ML24296B235

Issuance of Amendment No. 291 License Amendment to Support the Implementation of Framatome Gaia Fuel (EPID L-2023-LLA-0150) (Non-Proprietary)

ML24296B235
Person / Time
Site:	Millstone
Issue date:	11/19/2024
From:	Richard Guzman NRC/NRR/DORL/LPL1
To:	Carr E Dominion Energy Nuclear Connecticut
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ML24296B236	List: ML24296B235
References
EPID L-2023-LLA-0150
Download: ML24296B235 (1)

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OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION November 19, 2024 Eric S. Carr President - Nuclear Operations and Chief Nuclear Officer Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

MILLSTONE POWER STATION, UNIT NO. 3 - ISSUANCE OF AMENDMENT NO. 291 TO SUPPORT IMPLEMENTATION OF FRAMATOME GAIA FUEL (EPID L-2023-LLA-0150)

Dear Eric Carr:

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed Amendment No. 291 to Renewed Facility Operating License No. NPF-49 for the Millstone Power Station, Unit No. 3 (Millstone, Unit 3). This amendment is in response to your application dated October 30, 2023, as supplemented by letter dated September 16, 2024.

The amendment revises the Millstone, Unit 3, technical specifications (TSs) to support the implementation of Framatome GAIA fuel, which is currently scheduled for insertion into the Millstone, Unit 3, reactor during the spring 2025 refueling outage. Specifically, the TS changes include updating the reactor core safety limits (TS 2.1.1.1), reducing the Reactor Trip System Instrumentation Trip Setpoint for the P-8 Interlock (TS Table 2.2-1, Item 18.c), and adding to the list of approved methodologies for the Core Operating Limits Report (TS 6.9.1.6.b). Additionally, the amendment approves mixed-core departure from nucleate boiling (DNB) penalties for application to retained DNB margin and the use of the design basis limits for a fission product barrier associated with MPS3 specific application of methods needed to support GAIA fuel implementation.

The NRC staff has determined that the related safety evaluation (SE) contains proprietary information pursuant to Title 10 of the Code of Federal Regulations section 2.390, Public inspections, exemptions, request for withholding. The proprietary information is indicated by bold text enclosed with ((double brackets)). The proprietary version of the SE is provided as enclosure 2. Accordingly, the NRC staff has also prepared a non-proprietary version of the SE, which is provided as enclosure 3.

to this letter contains proprietary information. When separated from, this document is DECONTROLLED.

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

Sincerely,

/RA/

Richard V. Guzman, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-423

Enclosures:

1. Amendment No. 291 to NPF-49

2. Safety Evaluation (Proprietary)

3. Safety Evaluation (Non-Proprietary) cc: Listserv without enclosure 2

DOMINION ENERGY NUCLEAR CONNECTICUT, INC., ET AL DOCKET NO. 50-423 MILLSTONE POWER STATION, UNIT NO. 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 291 Renewed License No. NPF-49

1.

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

A.

The application for amendment by Dominion Energy Nuclear Connecticut, Inc.

(DENC, the licensee), dated October 30, 2023, as supplemented by letter dated September 16, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-49 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 291 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated into the license. DENC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

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

Attachment:

Changes to Renewed Facility Operating License No. NPF-49 and the Technical Specifications Date of Issuance: November 19, 2024

ATTACHMENT TO LICENSE AMENDMENT NO. 291 MILLSTONE POWER STATION, UNIT NO. 3 RENEWED FACILITY OPERATING LICENSE NO. NPF-49 DOCKET NO. 50-423 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert 4

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

Remove Insert 2-1 2-1 2-7 2-7 6-21 6-21

(2)

Technical Specifications The Technical Specifications contained in Appendix A, revised through Amendment No. 291 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated into the license. DENC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

DENC shall not take any action that would cause Dominion Energy, Inc.

or its parent companies to void, cancel, or diminish DENC's Commitment to have sufficient funds available to fund an extended plant shutdown as represented in the application for approval of the transfer of the licenses for MPS Unit No. 3.

(4)

Immediately after the transfer of interests in MPS Unit No. 3 to DNC*, the amount in the decommissioning trust fund for MPS Unit No. 3 must, with respect to the interest in MPS Unit No. 3, that DNC* would then hold, be at a level no less than the formula amount under 10 CFR 50.75.

(5)

The decommissioning trust agreement for MPS Unit No. 3 at the time the transfer of the unit to DNC* is effected and thereafter is subject to the following:

(a)

The decommissioning trust agreement must be in a form acceptable to the NRC.

