Transient and Accident Analysis Methods, (Reference 19) describes the evaluation model development and assessment process (EMDAP), which is one means the NRC staff considers acceptable for use in developing and assessing evaluation models for analyzing reactor transient and accident behavior.

NRC Generic Letter (GL) 88-16, Removal of Cycle-Specific Parameter Limits from Technical Specifications, (Reference 20) specifies actions required to adopt an alternative to submitting cycle-specific parameter limits, provided that these changes are determined using an NRC-approved methodology and are consistent with all applicable limits in the plant safety analysis.

3.0 TECHNICAL EVALUATION

As described in the following sections of this safety evaluation, the NRC staff approached the review by:

evaluating the acceptability of unique aspects of the FVI SBLOCA methodology that represent modifications to Framatomes generically approved EMF-2328(P)(A) SBLOCA Evaluation Model for PWRs (Section 3.1),

evaluating implementation of the FVI SBLOCA methodology to ensure appropriate adherence to limitations and modeling requirements associated with the EMF-2328(P)(A) base methodology (Section 3.2),

evaluating implementation of the FVI SBLOCA methodology to ensure selection of appropriate plant parameters and initial conditions (Section 3.3), and evaluating implementation of the FVI SBLOCA methodology to ensure that the calculated results conform to expectations and that the resulting figures of merit comply with applicable acceptance criteria specified in 10 CFR 50.46(b) (Section 3.4).

3.1 Unique Aspects Of FVI SBLOCA Methodology 3.1.1 Description of Proposed Plant-Specific Methodology The licensee has proposed the adoption of a methodology for analyzing the SBLOCA event that is based upon the generically approved Framatome methodology described in EMF-2328P (References 3, and 4). Aspects of the FVI SBLOCA methodology that remain consistent with the EMF-2328 SBLOCA evaluation model have been reviewed previously on a generic basis by the NRC staff and are not the focus of the present review. Rather, the NRC staff focused its present review upon two new and unique aspects of the plant-specific FVI SBLOCA methodology:

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The licensee originally requested to broaden the application scope of its plant-specific variant of the EMF-2328P methodology to include any fuel designs meeting a set of performance-based criteria. Since fuel designs not manufactured by Framatome could satisfy these criteria, the licensee deemed it a fuel-vendor-independent approach.

However, the licensee's response to RAI 12.b.S.1 narrowed the proposed application scope to cover applicable Framatome fuel designs and the Westinghouse fuel designs in current use at each plant. The NRC staff's review of the licensee's proposed FVI application scope is discussed below in Section 3.1.2.

The licensee has proposed that its FVI SBLOCA methodology incorporates a modification to the EMF-2328P base methodology that ((

)). The NRC staff's review of this modification to the EMF-2328P base methodology is discussed below in Section 3.1.3.

3.1.2 Expanded FVI Application Scope The FVI aspect of the proposed methodology is described in Section 3.2.2 of ANP-3467P and ANP-3676P, as supplemented by the licensees responses to RAIs 11-15 and 11.S1, 12.a.S1, 12.b.S1 and 13.S1. Based on its review of this information, the NRC staff noted the process to include the following:

((

)) the assessment described in Item 1 is reviewed in this safety evaluation. Since reference to this safety evaluation will be included in the COLR references section of the TS for North Anna and 4

In addition to new Framatome fuel assemblies, this applicability evaluation would also apply to permissible modifications to the current fuel design by Westinghouse in accordance with its NRC-approved fuel design change process, as may be implemented at either station in accordance with applicable requirements, including 10 CFR 50.59, Changes, tests, experiments, and as applicable, 10 CFR 50.90.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Surry, the Item 1 assessment is expected to remain static, absent future revisions to the FVI SBLOCA methodology that are submitted for NRC staff review.

Based on the licensees response to RAI 12.a.S1, the NRC staff expects that the ((

)). The licensee also stated:

Given that the typical cycle-specific variations are limited for a normal reload design, fuel products and core operational plans do not change in a manner that impacts the approved LOCA analyses. Assessment of the North Anna and Surry FVI-SBLOCA analyses shall be performed on a reload basis as part of the Dominion Energy process [i.e., in accordance with NRC-approved topical report VEP-FRD-42-A, Reload Nuclear Design Methodology].

This safety evaluation reviewed the analyses to ensure compliance with applicable regulatory requirements. In accordance with 10 CFR 50.46, the effect of any changes or errors that affect the calculated peak cladding temperature in the existing analyses must be estimated and reported to the NRC. However, the licensee may also opt to reperform analysis using the FVI SBLOCA methodology, as needed or desired, following the NRC staff's approval of this license amendment to add the topical reports describing the FVI SBLOCA methodology into the COLR references section of the TS.

((

)). This safety evaluation reviewed the process and criteria the licensee has proposed to perform these applicability evaluations. In general, ((

)) are typically performed by licensees in accordance with Generic Letter 88-16, without prior NRC staff review.

Licensees retain responsibility for ensuring that their actions comply with requirements in 10 CFR 50.59 concerning changes, tests, and experiments. Furthermore, cycle-specific design changes implemented by licensees may invoke regulatory reporting requirements in 10 CFR 50.46(a)(3).

The NRC staffs safety evaluation also reviewed the applicability of the FVI SBLOCA methodology to the fuel designs currently resident at North Anna (17x17 RFA-2 fuel with Optimized ZIRLO cladding) and Surry (15x15 Upgrade fuel with Optimized ZIRLO).

3.1.2.1 Phenomena Identification and Ranking Table (PIRT)

As originally proposed, the licensee ((

)). While the licensee's response to RAI 12.b.S.1 restricted the fuel designs substantially within the proposed application scope, evaluation of these key phenomena remains relevant to assessing applicability of the FVI SBLOCA methodology to the Westinghouse fuel designs currently in use at each plant.

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((

)).

Based on the licensee response to RAI 11, ((

)).

Based on the above, three potentially important fuel-related phenomena were identified for further review, as discussed below.

(( ))

The response to RAI 11 states that ((

)). However, the PIRT table provided for fuel rod performance in NUREG/

CR-6744, Phenomenon Identification and Ranking Tables (PIRTs) for Loss-of-Coolant Accidents in Pressurized and Boiling Water Reactors Containing High Burnup Fuel, (Reference

21) indicates that gap heat resistance, gap size, and gas pressure and composition are all ranked of high importance. (( )) correlates to the initial fill pressure of the fuel rod, fuel rod temperature profile, and production of gaseous fission products as a function of burnup. The plenum gas composition is a function of the initial composition of the fuel in terms of integral burnable absorber loading and fissile material composition and enrichment. These items may be constrained by either sites TS design features, safety limits, limiting conditions for operation, etc., as well as the documented applicability basis of the generically approved EMF-2328 evaluation model.

(( ))

The response to RAI 11 notes that ((

)). This phenomenon is understood to be dependent on both coolant and cladding material properties, specifically alloy material and surface roughness. In a supplement dated May 28, 2020, the licensee ((

)), because while important following loop-seal clearing, the ((

)). The licensee also stated, in the supplemental response, that following loop-seal clearing, the core liquid level recovers rapidly to about the level of the cold leg. Since the licensee clarified that the phenomenon has overall ((

)), the NRC staff finds that ((

)) and, as such, determined that further review of the phenomenon and its analytic treatment was not necessary.

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(( ))

The NRC staff has observed a tendency for larger small breaks to be potentially limiting for some PWRs. As such, the NRC staff questioned the importance of ((

)).

((

)).

The NRC staff identified that the ((

)).

((

)).

The licensee provided additional analytic support for its position in the supplemental response to RAI 16, dated May 28, 2020 (Reference 10). In its supplemental response, the licensee OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION evaluated ((

)).

The NRC staff reviewed the supplemental information provided by the licensee and noted that

((

)) The NRC staff also noted that the analytic sensitivity study demonstrated that, even though the conditions of the specific transients, (( )), suggested the prior results for other plants may have questionable applicability to the present analyses, ((

)) Based on its review of the prior studies as supplemented with the analytic sensitivity studies, the NRC staff determined that the modeling approach described in EMF-2328(P)(A) remains acceptable for the Surry and NorthAnna plant-specific applications ((

)). The NRC staff conclusion is based on the present transient timings and PCT magnitudes.

3.1.2.2 Fuel Design-Specific Applicability of FVI Approach The NRC staffs review considered whether the licensee has proposed adequate constraints to ensure application of the FVI SBLOCA methodology only to fuel designs for which its physical models have been demonstrated to be applicable. In particular, in RAI 12.b.S1, the NRC staff questioned whether past NRC staff reviews of the Framatome analysis Codes supporting the proposed FVI SBLOCA methodology (e.g., RODEX2 and S-RELAP5) have demonstrated sufficient capabilities to model the complete range of fuel designs to which the proposed FVI SBLOCA methodology may be applied.

In its response to RAI 12.b.S1, the licensee narrowed the scope substantially for the fuel assembly designs to which the proposed FVI SBLOCA methodology may be applied. The response indicated that the proposed method (including the submitted analyses) would apply to the currently loaded fuel at both plants, with allowance for minor fuel component changes made by the vendor in accordance with its NRC-approved change process. The response also indicated that the proposed method would also be applicable to Framatome fuel.

Previous reviews by the NRC staff (Reference 4) have found application of the evaluation model described in EMF-2328(P)(A) to be acceptable for current Framatome-designed fuel assemblies. Framatome has existing processes and practices, subject to NRC oversight, for assessing the continued applicability of its analysis methods to its own new fuel designs.

Therefore, the NRC staff did not review such applications on a plant-specific basis for either station. However, the applicability of the evaluation model (EM) to other vendors fuel designs is beyond the generic approval scope of EMF-2328(P)(A). Therefore, the NRC staff evaluated whether the FVI SBLOCA approach proposed by the licensee may be applied to the current fuel designs loaded at each plant, namely the Westinghouse RFA-2, or Upgrade, fuel assembly design with ZIRLO or Optimized ZIRLO cladding.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION In response to RAI 14, the licensee discussed specific considerations concerning when an updated analysis or evaluation would require prior NRC staff review and approval. The NRC staff agrees that Dominion must estimate the effect of changes to, or errors in, both the FVI evaluation model and its application, in conformance with 10 CFR 50.46(a)(3) requirements.

However, the NRC staff also considers changes to the fuel system design to fall within the definition of change provided in 10 CFR 50.59(a)(1), and hence, the NRC staff determined that 10 CFR 50.59 requirements apply to changes in the fuel system design. Thus, if changes to the fuel system design are made, the criteria provided in Paragraph (c) of 10 CFR 50.59 would need to be considered, to address whether the effects of any changes to the FVI EM or its application are appropriately treated and, if determined from the screening, submit an amendment. In Regulatory Issue Summary 2016-04, Clarification of 10 CFR 50.46 Reporting Requirements and Recent Issues with Related Guidance not Approved for Use, (Reference 25) the NRC staff provides discussion on the inter-relationship between the two requirements, stating:

Based on the nature and the effect of a particular change to an evaluation model, both 10 CFR 50.46 and 10 CFR 50.59 should be independently applied on their own merits. Specifically, 10 CFR 50.59(c)(4) states:

The provisions in this section do not apply to changes to the facility or procedures when the applicable regulations establish more specific criteria for accomplishing such changes.

Since 10 CFR 50.46 establishes more specific criteria for estimating the effects of and reporting changes to, or errors in, evaluation models, or in the applications thereof, 10 CFR 50.59 reporting requirements do not apply to changes made to evaluation models. Note, however, that facility changes must be screened and evaluated according to 10 CFR 50.59 criteria, while associated changes to, or errors in, the evaluation model, or in applications thereof, that affect the PCT calculation, are subject to 10 CFR 50.46 reporting requirements.

Applicability Assessment for Current Westinghouse Fuel Designs The reload fuel design currently loaded at North Anna is Westinghouse 17x17 RFA-2 fuel with Optimized ZIRLO cladding. The reload fuel design currently loaded at Surry is Westinghouse 15x15 Upgrade fuel with Optimized ZIRLO cladding. The Optimized ZIRLO cladding material used in the fuel assemblies currently loaded at North Anna and Surry is beyond the scope of the existing approvals for the Framatome Codes supporting the FVI SBLOCA methodology, as are the zirconium diboride burnable absorbers that may be incorporated into these fuel assemblies.

Acceptable Cladding Materials

((

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))

In response to RAI 18S1, ((

)).

Fuel Pellet Composition The current resident fuel in use at Surry and North Anna may differ slightly in pellet composition from the NRC-approved Framatome fuel designs. Specifically, Surry and North Anna use fuel with Westinghouse Integral Fuel Burnable Absorber (IFBA). An evaluation of this design is provided in Westinghouse Topical Report WCAP-10444-P-A, Westinghouse Reference Core Report: VANTAGE 5 Fuel Assembly (Reference 26). IFBA fuel rods contain regions of pellets that have an annular blanket of neutron-absorbing material. They differ from conventional fuel rods in that:

An additional gas constituent is included in the IFBA rod. This additional gas is the helium that is created and is released from the depletion of the [neutron-absorbing material] contained in the burnable absorber coating.

Based on an NRC staff review of the NRC-approved Framatome fuel performance modeling tools (i.e., RODEX2-2A), it was not clear that the RODEX Codes have the ability to model the performance of IFBA fuels. ((

)).

((

)).

In the response to RAI 15, the licensee stated that ((

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))

Based on the information provided by the licensee, the NRC staff determined that the burnup dependent fuel rod characteristics associated with burnable absorber content, including IFBA, will be adequately addressed because ((

)).

((

)).

The licensee provided this information in the response to RAI 19.

The response to RAI 19 provided several sensitivity studies. ((

))

Acceptable Fuel Assembly Geometry In its responses to RAIs 12.b and 17 (Reference 8), the licensee had proposed to apply the FVI SBLOCA methodology to all fuel designs satisfying a specific set of criteria, which the NRC staff summarized in second-round RAI 12.b.S1 (Reference 9). The NRC staff questioned the proposed criteria, in part, because they did not appear to ensure application of the FVI SBLOCA methodology would apply only to fuel assemblies that are geometrically similar to those considered in the curent review. The NRC staff's question was based on the observation that, while the small-break LOCA accident progression is influenced significantly by overall system behavior, phenomena associated with fuel rod design and performance can also affect the predicted figures of merit sunstantially.

In its response to RAI 12.b.S1 (Reference 10), the licensee revised its request by proposing to limit the scope of the FVI SBLOCA methodology to applicable Framatome fuel designs and the current Westinghouse fuel design in use at each plant (subject to minor component changes).

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The more restrictive criteria the licensee proposed in its response to RAI 12.b.S1 provides reasonable assurance that the FVI SBLOCA methodology will be applied only to fuel designs with an appropriate geometry.

3.1.2.3 Cycle-Specific Verification of FVI Analyses Reactor licensees typically do not perform a unique LOCA analysis for each operating cycle.

Rather, a licensee typically analyzes to bounding inputs meant to envelop anticipated operating parameter ranges for a number of future operating cycles. Prior to the start of each such operating cycle, in accordance with 10 CFR 50.59, the licensee would either (1) ensure that the existing LOCA analysis continues to bound the planned operating cycle or (2) reanalyze the LOCA event with a new set of plant parameters and inputs sufficient to bound the planned operating cycle.

One reason a licensee may reanalyze the postulated LOCA or other design-basis events is to support planned fuel cycles that involve loading new fuel assemblies of a different design relative to the existing resident fuel burned in past cycles (e.g., as part of a fuel transition).

Reanalysis is especially likely when the new fuel assembly is supplied by a different vendor with different proprietary analysis methods (i.e., mixed cores).

In Section 3.2.2 of ANP-3467P and ANP-3676P, the licensee stated that the analyses contained therein are applicable to 17x17 (for North Anna) and 15x15 (for Surry) fuel products with ZIRLO and Optimized ZIRLO. The foregoing discussion evaluates the applicability of the methods supporting the proposed FVI SBLOCA methodology to the current Westinghouse fuel designs used at each plant, which contain specific design features that differ substantially from those of the Framatome fuel for which the proposed analysis methods were originally developed. The NRC staff expressed concern with the extent to which specific analyses performed using the FVI SBLOCA methodology (e.g., the analyses presented in ANP-3467P and ANP-3676P) may be used to represent different fuel designs to which the FVI SBLOCA methodology has been determined to be applicable.

For instance, Section 3.2.2 of the plant-specific methodology reports appears to acknowledge the importance of modeling the fuel in the reactor core with representative detail: ((

)) These reports further state that ((

)) However, in the event that the analyses contained in ANP-3467P and ANP-3676P were applied to fuel designs different than the existing resident assemblies, the representativeness of the input parameters cited in these plant-specific methodology reports may not be maintained.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION The NRC staffs review of the licensees submittal did not identify specific criteria the licensee would use to verify the continued applicability of a given analysis performed in accordance with the FVI SBLOCA methodology on a cycle-specific basis. ((

)). In RAI 12.a.S.1, the NRC staff requested that the licensee provide further justification for these proposed criteria.

In response to RAI 12.a.S.1 (Reference 10), the licensee proposed a revised discussion of its cycle-specific verification criteria that are based upon NRC-approved topical report VEP-FRD-42-A (Reference 27). This RAI response references a subsequent revision of VEP-FRD-42-A with respect to assessing cycle-specific variations, which the licensee defined as parameters that can vary during a typical core reload:

uranium enrichment fuel loading pattern core peaking factors burnup power history design of minor fuel components, in accordance with NRC-approved change process The licensee also identified a set of factors that would not be expected to vary during a typical core reload:

fuel assembly design, beyond minor component changes peaking factor limits nominal cycle length burnable absorber (integral or discrete)

In addition, the licensee stated that the (( )) cited in the licensee's response to RAI 12.a (noted above) would also be incorporated into reload checks associated with the FVI SBLOCA methodology. The licensee stated that changes to fuel assembly design and core operational plans in excess of typical cycle-specific variations, such as those listed above, would require additional engineering review to determine impacts to the plant design and licensing bases.

In its response to RAI 12.a.S.1, the licensee stated that the cycle-specific verification process for the FVI SBLOCA methodology would follow its approved topical report governing the core reload analysis process. As summarized above, the licensee's approved cycle-specific criteria OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION for verifying continued analysis applicability are based upon the assumption that the basic fuel design remains the same. This position is consistent with that reflected in the NRC staff's safety evaluation on Dominions topical report VEP-FRD-42-A (Reference 27), which indicates that key fuel features will be represented by detailed inputs in core design and safety analysis models.

The NRC staff considers this treatment appropriate, since changes in fuel design may, in general, significantly influence the figures of merit calculated by the FVI SBLOCA methodology.

3.1.2.4 Influence of Fuel Design on Evaluation Model Predictions Initially, the NRC staffs review had intended to assess whether the licensee had adequately validated the FVI SBLOCA methodology for the broad spectrum of possible fuel designs satisfying the licensee-proposed criteria summarized by the NRC staff in RAI 12.b.S.1.

However, the licensee's response to RAI 12.b.S.1 simplified this portion of the NRC staff's review by restricting the scope of the proposed FVI SBLOCA methodology only to applicable Framatome fuel designs and the current Westinghouse fuel design at each plant (subject to minor changes to fuel components). Applicability to the current Westinghouse fuel design for each plant has been specifically evaluated above in this safety evaluation and is, based on the above, acceptable.

Because the influence of fuel assembly design remains topical to the assessment of cycle-specific analysis applicability, the NRC staff has documented relevant portions of its review in its safety evaluation. In particular, several responses to requests for additional information, remain relevant to assessing cycle-specific analysis applicability:

In RAI 20, the NRC staff requested that the licensee perform a real-world demonstration of the FVI SBLOCA methodology by explicitly analyzing a switch between two actual fuel designs that fall within the scope of the methodology, while holding all other input parameters constant.

In RAI 21, the NRC staff requested that the licensee demonstrate the capability of the FVI SBLOCA methodology to model fuel designs clad with zirconium-based alloys by performing a sensitivity analysis that examines the impact of toggling the fuel cladding material between M5 and Zircaloy-4 for a 3-inch and a 6.5-inch break.

In RAI 22, the NRC staff requested that the licensee perform a sensitivity analysis using the FVI SBLOCA methodology to assess whether conceivable variations in fuel-related parameters deemed insignificant in the methodology could significantly influence calculated figures of merit for the SBLOCA event.

The licensees responses to RAIs 20 and 22 did not did not provide the expected completion of the requested sensitivity studies. However, in RAI 12.b.S.1, the licensee discussed the results of a different sensitivity study, somewhat related to the earlier requests posed by the NRC staff.

((

OFFICIAL USE ONLY PROPRIETARY INFORMATION

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)). Therefore, the NRC staff did not consider the issue of fuel design influence on cycle specific analysis applicability to have been resolved by the information provided in response to RAIs 20 and 22; however, prior to making an overall regulatory conclusion on this issue, as described in the following paragraph, the NRC staff considered additional evidence related to this issue in response to RAI 21.

((

))

Figure 1: Licensee-Calculated Sensitivity Results from Response to RAI 12.b.S.1 In response to RAI 21, the licensee stated that a similar sensitivity study concerning the impact of cladding material on swelling and rupture had already been performed for Millstone Power Station, Unit 2 (a PWR designed by Combustion Engineering). The licensee stated that,

((

)). The licensee stated that the calculated difference in peak cladding temperature for the Millstone 2 sensitivity study was small (i.e., 4 °F). Although the licensees sensitivity study may have been appropriate for assessing impacts on the limiting break for Millstone 2, the NRC staff observed significant variation in the calculated peak cladding temperature at other break sizes. This behavior is illustrated below in Figure 2, which plots the results of the licensees calculations for Millstone 2. Moreover, given that the limiting break sizes for both North Anna and Surry are outside the range of the sensitivity study results the licensee provided for Millstone 2, the implication of the Millstone 2 sensitivity results to the present review is not clear. Based on secondary NRC staff questions, the licensee provided further supplemental information.

The licensee's RAI responses reflect its view that the small-break LOCA event is governed primarily by system-wide thermal-hydraulic behavior, and that fuel-related parameters have only secondary influence. While the design of the reactor coolant system and ECCS play a major role in the small-break LOCA event response, fuel rod behavior is also significant. In particular, OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION the influence of fuel rod behavior is highly non-linear and temperature dependent. ((

)). As such, the licensee's responses to RAIs 12.b.S.1, 20, 21, and 22, did not independently support its view on the importance of any particular fuel-related parameters.