(b)

With respect to the decommissioning trust fund, investments in the securities or other obligations of Dominion Energy, Inc. or its affiliates or subsidiaries, successors, or assigns are prohibited.

Except for investments tied to market indexes or other non-nuclear-sector mutual funds, investments in any entity owning one or more nuclear power plants are prohibited.

(c)

The decommissioning trust agreement for MPS Unit No. 3 must provide that no disbursements or payments from the trust, other than for ordinary administrative expenses, shall be made by the trustee until the trustee has first given the Director of the Office of Nuclear Reactor Regulation 30 days prior written notice of payment. The decommissioning trust agreement shall further contain a provision that no disbursements or payments from the trust shall be made if the trustee receives prior written notice of objection from the NRC.

(d)

The decommissioning trust agreement must provide that the agreement cannot be amended in any material respect without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation.

• On May 12, 2017, the name Dominion Nuclear Connecticut, Inc. changed to Dominion Energy Nuclear Connecticut, Inc.

Renewed License No. NPF-49 Amendment No. 270-290, 291

MILLSTONE - UNIT 3 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, Reactor Coolant System highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the CORE OPERATING LIMITS REPORT; and the following Safety Limits shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained greater than or equal to 1.14 for the WRB-2M DNB correlation, or greater than or equal to 1.13 for the ORFEO-GAIA DNB correlation.

2.1.1.2 For Westinghouse fuel, the peak fuel centerline temperature shall be maintained less than 5080F, decreasing by 9F per 10,000 MWD/MTU of burnup. For Framatome fuel, the peak fuel centerline temperature shall be maintained less than 4901°F, decreasing linearly by 13.7°F per 10,000 MWD/

MTU of burnup.

APPLICABILITY:

MODES 1 and 2.

ACTION:

Whenever the Reactor Core Safety Limit is violated, restore compliance and be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

APPLICABILITY:

MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2750 psia be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MODES 3, 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

Amendment No. 173, 217, 236, 242, 281, 280, 290, 2-1 291

MILLSTONE - UNIT 3 TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT NOMINAL TRIP SETPOINT ALLOWABLE VALUE 16.

Turbine Trip a.

Low Fluid Oil Pressure 500 psig 450 psig b.

Turbine Stop Valve Closure 1 open 1 open 17.

Safety Injection Input from ESF N.A.

N.A.

18.

Reactor Trip System Interlocks a.

Intermediate Range Neutron Flux, P-6 1 x 10-10 amp 9.0 x 10-11 amp b.

Low Power Reactor Trips Block, P-7

1) Power Range Neutron Flux, P-10 input (Note 5) 11 of RTP**

11.6 of RTP**

2) Turbine Impulse Chamber Pressure, P-13 input 10 RTP** Turbine Impulse Pressure Equivalent 10.6 RTP**

Turbine Impulse Pressure Equivalent c.

Power Range Neutron Flux, P-8 35.0% of RTP**

35.6% of RTP**

• • RTP = RATED THERMAL POWER 2-7 Amendment No. 60, 85, 159, 217, 220, 242, 291

MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Cont.)

23.

DOM-NAF-2-P-A, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code, including Appendix C, Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code, Appendix D, Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code, and Appendix F, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code. Methodology for Specifications:

• 3.2.3.1 RCS Flow Rate, Nuclear Enthalpy Rise Hot Channel Factor

• 3.2.5 DNB Parameters 24.

EMF-2328-P-A, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, (Framatome Propietary). Methodology for Specification:

• 3.2.2.1 Heat Flux Hot Channel Factor 25.

EMF-2103-P-A, Realistic Large Break LOCA Methodology for Pressurized Water Reactors, (Framatome Proprietary). Methodology for Specification:

• 3.2.2.1 Heat Flux Hot Channel Factor 26.

ANP-10349-P-A, GALILEO Implementation in LOCA Methods, (Framatome Proprietary). Methodology for Specification:

• 3.2.2.1 Heat Flux Hot Channel Factor 27.

ANP-10342-P-A, GAIA Fuel Assembly Mechanical Design, (Framatome Proprietary). Methodology for Specification:

• 3.2.2.1 Heat Flux Hot Channel Factor 6.9.1.6.c The core operating limits shall be determined so that all 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.

6.9.1.6.d The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

Amendment No. 24, 40, 50, 69, 104, 173, 212, 215, 229, 238, 245, 249. 252, 255, 256, 279, 289, 290, 6-21 291

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION ENCLOSURE 3 (NON-PROPRIETARY)

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 290 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-49 DOMINION ENERGY NUCLEAR CONNECTICUT, INC., ET AL MILLSTONE POWER STATION, UNIT NO. 3 DOCKET NO. 50-423 Proprietary information pursuant to Section 2.390 of Title 10 of the Code of Federal Regulations has been redacted from this document.