However, the NRC staff determined that the empirical evidence presented in the licensee response is consistent with the conclusion above concerning the adequacy of its existing core reload criteria for verifying continued analysis applicability, which are predicated upon the condition that the fuel design remains essentially the same.

((

))

Figure 2: Results of Dominions Cladding Material Sensitivity Analysis for Millstone, Unit 2 3.1.3 Additional Proposed Modification to EMF-2328 Base Methodology Beyond the proposed FVI application scope, the licensees submittals identified one additional deviation from the approved EMF-2328 base methodology, in Section 3.2.3 of ANP-3467 for North Anna and ANP-3676 for Surry. Subsequent to the NRC staffs approval of the EMF-2328 evaluation model, Framatome implemented a ((

)) .

((

5 As described in response to RAI 8, Framatomes ((

)).

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)).

In RAI 8, the NRC staff requested additional information concerning this change. In particular, the NRC staff requested that the licensee describe the discovery of the issue, explain ((

)).

The licensees response to RAI 8 ((

)).

Both conservative requirements in Appendix K to 10 CFR 50 and more recent guidance in Regulatory Guide 1.203 (Reference 19) reflect the importance of performing comparisons of evaluation model predictions against relevant test data. The NRC staff issued RAI 8 S1 to request additional information necessary to confirm whether (1) the existing EMF-2328 evaluation model assessment remains valid, or (2) a new assessment is necessary with the modified evaluation model the licensee proposes to apply to North Anna and Surry.

In its response to RAI 8 S1, the licensee provided additional information concerning the impact of Framatomes modification on the validation and qualification of its approved SBLOCA evaluation model. ((

)) issue does not affect the validation of the approved evaluation model. The licensees response ((

)).

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The NRC staffs review concluded that adequate evidence was presented by the licensees responses to RAIs 8 and 8 S1 to demonstrate that the continued applicability of the validation of Framatomes SBLOCA evaluation model was not adversely impacted by the modification Framatome implemented ((

)). Further, based on substantial design similarities between Surry and North Anna that result in a similar SBLOCA response across the break spectrum (e.g., see Figure 3 and Figure 4, below), the NRC staff has reasonable assurance of similar qualitative behavior for North Anna in the same break range.

3.1.4 Acceptability of Modified Evaluation Model The NRC staff has concluded that the proposed, plant-specific, fuel vendor-independent variant of EMF-2328(P)(A) is acceptable for implementation at Surry Units 1 and 2 and North Anna Units 1 and 2, based on the foregoing determination that the EM, as modified, complies with the requirements of 10 CFR 50.46(a)(1)(i), insofar as it is an acceptable evaluation model.

Furthermore, the NRC staff determined that the changes proposed by the licensee remain consistent with the required and acceptable features for ECCS evaluation models specified in Appendix K to 10 CFR 50.

The NRC staffs conclusion applies strictly to the application of the proposed evaluation model to North Anna and Surry. As discussed in this safety evaluation, the NRC staffs conclusion is based upon a number of plant-specific considerations, including:

plant-specific analyses, including a number of sensitivity studies, performed for a specific set of conditions applicable to North Anna and Surry, the existence of robust, NRC-approved licensee reload analysis methods that contain adequate controls upon the representativeness of the modeling of fuel assemblies, the existence of sufficient analytical margins, considering both conservative inputs and predicted margins to regulatory acceptance criteria, and confirmatory analysis performed by the NRC staff using input conditions intended to represent North Anna and Surry.

The NRC staff has not generically reviewed the proposed FVI SBLOCA methodology for application to other reactors; nor has the NRC staff reviewed Framatomes EMF-2328, Supplement 1, evaluation model for generic application to non-Framatome fuel designs.

3.2 Adherence to Limitations on EMF-2328(P)(A) Base Methodology 3.2.1 EMF-2328(P)(A), Revision 0 EMF-2328(P)(A), Revision 0 (Reference 3) contains one limitation:

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OFFICIAL USE ONLY PROPRIETARY INFORMATION While it has been shown in [NUREG/CR-4945, Summary of the Semiscale Program (1965-1986) [(Reference 28)] that the thermal-hydraulic phenomena observed for breaks up to 10 percent of the cold leg flow area are the same, if the code is used for break sizes larger than 10 percent of the cold leg flow area additional assessment must be performed to ensure that the code is predicting the important phenomena which may occur.

Both North Anna and Surry use the EMF-2328(P)(A) methodology for breaks less than 10 percent of the cold leg flow area. Therefore, the staff determined that it is not necessary for the licensee to assess breaks larger than 10 percent of the cold leg flow area, and the licensee meets this limitation.

3.2.2 EMF-2328(P)(A), Supplement 1 The NRC staff reviewed the North Anna and Surry applications to ensure compliance with EMF-2328(P)(A), Supplement 1 (Reference 4), which, in conjunction with Section 4.0 of the NRC staff's associated safety evaluation, provides explicit modeling requirements in the following eight areas:

1. Spectrum of break sizes

2. Core bypass flow paths in the reactor vessel

3. Reactivity feedback

4. Delayed reactor coolant pump trip

5. Maximum accumulator/refueling water storage tank temperature

6. Loop seal clearing and crossover leg modeling

7. Breaks in attached piping

8. Core nodalization The following subsections contain the NRC staffs review of the EMF-2328(P)(A), Supplement 1 (Reference 4) methodology requirements and sensitivity studies.

3.2.2.1 Spectrum of Break Sizes In accordance with 10 CFR 50.46(a)(1)(i), ECCS cooling performance must be calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated. Section 2.1 of EMF-2328(P)(A), Supplement 1, identifies that Framatomes approved small-break LOCA methodology is intended to cover a range of breaks up to, and including, 10 percent of the cold leg cross-sectional area. In typical applications of the EMF-2328 methodology, the remainder of the break spectrum is analyzed by Framatomes large-break LOCA evaluation model documented in EMF-2103(P)(A), Realistic Large Break LOCA Methodology for Pressurized Water Reactors, (Reference 29). However, neither North Anna nor Surry currently use EMF-2103(P)(A) to analyze the postulated spectrum of large-break LOCAs. Therefore, a discontinuity could arise in the break spectra covered by the EMF-2328(P)(A)-based FVI SBLOCA methodology and the large-break LOCA EMs employed at North Anna and Surry, such that an unanalyzed range of break sizes could exist between the largest break analyzed using the FVI SBLOCA methodology and the smallest break analyzed using each plants respective large break loss-of-coolant accident (LBLOCA) EM.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION Revision 3 (Reference 29), for several sample plant designs that illustrate the general physical behavior described for intermediate breaks in the licensee's responses to RAIs 2 and 2 S1.

In addition to the information presented by the licensee, the NRC staff identified additional information supporting the conclusion that, at the present time, the portion of the break spectrum between the FVI SBLOCA methodology and the existing ASTRUM large-break LOCA evaluation model is not limiting. First, a 2004 analysis performed using Framatomes Realistic Large Break LOCA evaluation model explicitly analyzed down to a minimum break size of approximately 10 percent of the cold leg cross sectional area; the analysis showed that, for the evaluated conditions, the calculated figures of merit were not limiting. Second, while the results of ASTRUM analyses submitted for both North Anna (Figure 1 of Reference 15) and Surry (Figure 1 of Reference 31) did not explicitly analyze breaks in the range of interest to RAI 2, they showed a significant trend of decreasing peak cladding temperatures as the range of interest is approached. Third, as shown subsequently in Table 2, the accumulator line break, which has a slightly larger effective flow area than the largest reactor coolant system break considered in the FVI SBLOCA analysis, is not limiting for North Anna or Surry. Fourth, the results of confirmatory analysis performed by the NRC staff using the TRACE code are consistent with the contention that, for the analyzed existing plant conditions, limiting values for the calculated figures of merit are not expected in the range of interest to RAI 2.

Based on the above, the NRC staff agrees that the licensee has provided sufficient technical basis to conclude that the portion of the break spectrum between the approved applicability ranges of the proposed FVI SBLOCA methodology and the existing ASTRUM large-break LOCA evaluation model is non-limiting for the current configuration of North Anna and Surry.

Although the NRC staff agrees with a significant part of the licensee's discussion concerning the general physical behaviors that occur across the LOCA spectrum, it was not possible to conclude that a certain portion of the LOCA spectrum can never become limiting, irrespective of future changes licensees may implement to plant designs and analysis methods. Therefore, the NRC staff does not find that the portion of the break spectrum between the proposed FVI SBLOCA methodology and its current large-break LOCA evaluation model is inherently non-limiting, and further notes that 10 CFR 50.46 requires acceptable emergency core cooling system performance for all breaks large enough to result in a loss of reactor coolant greater than the capability of the reactor coolant makeup system, up through a break equivalent to a double-ended rupture of the largest reactor coolant system pipe. As such, the conclusion reached by the NRC staff concerning the current configuration of North Anna and Surry is that it in no way relieves the licensee of its regulatory obligation to ensure that 10 CFR 50.46 is satisfied for the entire postulated spectrum of LOCAs, including the range of break sizes of interest to RAI 2. Iffuture analysis of this region of break sizes should become necessary (e.g.,

following future plant design or analysis method changes), and in accordance with 10 CFR 50.46(a)(1)(i), the licensee will first need to define an acceptable evaluation model for this portion of the break spectrum.

The approved EMF-2328 methodology requires a break spectrum ((

)). The analyzed spectrum of breaks for North Anna and Surry considers break diameters from 1.0 inch to 8.7 inches (equivalent to 10 percent of the cold leg area for both plants). The limiting break sizes were the 6.5-inch break for North Anna and the 2.6-inch break for Surry. The NRC staff determined for each plant that an appropriate break-size resolution was ((

)) and that the limiting break was appropriately identified.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION Additionally, the EMF-2328 base methodology requires that the largest small break that depressurizes to a pressure just above the accumulator actuation pressure be included in the break spectrum. ((

)). The NRC staff noted that Table 4-2 of ANP-3676P (Reference 6) ((

)). Therefore, the NRC staff determined that the break size for which accumulator injection would not occur was

(( )).

In RAI 3, the NRC staff questioned how the methodology considers the azimuthal orientation of the break (e.g., top, bottom, or side) with respect to its influence on the liquid/vapor composition at the break exit plane. The state of the fluid at the break plane influences the rates at which RCS pressure declines and RCS fluid mass is lost. In some SBLOCA scenarios, stratification may occur for an extended period in the RCS loops, which results in inhomogeneous fluid conditions at the break plane. Therefore, the azimuthal orientation of the break may in general affect the state of the fluid at the break plane, and, hence, the overall event response.

In response to RAI 3, the licensee stated that the methodology ((

)) but that the Moody break flow model is used in accordance with Appendix K to 10 CFR 50. The Moody model is a homogenous flow correlation, which assumes that the two phases are in thermodynamic equilibrium (Reference 32). The model is generally understood to overestimate the rate of flow discharging from the break. Appendix K to 10 CFR 50 requires consideration of at least three multipliers to the discharge coefficient ranging from 0.6 to 1.0, but in practice, this requirement is generally applied to large breaks (i.e., double-ended guillotine breaks). For split breaks, the postulated range of breaks is typically examined by considering an adequately resolved set of discrete break sizes in the range of concern.6 The NRC staffs review of the licensee's response to RAI 3 determined that (1) the Moody model is not incompatible with the modeling of break orientation and fluid stratification, which simply affect the fluid conditions at the break that would be input to the model and (2) overestimation of the critical flow rate from the break does not necessarily add conservatism for an analysis of the small-break LOCA event.

Regarding the latter point, in the small-break LOCA event, the critical flow rate from the break has multiple, competing effects. While overestimating the critical flow rate tends to lead to an earlier core uncovery, it also tends to result in earlier depressurization of the reactor coolant system, higher ECCS flow rates, and earlier injection of coolant from the accumulators. In judging the net impact of these countervailing influences, it is important to recall that the tendency for overprediction associated with the Moody critical flow model arises predominately in the two-phase break flow regime that is encountered later in the event, after a significant mass of RCS fluid has been lost. Conversely, with respect to the prediction of subcooled critical flow, such as occurs early in the small-break LOCA event, the Moody model tends to provide a more realistic estimate. Hence, use of the Moody model does not influence the predicted times to core uncovery and reflood uniformly; furthermore, the physical insights discussed above indicate that the Moody model does not necessarily add conservatism to a small-break LOCA analysis. The NRC confirmed these physical insights through sensitivity calculations using the 6 For example, Framatomes EMF-2328 small-break LOCA methodology considers a range from ((

)).

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OFFICIAL USE ONLY PROPRIETARY INFORMATION TRACE thermal-hydraulic code including the impact of applying two-phase critical flow multipliers representative of the Moody correlation.

Based on the licensees response to RAI 3, the NRC staff performed additional confirmatory analysis using the TRACE code to assess the impact of varying the azimuthal orientation of the break for the small-break LOCA event under existing plant conditions for North Anna and Surry.

While the azimuthal orientation of a break is expected to be significant, in general, the results of this sensitivity analysis did not identify significant variation in the peak cladding temperature or other figures of merit attributable to the azimuthal orientation of the break for the specific conditions analyzed. As a result, the NRC staff considered the issue of break orientation and fluid stratification at the break location to have been adequately addressed in light of overall modeling conservatisms reviewed in the licensees analysis. Therefore, the NRC staff concluded that the licensee has provided acceptable assurance that the most severe postulated LOCA has been calculated with respect to the underlying phenomena associated with variations in break flow rate, which may be considered to be among the varying sizes, locations, and other properties associated with the postulated spectrum of LOCAs that must be calculated.

Section 2.3 of EMF-2328(P)(A), Supplement 1, also ((

)). To confirm whether the hot leg nodalization of the North Anna and Surry plant models is consistent with the required modeling practice in EMF-2328, Supplement 1, the NRC staff requested additional information in RAI 4.a. The licensees response confirmed that the approved modeling approach specified in EMF-2328, Supplement 1, had been used.

3.2.2.2 Core Bypass Flow Paths in the Reactor Vessel The S-RELAP5 vessel nodalization presented in EMF-2328(P)(A), Supplement 1 (Reference 4),

((

)) will tend to produce a higher PCT for all small-break LOCAs. The NRC staff was unable to verify ((

)) for either the North Anna or Surry plant models. To confirm that the plant models for North Anna and Surry comply with this required modeling practice from EMF-2328, Supplement 1, the NRC staff requested additional information regarding ((

)) in RAI 4b.

The licensees response to RAI 4b (Reference 8) confirmed ((

)). The licensee further stated that, in accordance with EMF-2328, Supplement 1, ((

)) during a small-break LOCA. The NRC staff's review concluded that the clarification provided by the licensee concerning ((

)) reflects an appropriate implementation of the approved EMF-2328, Supplement 1, evaluation model.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.2.2.3 Reactivity Feedback Supplement 1 of EMF-2328(P)(A), discusses how, in addition to the modeling of reactivity feedback from control rod insertion upon reactor trip, the SBLOCA model will incorporate reactivity feedback due to changes in the moderator temperature, core void fraction, and fuel temperature. Excluding negative feedback is conservative when the moderator temperature coefficient (MTC) is negative. However, the MTC can be slightly positive at beginning-of-cycle conditions (e.g., during startup), causing positive reactivity feedback and increasing core power during the depressurization phase of a SBLOCA prior to control rod insertion. As such, the NRC staffs safety evaluation on EMF-2328P, Supplement 1, found that the maximum positive value of the MTC consistent with the plant TS should be incorporated into the analysis.

To confirm that the analysis for North Anna and Surry complies with the approved approach for reactivity feedback modeling, the NRC staff requested additional information, in RAIs 4.d and 4.e, to confirm that relevant reactivity feedback mechanisms are modeled with appropriately conservative values.

In response to RAIs 4.d and 4.e, the licensee confirmed that the moderator reactivity coefficients used in its analyses conservatively model beginning-of-cycle conditions. The licensee stated that the inputs used for the analyses apply the most positive (or least negative) reactivity feedback possible as the moderator density decreases due to voiding in the core during the small-break LOCA event. The licensee identified that the inputs represent an all-rods-out condition, which tends to have a positive influence on the resulting MTC value. The licensee further stated that its reactivity calculations used a conservative value for beta-effective (i.e., the effective fraction of delayed neutrons).

The NRC staffs review found that the licensees combined response to RAIs 4.d and 4.e acceptably addressed reactivity feedback associated with the moderator. The licensee's response did not specifically address modeling of fuel temperature feedback (i.e., Doppler coefficient of reactivity). Sensitivity studies performed by the NRC staff indicate that analyzed existing conditions for North Anna and Surry do not appear to have a strong sensitivity to the fuel temperature feedback. Nevertheless, as noted in the NRC staff's corresponding safety evaluation, EMF-2323PA, Revision 0, Supplement 1PA, Revision 0, requires modeling of fuel temperature feedback at its minimum calculated value, with an allowance for uncertainty. The licensee's proposed adoption of approved topical report EMF-2328, Supplement 1, into the Core Operating Limits Report section of the TS for each plant would require conservative modeling of fuel temperature feedback for analyses performed with the FVI SBLOCA methodology.

3.2.2.4 Delayed Reactor Coolant Pump Trip Consistent with EMF-2328, Supplement 1, the licensees baseline analysis for the small-break LOCA event assumed that the reactor coolant pumps (RCPs) trip on a loss of offsite power concurrent with the reactor trip. Supplement 1 of EMF-2328, also calls for sensitivity analyses to address scenarios where offsite power is assumed to be available, and the RCPs continue running until they are manually tripped by reactor operators in accordance with emergency operating procedures. As such, the licensees submittal contains analysis of a spectrum of hot-and cold-leg breaks assuming the RCPs are manually tripped with a 5-minute delay after a loss of subcooling margin in the RCS hot legs. The licensee stated that the results of its sensitivity analyses assuming a delayed RCP trip proved to be non-limiting by at least 150 °F relative to the baseline analyses for both Surry and North Anna.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION The NRC staffs review of the licensees delayed RCP trip analyses submitted in ANP-3467P and ANP-3676P resulted in two first-round RAIs.

In RAI 5, the NRC staff requested that the licensee provide additional information concerning the results of the delayed RCP trip sensitivity studies for the analyzed spectrum of hot- and cold-leg breaks for North Anna and Surry, such that the NRC staff could evaluate their scope and adequacy.

The licensees response to RAI 5 provided the requested data. The cold-leg break results were found to bound those for the hot leg. The licensees results from the delayed RCP trip sensitivity analysis for each plant for the more limiting cold-leg break spectrum are graphed below in Figure 3 and Figure 4, along with the baseline analysis results. The licensees response to RAI 5 confirmed that the delayed RCP trip sensitivity analyses ((

)).7

((

))

7 This procedure is discussed further below in Section 3.2.2.6.

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((

))

The licensees response to RAI 5 revealed ((

)).

Based on the points below, the NRC staffs review of this issue for Surry concluded that there is no reason to expect that breaks in the range (( )) would tend to experience higher peak cladding temperatures for the delayed RCP trip sensitivity analysis as compared to the baseline analysis.

First, the peak cladding temperatures calculated for Surry for the delayed RCP trip sensitivity study were comparable to or lower than those calculated for the baseline case for both the 2- and 3-inch break scenarios.

Second, Surry and North Anna are similarly designed Westinghouse 3-loop plants; the more rigorous analysis the licensee performed for North Anna in the (( ))

break diameter range shows the delayed RCP trip scenario to be bounded by the baseline analysis case for North Anna. Although, as expected, the numerical values for the calculated peak cladding temperature vary between the two plants, the qualitative behavior shown in Figure 3 and Figure 4 tends to be similar overall.

Third, RCP trip times of approximately 6-8 minutes into the event for the delayed-trip sensitivity cases for breaks in the (( )) diameter range remain a relatively OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION small fraction of the total time to the peak cladding temperature ((

)). Throughout the time prior to the tripping of the RCPs, and for some time continuing beyond the trip, the break location is predicted to remain covered with liquid. Therefore, the mass loss through the break should not be exacerbated by RCP operation. In fact, operation of the RCPs may tend to reduce the mass loss through the break during the early phase of the small-break LOCA event, when the primary system is only slightly voided. Break exit quality under slightly voided conditions may be incrementally larger due to the homogenizing effect of RCP operation, as compared to a stratified, natural circulation flow.

Fourth, the results of sensitivity studies performed by the NRC staff, with the TRACE Code using an input deck for a Westinghouse 3-loop plant, confirmed the staffs physical insights (e.g., see Figure 5 below) and further showed comparable qualitative behavior to the licensees calculations in the (( )) break size range with respect to the limited and typically slightly beneficial impact of an RCP trip on subcooling with a delay time of 5 minutes.

Figure 5: Dependence of Integrated Break Flow on RCP Trip Timing, NRC Staff Confirmatory Calculation Based on the above, which involve analysis completed by the licensee for Surry and North Anna, along with the NRC staff's physical insights and independent confirmatory calculations, the NRC staff considered the licensees response to RAI 5 to be acceptable.

In RAI 6, the NRC staff questioned the appropriateness of the licensees use of a single RCP trip delay time of 5 minutes across the entire range of breaks considered in its delayed RCP trip OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION sensitivity analysis. In particular, RAI 6 arose from the staffs expectation that a 5-minute RCP trip delay may not represent either the most limiting or most likely trip delay, particularly when considering breaks at the larger end of the small-break spectrum (e.g., break diameter 5 inches).

Following the accident at Three Mile Island (TMI), Unit 2, in 1979, analysis of RCP operation during postulated small-break LOCA events was performed to develop an improved understanding of the impact of this important phenomenon. As described in NUREG-0623 (Reference 33) and NUREG-1230 (Reference 34), these early analyses demonstrated that RCP operation during a small-break LOCA could have a beneficial impact on core cooling - if adequate cooling is maintained throughout the event. However, if RCPs operate initially, but are interrupted or cease to be effective partway through a small-break LOCA, the event could become significantly more severe.