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

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 291 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-49 DOMINION ENERGY NUCLEAR CONNECTICUT, INC., ET AL MILLSTONE POWER STATION, UNIT NO. 3 DOCKET NO. 50-423

1.0 INTRODUCTION

By letter dated October 30, 2023 (Reference 1), as supplemented by letter dated September 16, 2024 (Reference 2), Dominion Energy Nuclear Connecticut, Inc. (DENC, the licensee),

submitted a license amendment request (LAR) to revise the technical specifications (TS) for Millstone Power Station, Unit 3 (MPS3), to support the implementation of Framatome GAIA fuel with M5TM1, which is currently scheduled for insertion into the MPS3 reactor during the spring 2025 refueling outage. Specifically, the TS changes include updating the reactor core safety limits (TS 2.1.1.1), reducing the Reactor Trip System Instrumentation Trip Setpoint for the P-8 Interlock (TS Table 2.2-1, Item 18.c), and adding to the list of approved methodologies for the Core Operating Limits Report (TS 6.9.1.6.b). Additionally, the licensee requests approval of mixed-core departure from nucleate boiling (DNB) penalties for application to retained DNB margin and the use of design basis limits for a fission product barrier (DBLFPBs) associated with MPS3 specific application of methods needed to support GAIA fuel implementation.

The NRC staff identified the need for a regulatory audit to examine DENCs non-docketed information with the intent to gain understanding, to verify information, or to identify information that will require docketing to support the basis of the licensing or regulatory decision.

By letter dated April 4, 2024 (Reference 31), NRC issued an audit plan, which provided the list of requested information, documents, and other details pertaining to the audit. By letter dated September 3, 2024 (Reference 32), the NRC staff issued the Audit Summary.

The supplemental letter dated September 16, 2024, provided additional information that clarified the application, 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 (FR) on January 2, 2024 (89 FR 108).

1 M5 is a trademark or registered trademark of Framatome or its affiliates, in the United States of America or other countries.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION

2.0 REGULATORY EVALUATION

The NRC staff considered the following regulatory requirements and guidance during its review of the LAR.

Regulatory Requirements The regulations under 10 CFR 50.36, "Technical specifications," provide regulatory requirements related to the content of TSs. Section 50.36(b) of 10 CFR requires that each license authorizing the operation of a facility will include TSs and that the TSs will be derived from the safety analysis. Section 50.36(c) of 10 CFR specifies the categories that are to be included in the TSs including (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs);

(4) design features; and (5) administrative controls. Of relevance to this review, The regulation under 10 CFR 50.36(c)(1)(i)(A) and 10 CFR 50.36(c)(1)(ii)(A), Safety limits, limiting safety system settings and limiting control settings, states, in part, that:

[s]afety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity. If any safety limit is exceeded, the reactor must be shut down. Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions. Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded.

The regulation under 10 CFR 50.36(c)(5), Administrative controls, states, in part, that

[a]dministrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure the operation of the facility in a safe manner.

TSs are derived from a plants safety analysis report, which contains the plants design. As described in Section 3.1 of its updated final safety analysis report (FSAR) (Reference 3), the design of MPS3 satisfies and complies with the design criteria in Appendix A, General Design Criteria [GDC] for Nuclear Power Plants, to 10 CFR Part 50, with exceptions as noted therein.

The staffs review of the LAR considered GDC 10 particularly relevant to the GAIA fuel mechanical design:

GDC 10, Reactor Design, in 10 CFR 50 Appendix A requires that the reactor core and associated coolant, control, and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences (AOOs). SAFDLs ensure fuel is not damaged (i.e., fuel rods do not fail, fuel system dimensions remain within operational tolerances, and functional capabilities are not reduced below those assumed in the safety analysis). GDC 10 provides assurance that the integrity of the fuel and

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION cladding will be maintained, thus preventing the potential for release of fission products during normal operation or AOOs.

GDC 28, Reactivity Limits, in 10 CFR 50 Appendix A requires reactivity control systems be designed with appropriate limits to assure that the effects of the postulated reactivity accidents, including a rod ejection, can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. Compliance with GDC 28 provides assurance that the integrity of the reactor coolant pressure boundary and core coolability will be maintained.