Physically, RCP operation tends to homogenize the two-phase flows that emerge in the RCS as the small-break LOCA event progresses. Hence, after the RCS drains to the point at which break uncovery would occur under stratified, natural circulation flow conditions, continued RCP operation tends to exacerbate the ongoing loss of system fluid mass relative to an analogous scenario with the RCPs tripped at that point. Conversely, continued RCP operation can greatly enhance two-phase flow through the core, which may significantly improve core cooling under conditions with depleted inventory. Therefore, tripping the RCPs after they have accelerated the depletion of RCS inventory, but shortly prior to or during the core uncovery phase of the event, can produce more limiting conditions than either an early RCP trip or continuous RCP operation throughout the small-break LOCA event.

In RAI 6, the NRC staff specifically questioned the licensees sensitivity studys treatment of breaks larger than approximately 5 to 6 inches in diameter, for which the peak cladding temperature may occur prior to the simulated tripping of the RCPs with a 5-minute delay following the loss of hot leg subcooling margin. Based on the licensee's response to RAI 5 (see Figure 3 and Figure 4 above), the NRC staff expected that an assumed 5-minute RCP trip delay would provide non-limiting results in this portion of the postulated break spectrum.

In RAI 6, the NRC staff noted that, realistically, offsite power may remain available in many potential small-break LOCA scenarios, such that the RCPs would remain operational beyond the time at which the reactor trips. he licensee stated that plant operators would be expected to trip the RCPs manually following a loss of subcooling margin well prior to the 5-minute delay time assumed in the sensitivity analysis. The NRC notes thateffective operation of RCPs under the highly voided conditions that evolve in the RCS loops on the approach to core uncovery cannot be assured even in the absence of a manual trip. For this reason, the NRC staff questioned, not only whether the licensees assumption of a 5-minute delay period could be nonconservative for breaks larger than approximately 5 to 6 inches in diameter, but also whether shorter RCP trip delay times would not be more likely than the scenarios analyzed by the licensee.

The licensees response to RAI 6 provided a qualitative, historical review of background material concerning the analysis of RCP trip timing for the SBLOCA event. Much of the historical analyses and derivative insights discussed by the licensee originated in response to the 1979 accident at TMI. As discussed with the licensee during the regulatory audit conducted in October 2018 (Reference 35), the post-TMI analyses relied upon computer codes developed during the 1960s and 70s with significantly simpler modeling practices than modern codes OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION (e.g., 10-20 fluid nodes, simplified field equations). The post-TMI analyses also generally focused upon smaller break sizes (e.g., 2-4 inches), as opposed to the larger range of small breaks of concern to RAI 6 (i.e., 5 inches and larger) that contemporary analyses show have the potential to be limiting for many PWRs implementing modern analysis methods ((

)). The licensees response did not reference any RCP trip time sensitivity calculations using modern analysis methods that are applicable to North Anna or Surry for the ranges of break size and RCP trip delay time of interest to RAI 6.

Based on the above, the NRC staff performed confirmatory analyses using the TRACE thermal-hydraulic Code to obtain additional physical insight and an estimate of issue significance. For some breaks in the range of concern to RAI 6, the NRC staffs confirmatory simulations displayed sensitivities to reduced RCP trip delay times of approximately 100 °F or more.

Subsequently, in a regulatory audit conducted in January 2020 (Reference 36), the NRC staff observed the results of sensitivity analyses performed by the licensee with the EMF-2328, Supplement 1, methodology for North Anna using a 1-minute RCP trip delay time. The licensee characterized the 1-minute trip delay as a best-estimate of the expected operator response time. The licensee's sensitivity analyses demonstrated peak cladding temperature increases of a magnitude similar to the NRC staffs confirmatory calculations. However, the licensee's sensitivity analysis omitted the 6.5-inch break case that was found to be limiting from the baseline break spectrum for North Anna, such that the NRC staff could not reliably infer the impact on the global peak cladding temperature. Therefore, the NRC staff issued supplemental RAI 6 S1, which described the concerns summarized above and requested that the licensee address the potential for RCP trip delays less than 5 minutes to result in significantly higher peak cladding temperatures than those originally reported in the submitted amendment request, or provide justification that consideration of such trip delays is unnecessary following adoption of the proposed license amendment.

In response, the licensee provided both (1) justification that additional consideration of RCP trip delays is unnecessary, as well as (2) the results of additional sensitivity analyses, for North Anna only, demonstrating the impact of a reduced, best-estimate RCP trip delay time.

Concerning the first point, the licensees RAI response cited several passages from the NRC staff's safety evaluation on EMF-2328, Supplement 1, that the licensee considered as supporting its position that its proposed licensing basis analysis needs only to consider scenarios that assume an RCP trip contemporaneous with reactor trip, regardless of whether such scenarios lead to limiting results relative to the criteria of 10 CFR 50.46(b).

The NRC staff observed that the licensees interpretation of the approved EMF-2328, Supplement 1, evaluation model had not considered all relevant evidence in reaching its conclusion, including the following items from the topical report and associated safety evaluation:

The NRC staffs safety evaluation for EMF-2328, Supplement 1, in Section 4.4 calls for a spectrum of breaks to be performed to examine RCP trip time, if necessary to support the RCP trip procedure or address an RAI from the NRC staff.

Supplement 1 of EMF-2328, indicates in Section 5.2 that analysis of the delayed RCP trip action will be performed with the EMF-2328, Supplement 1, evaluation model. As defined in 10 CFR 50.46(c)(2), an evaluation model includes not just computer codes, OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION but further encompasses all other information necessary to perform the calculation, including specification of mathematical models, assumptions, and values of parameters.

The NRC staffs safety evaluation for EMF-2328, Supplement 1, indicates in Section 4.4 that analysis of RCP trip delays is necessary because such scenarios may be limiting and must be prevented from exceeding acceptance criteria limits (i.e., in 10 CFR 50.46(b)).

The licensees response further discussed Generic Letter 85-12, Implementation of TMI Action Item II.K.3.5, Automatic Trip of Reactor Coolant Pumps (Reference 37). Following the TMI accident in 1979, the NRC staff had considered requiring that PWR licensees implement safety-grade instrumentation capable of automatically tripping RCPs during a small-break LOCA.

However, as described in Generic Letter 85-12, industry owners groups proposed including manual RCP trip criteria in emergency operating procedures as an alternative.8 Generic Letter 85-12 documents the NRC staffs approval, on a generic basis, of the Westinghouse Owners Group proposal for establishing criteria for manually tripping the RCPs under small-break LOCA conditions, including the reactor coolant subcooling method the licensee implemented at Surry and North Anna. While the licensee's historical discussion provides general background context, the NRC staff noted that the licensee's proposed FVI SBLOCA methodology involves no changes to the previously approved manual RCP trip criteria discussed in Generic Letter 85-12.

The licensee further observed in its response to RAI 6 S1 that a 5-minute RCP trip delay was applied in the original Westinghouse analysis supporting its current manual RCP trip criteria.

However, the licensee acknowledged the NRC staffs concern that these analyses relied upon early computer code models with significantly simplified modeling practices and further focused on breaks below the range of interest to RAI 6. The licensee stated that it performed the following two sensitivity studies for North Anna using a best-estimate RCP trip time of 1 minute following a loss of subcooling margin:

A sensitivity study that applied the full EMF-2328, Supplement 1, evaluation model. The results of this sensitivity study, included in Figure 6, demonstrated a peak cladding temperature of 1801 °F at a break size of 6.5 inches.

A sensitivity study that applied a modified version of the EMF-2323PA, Revision 0, Supplement 1PA, Revision 0, evaluation model, using a nominal decay heat power and a critical flow model considered more realistic than the Moody model required by Appendix K to 10 CFR 50. The peak cladding temperature from this sensitivity study was (( )), which demonstrates a sensitivity associated with the decay heat and critical flow modeling assumptions (( )).

Of these two sensitivity studies, the NRC staff considered the first study particularly relevant to the concerns identified in RAI 6 S1, since, in accordance with Section 5.2 of EMF-2328, Supplement 1, and 10 CFR 50.46(c)(2), since it was performed in accordance with the full, approved EMF-2328, Supplement 1, evaluation model.

8 Generic Letter 85-12 specifically addresses the effort performed by the owners group for Westinghouse PWRs.

Similar correspondence addresses the parallel efforts undertaken by the owners groups for the other PWR vendors.

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((

))

Regarding the second sensitivity study, the NRC staff expects that the calculated margin was generated predominately by assuming a nominal decay heat power in lieu of the 1.2 multiplier required by Appendix K.9 Appendix K requires incorporation of conservative modeling practices for decay heat and other significant parameters to account for phenomenological uncertainty.

The omission of conservative assumptions for decay heat power and critical flow would not conform to Section 5.2 of EMF-2328, Supplement 1, or 10 CFR 50.46(a)(1)(i). Furthermore, considering that the conservative modeling practices required by Appendix K pertain mainly to the large-break LOCA event, a calculation using a nominal value for one of the few dominant phenomena influencing the small-break LOCA event would not provide sufficient confidence that regulatory limits are satisfied for postulated scenarios involving a manual RCP trip.

While the requirements of Appendix K do not specifically address RCP modeling practices for a small-break LOCA event, insofar as Appendix K recognizes the significance of RCP behavior under LOCA conditions, it calls for a conservative treatment. For instance, as relevant to the large-break LOCA scenario, Appendix K requires sensitivities to determine the more conservative condition between a locked or freely spinning pump impeller. Moreover, when considering past calculations of RCP trip behavior during a small-break LOCA event, for the smaller break size range typically examined in those calculations (e.g., 2- to 4-inch breaks), the NRC staff observes that the assumption of a 5-minute delay time for the manual RCP trip may be regarded as an analogous conservatism. Specifically, the assumed 5-minute delay time significantly overestimates the expected time required for operators to trip the RCPs, which was 9

As discussed above in Section 3.2.2.1, the influence of critical flow modeling is expected to be more subtle, due to countervailing impacts. While a larger critical flow rate may lead to more a more rapid loss of RCS inventory and earlier core uncovery, it will also bring about a more rapid injection of accumulator inventory and earlier recovery.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION intended to produce conservative results for scenarios with an extended event progression (e.g., on the order of half an hour or more).10

((

)). For this range of the small-break spectrum (e.g., (( ))), where the peak cladding temperature may be reached within a few minutes of event initiation, a 5-minute RCP trip time is potentially nonconservative, and the assumption of a 1-minute trip time represents a best-estimate value without additional conservatism, such as had been included in past analyses in the smaller range of the break spectrum that selected a longer RCP trip delay time.

By letter dated October 22, 2020 (Reference 38), the licensee responded to RAI 6 S1 bysubmitting information relevant to the 1-minute delayed RCP trip scenario for both North Anna and Surry:

For North Anna, the licensee stated that, following implementation of the FVI SBLOCA methodology, the licensing basis peak cladding temperature would be established as 1801 °F, based on the result calculated for the 1-minute delayed RCP trip scenario. The licensee further confirmed that this scenario would be considered when estimating the effect of future changes or error corrections to its small-break LOCA analysis.

The licensee stated that the RCP trip sensitivity study performed for North Anna also applies to Surry. The licensee based its position on several points, including the similarity of the plants reactor coolant system and ECCS designs, the use of the same analytical methodology, and similar predicted peak cladding temperature trends exhibited in the small-break LOCA break spectrum analyses for both plants. The licensee concluded that the limiting peak cladding temperature break scenario for Surry is presently not impacted by the 1-minute delayed RCP trip assumption.

The consideration of the predicted limiting event in the present small-break LOCA analysis for North Anna (i.e., which involves a 1-minute delayed RCP trip) and will be adequately controlled under the reporting requirements specified in 10 CFR 50.46(a)(3) and thus, provides confidence that the licensee will maintain compliance with the acceptance criteria of 10 CFR 50.46(b).

Based on the following key points, the NRC staff finds the licensees supplemental response for Surry to address the NRC staff concerns to be acceptable:

The calculated limiting break size for Surry is 2.6 inches, which results in a peak cladding temperature of 1673 °F. ((

)).

10 Note that the 2016 adoption of EMF-2328, Supplement 1, at Millstone Power Station, Unit 2, cited in the licensee's response to RAI 6 S1, falls into this mold. The limiting break size for the Millstone 2 small-break LOCA analysis was 3.78 inches, and its RCP trip sensitivity focused on breaks between 3 and 5 inches. These break sizes are below the range where the concerns expressed by the staff in RAI 6 of the present review become significant.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION Both the licensee's simulations for North Anna (see Figure 6 above) and confirmatory simulations performed by the NRC staff for a Westinghouse 3-loop plant indicate that the expected difference between modeling an RCP trip (1) at the time of reactor trip as compared to (2) one minute following a loss of RCS subcooling should diminish for a

(( )). Physically, as the break size increases, the time to the peak cladding temperature is reduced, such that the cooling effect from the operation of the RCPs and their subsequent coastdown may begin to delay or, for larger ruptures in the small break spectrum, even prevent core heatup.

Furthermore, the more rapid depletion of RCS inventory for larger breaks tends to diminish the influence of RCP operation.

Because calculated peak cladding temperatures for the larger range of small-break LOCA scenarios tend to be slightly reduced for Surry as compared to North Anna, it is expected that, all else being equal, the sensitivity associated with the difference in RCP trip modeling would likewise be slightly reduced for Surry.

Based upon the above, the NRC staff determined that there is sufficient margin to account for the expected sensitivity between (1) the RCP trip on reactor trip scenario modeled for Surry and (2) the 1-minute delayed trip following a loss of subcooling margin scenario that was not modeled explicitly. While the NRC staff has reasonable assurance in this conclusion for the current plant configuration, the present review cannot encompass future plant or analysis method modifications the licensee may implement. Therefore, the NRC staff's does not rule out the possibility that applicable regulations may require the licensee to perform analysis of scenarios at Surry involving a best-estimate RCP trip delay in the future.

On the above, the NRC staffs review found that the licensees proposed FVI SBLOCA methodology would appropriately address potentially limiting small-break LOCA scenarios involving a delayed trip of the reactor coolant pumps. The NRC staffs review found the licensee's proposed treatment of delayed RCP trip scenarios, as modified by its RAI responses and supplementary submittal, to provide an acceptable basis for the NRC staffs conclusions supporting reasonable assurance for approval of the methodology described in EMF-2328, Supplement 1. The licensee's proposed treatment of delayed RCP trip scenarios is consistent with the requirement to ensure that the acceptance criteria of 10 CFR 50.46(b) remain satisfied for the most severe LOCAs in the range of accident conditions postulated in 10 CFR 50.46(a)(1)(i). Furthermore, the licensee's proposed consideration of delayed RCP trip scenarios when assessing changes and error corrections to its small-break LOCA analysis is consistent with the reporting requirements in 10 CFR 50.46(a)(3) and is, therefore, acceptable.

Finally, while the NRC staff has concluded that consideration of a 1-minute delayed RCP trip time within the safety analysis for North Anna and Surry is necessary to assure satisfaction of the acceptance criteria in 10 CFR 50.46(b) for larger small-break LOCA events, the NRC staff also recognizes that the licensees Time Critical Action program currently applies a 5-minute RCP trip delay time. While the licensee's Time Critical Action program is beyond the scope of the present review, the NRC staff recognizes that assuring that plant operators can trip the RCPs within 5 minutes of indication of a loss of RCS subcooling remains a part of the licensee's strategy for assuring satisfaction of the acceptance criteria in 10 CFR 50.46(b) across the full range of postulated small-break LOCA events.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.2.2.5 (( ))

The licensee completed a safety injection (( )) following the direction provided in EMF-2328(P)(A), Revision 0, Supplement 1 (Reference 4). The NRC staff reviewed the applicable sections of the FSAR and the TS for Surry and North Anna to confirm

((

)).

The NRC staffs SE for EMF-2328, Supplement 1, ((

)). The NRC staff verified ((

)). The NRC staff also reviewed the licensees sensitivity study for safety injection (( )) and concluded that the approach the licensee took concerning the initial safety injection (( )) is adequate and compliant with the NRC staffs safety evaluation.

Beyond this, the NRC staff considered additional sources of safety injection within the present review that could impact the determination of an upper bound PCT. The NRC staff notes that, after the refueling water storage tank (RWST) is depleted by ECCS and containment sprays, transferring suction of ECCS pumps to the containment sump ((

)). Depending on the break size, switchover to the containment sump could be required before the core is quenched. Hence, realistically modeling the switchover of the safety injection source (( )) could result in the calculation of a higher PCT. The NRC staff requested information in RAI 7 to determine whether an RWST depletion study had been performed for the limiting break, and if a

(( )) was utilized.

The licensees response indicated an RWST depletion sensitivity study had been performed that supports the conclusion that the PCT results of the analyses of record, as presented in the submittals, remain bounding. The licensees response also identified the safety injection water

((

)). While the NRC staff found the ((

)). The response also did not indicate whether the sensitivity analyses examined RWST depletion and switchover at times before or after the core is quenched; the timing of safety injection switchover could significantly impact the calculation of PCT.

In a supplemental response to RAI 7, the licensee provided additional justification for the post-RWST drain down safety injection water temperature. The licensee indicated that the value was derived from the containment response to the LBLOCA conditions analyzed in the North Anna and Surry UFSARs. In these analyses, the rate of mass and energy release is greater than that for SBLOCA conditions, resulting in a bounding sump temperature for SBLOCA conditions.

Additionally, substantial conservatisms were incorporated into the analyses. These OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION conservatisms included the assumption of one train of recirculation spray operating at minimum flow rates, one train of service water operating at minimum flow rates, and the modeling of degraded recirculation spray heat exchanger performance. The NRC staff finds these conservatisms (( )). Therefore, based on the information provided in the supplemental response, the NRC staff finds the containment sump temperature of (( )) used for safety injection following depletion of the RWST is acceptable for the SBLOCA conditions present at North Anna and Surry.

The supplemental response to RAI 7 also indicated all the break sizes included in the analyses of record were evaluated for the RWST depletion sensitivity studies, and re-analyses were performed for those break sizes wherein the RWST drains before the core is quenched. The justification provided for this approach is that, if the core is already quenched by the time of RWST depletion, then continued safety injection ((

)) would not be capable of causing an additional temperature excursion. The NRC staff finds this approach acceptable because, in a post-core-quench switchover scenario, abundant injection of core coolant ((

)) would be sufficient to ensure continued removal of decay heat from the core. The supplemental response results indicate that, in all re-analyzed cases (i.e., those for which the RWST drains before core quench), ((

)). The licensee found that the limiting values of the figures of merit for the RAI 7 sensitivity studies remain bounded by the limiting case from the analyses of record.

The NRC staff reviewed the results from the licensee's safety injection temperature sensitivity study, the base break spectrum analysis, the response to RAI 7 concerning containment sump temperature, and the information provided in the supplemental response to RAI 7. In light of the conservative nature (( )) assumed by the licensee, which plays a significant role in these analyses, the NRC staff finds these results acceptable and supportive of the licensees assertion that the timing of RWST depletion does not challenge the PCT results in the analyses of record and is, therefore, acceptable.

3.2.2.6 Loop Seal Clearing and Crossover Leg Modeling The NRC staffs safety evaluation approving Supplement 1 to EMF-2328P imposed a requirement for modeling loop seals on the RCP suction piping in a manner that ensures appropriate conservatism. ((

)) since the test data reviewed by the NRC staff indicated that additional loop seals may realistically be expected to clear.

The approved Supplement 1 methodology ((

)).

Based upon piping dimensions and specifications included in the plant FSARs, the NRC staff confirmed that, in accordance with the methodology described in Section 7.3.1 of OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION

)). The NRC staff examined the North Anna and Surry core models and confirmed core nodalizations of (( ))

are used, which is consistent with the approved EMF-2328, Supplement 1, topical report.

3.3 Plant Parameters and Initial Conditions The NRC staff reviewed the plant parameters and initial conditions used in the analysis for North Anna and Surry to ensure that the values are appropriate for the SBLOCA event and consistent with the plants current design bases. Specifically, the NRC staff reviewed Table 3-1 in the proprietary vendor licensing reports attached to the LARs for North Anna (ANP-3467) and Surry (ANP-3676), which contain the plant parameters and initial conditions used in the FVI SBLOCA analysis. The NRC staff confirmed that all parameters in these tables are consistent with the current licensing basis for North Anna and Surry , are bounding, or are governed by the analysis (e.g., FQ), with one exception.

The value of the low pressurizer pressure setpoint for reactor trip in Table 3-1 of the LAR for Surry (1899.7 psia) is inconsistent with the value in Surry TS 2.3 (1889.7 psia). For a SBLOCA, a lower setpoint may be more conservative, since it will delay the reactor trip, which allows additional fission heat to be generated in the core and increases the decay heat at the time of core uncovery. Therefore, the NRC staff requested in RAI 9 that the licensee justify use of a value for the reactor trip setpoint on low pressurizer pressure in the analysis performed for Surry that exceeds the TS value.

In response to RAI 9, the licensee stated that the value of the low pressurizer pressure trip setpoint in Table 3-1 of the proprietary vendor licensing report supporting the Surry LAR was in error; the value the licensee actually used in its analysis was (( )). Since the analyzed value was boundingly low relative to the TS value, the NRC staff determined that the analysis was conservative with respect to modeling the low pressurizer pressure setpoint for reactor trip.

3.4 Calculated Results 3.4.1 Review of Licensees Calculations The licensee provided results for its break spectrum analyses in Tables 4-1 and 4-2 of ANP-3467 for North Anna and ANP-3676 for Surry. The peak cladding temperatures the licensee predicted for the limiting cold leg break spectrum calculations have been shown above graphically in Figure 3, Figure 4, and Figure 6. As shown in Figure 6, the licensee's supplementary submittal (Reference 38) confirmed that a break scenario involving a manual RCP trip 1-minute after a loss of RCS subcooling is more limiting for North Anna than the contemporaneous pump trip scenario originally presented as limiting in ANP-3467.

As discussed above in Section 3.2.2.1, the NRC staffs review confirmed that the licensee adequately addressed the modeling requirements concerning break spectrum analysis in OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION EMF-2328, Supplement 1. The NRC staffs review of the licensee's break spectrum results further found that the licensee's break spectrum predictions display small-break LOCA response characteristics typical of many pressurized-water reactors. In particular, the break spectrum calculations for North Anna and Surry ((

)). While specific details of the break spectrum curves can vary significantly between plants, general features of the break response tend to be consistent based on the similarity of the design characteristics shared by many pressurized-water reactors.