Regulatory Guidance The NRC staff relied on the following section of NUREG-0800, Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [light water reactor]

Edition in its review of this LAR. Review guidance for satisfying GDC 10 and other regulatory requirements is provided in Chapter 4 and 15 of the SRP.

SRP Section 4.2 (Fuel System Design) describes known fuel damage criteria and acceptable means for their evaluation. SRP section 4.2 further describes the objective of a fuel system safety review as providing assurance that (1) the fuel system is not damaged as a result of normal operation and AOOs, (2) fuel system damage is never so severe as to prevent control rod insertion when it is required, (3) the number of fuel rod failures is not underestimated for postulated accidents, and (4) coolability is always maintained (Reference 4).

SRP Section 4.3 (Nuclear Design) establishes review criteria for core power distribution, reactivity coefficients, and reactivity control requirements (Reference 5).

SRP Section 4.4 (Thermal and Hydraulic Design) provides specific thermal-hydraulic review criteria for the core and reactor coolant system (Reference 6).

SRP Section 15.4.8 (Spectrum of Rod Ejection Accidents) discusses the postulated control rod ejection accident (REA) and associated criteria for evaluation of reactor coolant pressure boundary damage and cooling flow impairment (Reference 7).

NRC Generic Letter (GL) 88-16, Removal of Cycle-Specific Parameter Limits from Technical Specifications, states, in part, that it is acceptable for licensees to control reactor physics parameter limits by specifying an NRC-approved calculation methodology. Such parameter limits may be removed from the TS and placed in a cycle-specific COLR [Core Operating Limits Report], which is required to be submitted to the NRC every operating cycle or each time it is revised. (Reference 8)

Regulatory Guide (RG) 1.236, Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents, (Reference 9) provides guidance for the methods and procedures that the NRC considers acceptable for use in the pressurized water reactor (PWR) rod ejection and boiling water reactor rod drop analysis. It describes analytical limits for analyzing the short-term reactivity insertion and demonstrating compliance with GDC

28. It also defines fuel cladding failure thresholds to support radiological consequence

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION assessments.

The licensee proposed to apply the RG 1.236 guidance to demonstrate the relevant requirements of GDC 28 and Sections 4.2, 4.3, and 15.4.8 were met. The NRC staffs evaluation of the licensees compliance with the applicable regulatory guidance and requirements is discussed in Section 3.0 below.

3.0 TECHNICAL EVALUATION

3.1 Description of Proposed Changes The proposed TS changes include updating the Reactor Core Safety Limits (RCSLs) in TS 2.1.1.1, the Power Range Neutron Flux, Permissive 8 (P-8) setpoint in TS Table 2.2-1, Item 18.c; and the list of approved methodologies for the Core Operating Limits Report (COLR) in TS 6.9.1.6.b.

The licensee also requests NRC approval of the following items:

the use of DBLFPBs associated with MPS3-specific application of the following methodologies:

(1) DOM-NAF-2-P-A, Appendix F, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF [critical heat flux] Correlations in the Dominion Energy VIPRE-D Computer Code, (2) VEP-NE-2-A, Statistical DNBR Evaluation Methodology, (3) ANP-10338-P-A, AREA-ARCADIA Rod Ejection Accident, mixed-core penalties for application to DNBR analysis results of MPS3 cores containing both Framatome GAIA fuel and the resident Westinghouse fuel, The request related to DOM-NAF-2-P-A and VEP-NE-2-A would allow the licensee to perform licensing basis DNBR calculations for GAIA fuel in MPS3 cores. The request related to ANP-10338-P-A would permit the licensee implementation of the MPS3 REA analysis for GAIA fuel. The transition core DNBR penalties address the thermal-hydraulic effects resulting from physical differences between the resident Westinghouse fuel and prospective Framatome GAIA fuel.

The licensee stated that the proposed changes are needed to support the use of Framatome GAIA fuel with M5 cladding at MPS3. DENC and Framatome have entered into an agreement to implement the GAIA fuel product at MPS3. The first reload batch of GAIA fuel assemblies are planned for insertion in MPS3 Cycle 24, which is currently scheduled to begin operation in the spring 2025. The first MPS3 core full of GAIA product is expected in Cycle 26.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION DOM-NAF-2-P-A and VEP-NE-2-A Application to GAIA Fuel in MPS3 Cores The licensee previously received approval for the DNBR design basis for the resident Westinghouse fuel product using NRC-approved Fleet Report DOM-NAF-2-P-A for MPS3 (Reference 10), in which Appendix C describes the qualification of the WRB-2M Critical Heat Flux (CHF) correlation with the VIPRE-D code and includes the WRB-2M Deterministic Design Limit (DDL). Appendix D of the report includes the qualification of the ABB-NV and WLOP CHF correlations with the VIPRE-D code and includes the DDLs for both CHF correlations.