In addition to providing results sufficient to characterize the spectrum of analyzed breaks, the licensee provided plots of key parameters for break scenarios representative of conditions that would produce the limiting peak cladding temperature for each plant:

For North Anna, plots of key parameters for the 6.5-inch break were provided in Figures 4-2 through 4-20 of ANP-3467.11 For Surry, plots of key parameters for the 2.6-inch break were provided in Figures 4-2 through 4-20 in ANP-3676.

The NRC staff reviewed the information provided for these scenarios to ensure that the licensee's predictions for key event parameters are consistent with the physical behavior expected for an SBLOCA event. In particular, the NRC staff observed the physical behavior predicted during each major phase of the SBLOCA event (e.g., blowdown, natural circulation, loop seal clearing, core boil-off, and core recovery) to be generally consistent with expected event behavior. Following the resolution of RAI 10 (discussed below), the NRC staff further found that the licensee's analyses overall reflected appropriate plant responses after reaching the various instrumentation setpoints relevant to the small-break LOCA event.

As addressed in RAI 10, the NRC staffs review of the results contained in Section 4 of the plant-specific methodology reports, ANP-3467P for North Anna and ANP-3676P for Surry, identified a question concerning the modeling of the reactor trip on low pressurizer pressure.

((

11 Note that, as discussed above in Section 3.2.2.4, the licensee's supplementary submittal (Reference 38) confirmed that the 6.5-inch break scenario modeling a manual RCP trip with a 1-minute delay discussed in response to RAI 6 S1 (Reference 10) is more limiting than the case in ANP-3467 that simulated an RCP trip contemporaneous with reactor trip. However, the overall physical behavior for both scenarios remains generally consistent, such that the information the licensee presented for the 6.5-inch break scenario in ANP-3467 sufficiently demonstrates the prediction of key phenomena relevant to the limiting event scenario.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAb U$1! ONbY PROPRll!TARY INFORMATION

)) whereas a break size ([ )) yields a low pressurizer pressure tnp time of 0.69 seconds. A similar dIscont1nuIty appears for Surry, albeit at a break size below ((

)).

The licensee's response to RAI 1O confirmed that the results submitted in plant-specific methodology reports ANP-3467P and ANP-3676P are correct. The response further explained that the apparent discrepancies identified during the NRC staff's review are associated with the processing of the pressurizer pressure signal through lead-lag circuitry designed to provide an anticipatory function to the low-pressurizer pressure trip. As suggested in RAI 10, the licensee supported its response with plots at a significantly finer time resolution for several analysis cases, including the two examples from RAI 10 that are noted above. These refined plots illustrate expected behavior of the calculated pressurizer pressure along with the lead-lag compensated signal used for the low-pressurizer pressure trip.

The NRC staff's review found the licensee's response to RAI 10 acceptable because it adequately addressed the apparent inconsistencies associated with low-pressurizer-pressure induced reactor trips identified during the initial review and provided sufficient supporting evidence. The NRC staff further confirmed that Section 7.2 of the UFSAR for both plants refers to lead-lag compensation of the pressurizer low-pressure reactor trip.

3.4.2 Compliance with 10 CFR 50.46 Acceptance Criteria The limiting values of the peak cladding temperature (PCT), the transient maximum local oxidation (MLO), and the core-wide oxidation (CWO) calculated by the licensee for the postulated small-break LOCA event (References 1, 2, and 38) are summarized below in Table 3. The limiting results for each plant are bolded and underlined. The relevant acceptance criteria from 10 CFR 50.46 are summarized below in Table 4.

Table 3: Limiting Results for Each 50.46(b) Acceptance Criterion Plant Break Size in Transient MLO %

6.5 1.71 --------

NAPS

(( )) -=-+--

(( ],..._

] ------1--

2.6 --- (( ))---

SPS 2.5 1.43 Table 4: Relevant Acceptance Criteria from 1 0 CFR 50.46 Paragraph Figure of Merit Acceptance Criterion (b)( 1) Peak Cladding Temperature  ::. 2200 °F (b)(2) Maximum Local Oxidation :s 17% of Unoxidized Thickness (b)(3) Core-Wide Oxidation :s 1% of Hypothetical Amount (b)(4) Core Geometry Amenable to Cooling 12 Reported as< 0.04% in Section 2.0 of ANP-3467 13 Reported as< 0.06% in Section 2.0 of ANP-3676 OFFICIAb U$1! ONbY PROPRll!TARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION The values reported for MLO in Table 3 represent the oxidation occuring during the LOCA transient and do not include the additional oxidation that occurs during normal plant operation.

However, as discussed in Information Notice 98-29, consideration of pre-transient oxidation is also necessary to ensure compliance with 10 CFR 50.46(b)(2). As discussed in its submittals (References 1 and 2), the licensee recognizes that 10 CFR 50.46(b)(2) requires that the maximum total cladding oxidation (i.e., pre-transient plus transient) remain less than 17% of the original, unoxidized cladding thickness. For both North Anna and Surry, the licensee has assessed the maximum total cladding oxidation and confirmed that it will remain below the 17%

acceptance criterion throughout the analyzed life of the fuel (References 1 and 2).

The cladding temperature and oxidation results the licensee calculated for both plants remain within the limits of 10 CFR 50.46(b)(1) through (b)(3) and are, therefore, acceptable. The satisfaction of these acceptance criteria further provides adequate assurance that a coolable core geometry will be maintained, in accordance with 10 CFR 50.46(b)(4).

3.5 Evaluation of Proposed TS Changes The TS changes proposed by the licensee, including deletions and additions, were detailed in Section 2.2 of this SE.

For North Anna, the licensee proposed to add the following to TS 5.6.5.b:

3. EMF-2328(P)(A), PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, as supplemented by ANP-3467P, North Anna Fuel-Vendor Independent Small Break LOCA Analysis, as approved by NRC Safety Evaluation Report dated [DATE].

For Surry, the licensee proposed to add the following to TS 6.2.C:

3. EMF-2328(P)(A), PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, as supplemented by ANP-3676P, Surry Fuel-Vendor Independent Small Break LOCA Analysis, as approved by NRC Safety Evaluation Report dated [DATE].

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 proposed addition to the COLR will adequately proscribe the usage of the aforementioned methodologies as approved by this amendment and, therefore, meets 10 CFR 50.36(c)(5).

4.0 CONCLUSION

The NRC staff has reviewed changes the licensee proposed to the Core Operating Limits Report section of the technical specifications for North Anna and Surry to implement the proposed FVI SBLOCA methodology. As described in Section 2.2 above, the proposed technical specification changes would replace the licensee's existing evaluation models for analyzing the postulated small-break LOCA event with a modified, plant-specific version of the OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION EMF-2328, Supplement 1, evaluation model developed by Framatome. The licensee's proposed modification to the EMF-2328, Supplement 1, evaluation model, as described in plant-specific methodology reports ANP-3467 for North Anna and ANP-3676 for Surry and further modified in the licensee's response to RAI 12.b.S.1 (Reference 8), would allow the licensee to use the EMF-2328, Supplement 1, evaluation model to analyze, not only applicable Framatome fuel designs, but also the Westinghouse fuel designs currently in use at North Anna and Surry.

Based on he above, the NRC staff determined the following:

The modifications the licensee proposed to the EMF-2328, Supplement 1, evaluation model, which support its application to the current Westinghouse fuel designs used at North Anna and Surry, are acceptable for the proposed application. The FVI SBLOCA methodology complies with the requirements of 10 CFR 50.46(a)(1)(i), insofar as it is an acceptable evaluation model for its intended purpose, and further conforms to the required and acceptable features specified in Appendix K to 10 CFR 50 (Section 3.1).

The licensee has addressed the limitations and required modeling practices associated with the EMF-2328, Supplement 1, evaluation model (Section 3.2).

The key analytical inputs to the FVI SBLOCA analyses described by the licensee in its submittals supporting the proposed license amendment are appropriate and, therefore, acceptable (Section 3.3).

The analytical results calculated by the licensee for the postulated small-break LOCA event for North Anna and Surry comply with the relevant acceptance criteria specified in 10 CFR 50.46(b). As such, the requirement for abundant core cooling in their respective design criterion is also satisfied (Section 3.4).

Therefore, as discussed above in Section 3.5, the NRC staff concludes that the proposed changes to the Core Operating Limits Report section of the technical specifications for North Anna and Surry, as described above in Section 2.2, are acceptable.

As discussed above in Section 3.1.4, however, the NRC staff has not generically reviewed or approved application of the FVI SBLOCA methodology for other plants. Nor has the NRC staff generically reviewed or approved the use of Framatomes EMF-2323PA, Revision 0, Supplement 1PA, Revision 0, evaluation model for application to non-Framatome fuel designs.

The NRC staff has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's 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.

5.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Commonwealth of Virginia official was notified of the proposed issuance of the amendments on June 23, 2019. On June 23, 2020 the state official confirmed that the Commonwealth of Virginia had no comments.

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OFFICIAL USE ONLY PROPRIETARY INFORMATION

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on October 9, 2018 (83 FR 50697) for Surry and December 4, 2018 (83 FR 62609) for North Anna. 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.

5.0 REFERENCES

1 Virginia Electric and Power Company, letter to U. S. Nuclear Regulatory Commission (NRC), "Virginia Electric and Power Company, North Anna Power Station, Units 1 and 2 -

Proposed License Amendment Request (LAR) Addition of Analytical Methodology to COLR Small Break Loss of Coolant Accident (SBLOCA)," July 12, 2018 (ADAMS Accession No. ML18198A133).

2 Virginia Electric and Power Company, letter to U. S. Nuclear Regulatory Commission (NRC), "Virginia Electric and Power Company, Surry Power Station, Units 1 and 2 -

Proposed License Amendment Request - Revision of Analytical Methodology Refeence in Core Operating Limits Report for Small Break Loss of Coolant Accident," July 31, 2018 (ADAMS Accession No. ML18218A181).

3 Framatome ANP Richland, Inc., (Formerly AREVA,NP, Inc.) letter to U. S. Nuclear Regulatory Commission (NRC), "Publication of EMF-2328(P)(A) Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," May 9, 2001 (ADAMS Accession No. ML011410426).

4 Framatome, Inc. (formerly AREVA, NP, Inc.,) letter to U. S. Nuclear Regulatory Commission (NRC), "Replacement Publication of EMF-2328(P)(A), Revision 0, Supplement 1PA, Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based," March 17, 2017 (ADAMS Accession No. ML17082A170).

5 Virginia Electric and Power Company (Dominion Energy Virginia) and Framatome, Inc.,

Report to U. S. Nuclear Regulatory Commission (NRC), "North Anna Fuel-vendor Independent Small Break LOCA Analysis, Licensing Report (ANP-3467P, Revision 0)," May 2018 (ADAMS Accession No. ML18198A120 - Non-Public).

6 Virginia Electric and Power Company (Dominion Energy Virginia) and Framatome, Inc.,

Report to U. S. Nuclear Regulatory Commission (NRC), "Surry Fuel-vendor Independent Small Break LOCA Analysis - Licensing Report (ANP-3676NP, Revision 0)," July 2018 (ADAMS Accession No. ML18218A171).

7 U. S. Nuclear Regulatory Commission (NRC), letter to Daniel G. Stoddard, "North Anna Power Station, Units Nos. 1 and 2; and Surry Power Station, Unit Nos. 1 and 2 - Request for Additional Information," February 8, 2019 (ADAMS Accession No. ML19032A055).

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OFFICIAL USE ONLY PROPRIETARY INFORMATION 8 Virginia Electric and Power Company, letter to U. S. Nuclear Regulatory Commission (NRC), "Virginia Electric and Power Company, North Anna and Surry Power Stations, Units 1 and 2 - Proposed License Amendment Requests - Addition of Analytical Methodology to the Core Operating Limits Report for a Small Break Loss of Coolant Accident (SBLOCA),"

Response to Request for Additional Information and Analysis Error Correction, July 9, 2019 (ADAMS Pkg Accession No. ML19196A124).

9 U. S. Nuclear Regulatory Commission (NRC), letter to Virginia Electric and Power Company, "Request for Additional Information Related to Virginia Electric and Power Company, Surry Power Station Units 1 and 2, North Anna Power Station, Units 1 and 2 -

Request to Implement Fuel Vendor Independent Evaluation Models," October 21, 2019 (ADAMS Pkg Accession No. ML19322B112 - Non-Public).

10 Virginia Electric and Power Company, letter to U. S. Nuclear Regulatory Commission (NRC), "Virginia Electric and Power Company, North Anna and Surry Power Stations Units 1 and 2, Proposed License Amendment Requests - Addition of Analytical Methodology to the Core Operating Limits Report for a Small Break Loss of Coolant Accident (SBLOCA) ,"

Response to Request for Additional Information and Analysis Error Correction, May 28, 2020 (ADAMS Accession No. ML20149K694).

11 Virginia Electric and Power Company, letter to U. S. Nuclear Regulatory Commission (NRC), "Virginia Electric and Power Company, North Anna and Surry Power Stations Units 1 and 2 - Proposed License Amendment Requests - Addition of Analytical Methodology to the Core Operating Limits Report for a Small Break Loss of Coolant Accident (SBLOCA),"

Supplement to the Response to Request for Additional Information, October 22, 2020 (ADAMS Accession No. ML20296A727).

12 Westinghouse Electric Corporation, letter to U. S. Nuclear Regulatory Commission (NRC),

"NOTRUMP Code and Small Break ECCS [emergency core cooling system] Evaluation Model (WCAP-10054-P-A)," August 19, 1985 (ADAMS Accession No. ML100050586 - Non-Public).

13 Westinghouse Electric Company, letter to U. S. Nuclear Regulatory Commission (NRC),

"NOTRUMP Code and Small Break ECCS [emergency core cooling system] Evaluation Model (WCAP-10079-P-A)," August 19, 1985 (ADAMS Accession No. ML100060364 - Non-Public).

14 Westinghouse Electric Company, letter to U. S. Nuclear Regulatory Commission (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 Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," October 2, 2017 (ADAMS Accession No. ML17277A131, Public and Non-Public).

15 Virginia Electric and Power Company, letter to U. S. Nuclear Regulatory Commission (NRC), "Virginia Electric and Power Company, North Anna Power Station, Units 1 and 2 -

Proposed License Amendment Request (LAR) Addition of Analytical Methodology to COLR Best-Estimate Large," Break Loss-of-Coolant Accident (BE-LBLOCA) October 21, 2010 (ADAMS Accession No. ML102980447).

16 U. S. Nuclear Regulatory Commission (NRC), letter to Virginia Electric and Power Company, "Surry Power Station, Units 1 and 2, Virginia Electric and Power Company, LLC, Exemption from the Requirements of 10 CFR Part 50, Section 50.46," December 15, 2010 (ADAMS Accession No. ML103481056).

17 U. S. Nuclear Regulatory Commission (NRC), letter to Virginia Electric and Power Company, "North Anna Power Station, Unit Nos. 1 and 2, Exemption from the OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION Requirements of Title 10 of the Code of Federal Regulations, Part 50, Section 50.46 and Appendix K," March 23, 2011 (ADAMS Accession No. ML110600388).

18 U. S. Nuclear Regulatory Commission (NRC), "NUREG-0800, Standard Review Plan, Chapter 15.6.5, Loss-of-Coolant Accidents Resulting from Spectrum Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary, Revision 3," March, 2007 (ADAMS Accession No. ML070550016).

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

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

21 U. S. Nuclear Regulatory Commission (NRC) and Los Alamos National Laboratory, "NUREG/CR-6744, Phenomenon Identification and Ranking Tables (PIRTs) for Loss-of-Coolant Accidents in Pressurized and Boiling Water Reactors Containing High Burnup Fuel," December 2001 (ADAMS Accession No. ML013540584).

22 U. S. Nuclear Regulatory Commission (NRC), "NRC Information Notice 2009-23: Nuclear Fuel Thermal Conductivity Degradation and Supplement 1," October 8, 2009 (ADAMS Accession No. ML091550527 and ML121730336).

23 Framatome, Inc. (formerly AREVA NP, Inc.), Letter to U. S. Nuclear Regulatory Commission (NRC), "Informational Transmittal Regarding Requested White Papers on the Treatment of Exposure Dependent Fuel Thermal Conductivity Degradation in Legacy Fuel Performance Codes and Methods, Attachment A - BWR White Paper," July 14, 2009 (ADAMS Accession No. ML092010158 - Non-Public).

24 Calvert Cliffs Nuclear Power Plant, LLC (A Joint Venture of Constellation Energy (CENG)),

"Calvert Cliffs Nuclear Power Plant, Units 1 and 2; Docket Nos. 50-317 & 50-318 -

Supplement to the License Amendment Request: Transition from Westinghouse Nuclear Fuel to AREVA Nuclear Fuel," December 30, 2019 (ADAMS Accession No. ML110040374).

25 U. S. Nuclear Regulatory Commission (NRC), "Regulatory Issue Summary (RIS) 2016-04, Clarification of 10 CFR 50.46 Reporting Requirements and Recent Issues with Related Guidance Not Approved For Use," April 19, 2016 (ADAMS Accession No. ML15324A296).

26 Westinghouse Nuclear Energy Systems, "Vantage 5 - Westinghouse Reference Core Report - Vantage 5 Fuel Assembly (WCAP 10444-P-A," September 1985 (ADAMS Accession No. ML080630434 - Non-Public).

27 Virginia Electric and Power Company, letter to U. S. Nuclear Regulatory Commission (NRC), "Virginia Electric and Power Company (Dominion) Submittal of Approved Topical Report VEP-FRD-42, Revision 2.0-A, Reload Nuclear Design Methodology for Application at North Anna and Surry Power Stations," August 29, 2003 (ADAMS Accession No. ML032680720).

28 Idaho National Engineering Laboratory for U. S. Nuclear Regulatory Commission (NRC),

"NUREG/CR-4945 - Summary of the Semiscale Program (1965-1986), Formal Report," July 1987 (ADAMS Accession No. ML20237G462).

29 Framatome, Inc. (Formerly AREVA, Inc.) letter to U. S. Nuclear Regulatory Commission (NRC), "Final Safety Evaluation for AREVA, NP Inc. Topical Report EMF-2103(P), Revision 3, "Realistic Large Break LOCA Methodology for Pressurized Water Reactors."," June 29, 2016 (ADAMS Pkg Accession No. ML16172A329).

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OFFICIAL USE ONLY PROPRIETARY INFORMATION 30 Virginia Electric and Power Company (Dominion), letter to U. S. Nuclear Regulatory Commission (NRC), "Virginia Electric and Power Company (Dominion) - North Anna Power Station, Unit 1, Revised Realistic Large Break LOCA (RLBLOCA) Results Addressing Error Corrections for Use of Framatome ANP Advanced Mark-BW Fuel," June 18, 2004 (ADAMS Accession No. ML041740346).

31 Virginia Electric and Power Company, letter to U. S. Nuclear Regulatory Commission (NRC), "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Proposed Technical Specifications Change Addition of ASTRUM Methodology to Core Operating Limits Report References and Revised Large Break LOCA Analysis," November 16, 2006 (ADAMS Accession No. ML063210473).

32 U. S. Atomic Energy Commission, "Rulemaking Hearing: Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors,"

December 26, 1973 (ADAMS Accession No. ML20236U832).

33 U. S. Nuclear Regulatory Commission (NRC), "NUREG-0623, Generic Assessment of Delayed Reactor Coolant Pump Trip during Small Break Loss-of-Coolant Accidents in Pressurized Water Reactors," November 1979 (ADAMS Accession No. ML19340E453).

34 U. S. Nuclear Regulatory Commission (NRC), "NUREG-1230, Compendium of ECCS Research for Realistic LOCA Analysis," December 1988 (ADAMS Accession No. ML20235R540).

35 U. S. Nuclear Regulatory Commission (NRC), "Audit Results Summary Report, Related to Virginia Electric and Power Company, North Anna Power Station Units 1 and 2 and Surry Power Station Units 1 and 2, Request to Implement Fuel Vendor Independent Evaluation Model For Small Break Loss-of-Coolant ," Accident Analysis, October 2018 (ADAMS Accession No. ML20266G233 - Non-Public).

36 U. S. Nuclear Regulatory Commission (NRC), letter to Daniel G. Stoddard, Virginia Electric and Power Company, "North Anna Power Station, Units 1 and 2 and Surry Power Station, Units 1 and 2 - Regulatory Audit Report Regarding License Amendment Request for Small Break Loss-of-Coolant Accident Analysis Methodology," February 13, 2020 (ADAMS Accession No. ML20034F330).

37 U. S. Nuclear Regulatory Commission (NRC), "Generic Letter (GL)85-112, Implementation of TMI Action Item II.K.3.5, "Automatic Trip of Reactor Coolant Pumps,"" June 28, 1985 (ADAMS Accession No. ML031150698).

38 Virginia Electric and Power Company, letter to U. S. Nuclear Regulatory Commission (NRC), "Virginia Electric and Power Company, North Anna and Surry Power Stations Units 1 and 2, Proposed License Amendment Requests, Addition of Analytical Methodology to the Core Operating Limits Report for a Small Break Loss-of-Coolant Accident," (SBLOCA)

Supplement to the Response to Request for Additional Information, Ocotber 22, 2020 (ADAMS Accession No. ML20296A272).

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OFFICIAL USE ONLY PROPRIETARY INFORMATION Principal Contributors: J. Borromeo, NRR B. Parks, NRR K. Heller, NRR J. Lehning, NRR Date: March 19, 2021 OFFICIAL USE ONLY PROPRIETARY INFORMATION

ML20325A088 (Package)

ADAMS Accession No. ML20325A096 (Proprietary SE)

ADAMS Accession No. ML20325A095 (Non-Proprietary SE) *Via SE Input OFFICE NRR/DORL/LSPB/PM NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DSS/SFNB/BC NAME GEMiller VThomas KGoldstein RLukes*

DATE 03/16/2021 01/25/2021 01/12/2021 10/07/2020 OFFICE NRR/DSS/STSB/BC NRR/DSS/SNSB/BC OGC - NLO NRR/DORL/LPL2-1/BC NAME VCusumano SKrepel KGamin MMarkley DATE 12/02/2020 01/22/2021 02/24/2021 03/19/2021