To obtain the statistical design limits (SDLs) applicable to the VIPRE-D/WRB-2M and VIPRE-D/ABB-NV code/correlation pairs for the resident Westinghouse fuel in MPS3, DOM-NAF-2-P-A was used along with the NRC-approved Fleet Report VEP-NE-2-A (Reference 11) which describes Dominion Energys statistical DNBR methodology.

The licensee submitted Appendix F to Fleet Report DOM-NAF-2-P-A for NRC approval on December 19, 2022 (Reference 12) as supplemented by letters dated April 6 and July 26, 2023 (Reference 13 and 14, respectively). The submittal included the data evaluations qualifying the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs and developed the corresponding generic DDLs. Appendix F to Fleet Report DOM-NAF-2-P-A was approved by NRC staff on December 20, 2023 (Reference 15).

The proposed changes to the MPS3 TSs associated with DOM-NAF-2-P-A and VEP-NE-2-A application are shown below with bold text, and the deletions are shown in strikethrough.

TS 2.1.1.1 Reactor Core Safety Limits The LAR would revise TS 2.1.1.1 to add the VIPRE-D/ORFEO-GAIA DDL of 1.13. ORFEO-GAIA is the primary CHF correlation used in DNBR analysis of Framatome GAIA fuel. The TS would be revised as follows:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained greater than or equal to 1.14 for the WRB-2M DNB correlation, or greater than or equal to 1.13 for the ORFEO-GAIA DNB correlation.

TS 6.9.1.6.b Core Operating Limits Report The change to the TS 6.9.1.6.b COLR reference list would add DOM-NAF-2-P-A, Appendix F, which describes ORFEO-GAIA and ORFEO-NMGRID correlation application with VIPRE-D. TS 6.9.1.6.b requires that the cycle-specific COLR contain the complete identification for each of the TS referenced topical reports used (i.e., report number, title, revision, date, and any supplements); therefore, only the report number and title of the applicable topical reports, along with the associated TS parameter, is provided in the TS 6.9.1.6.b list. Specifically, the licensee proposed addition of DOM-NAF-2-P-A, Appendix F which describes ORFEO-GAIA and ORFEO-NMGRID correlation application with VIPRE-D to the TS 6.9.1.6.b COLR reference list.

The existing item 23 in TS 6.9.1.6.b would be revised as shown in bold text below:

23. DOM-NAF-2-P-A, Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code, including Appendix C, Qualification of the Westinghouse WRB-2M CHF Correlation in the Dominion VIPRE-D Computer Code, and

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Appendix D, Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code, and Appendix F, "Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code. Methodology for Specifications:

3.2.3.1 RCS Flow Rate, Nuclear Enthalpy Rise Hot Channel Factor 3.2.5 DNB Parameters DNBR Penalties Supporting the MPS3 Transition to GAIA Fuel The licensee requests NRC approval to apply a transition core penalty (TCP) of 2.4% for application to DNBR analysis results calculated by the VIPRE-D/ORFEO-GAIA code/correlation pair during the first and second transition cycles and a TCP of 2.7% using the VIPRE-D/ORFEO-NMGRID code/correlation pair above the GAIA first mixing vane spacer grid. As stated by the licensee, the TCPs address the thermal-hydraulics effects resulting from the physical differences between the resident Westinghouse fuel and prospective Framatome GAIA fuel. During the transition cycles, no TCP will be applied to DNBR analysis results calculated using the Westinghouse CHF correlations or below the GAIA first mixing vane grid.

TS Table 2.2-1, Item 18-C - Power Range Neutron Flux, P-8 The LAR would reduce the P-8 trip setpoint2 and allowable value to meet GAIA DNBR safety limits. The proposed changes to the applicable TS are shown below with bold text, and the deletions are shown in strikethrough. TS Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints, Item 18.c would be revised as follows:

FUNCTIONAL UNIT NOMINAL TRIP SETPOINT ALLOWABLE VALUE Power Range Neutron Flux, P-8 35% 50.0% of RTP**

35.6% 50.6% of RTP**

• • RTP = RATED THERMAL POWER ANP-10338-P-A Application to GAIA Fuel in MPS3 Cores The licensee requests NRC approval of the DBLFPBs associated with MPS3-specific application of Framatomes ARCADIA Rod Ejection Accident (AREA) methodology described in ANP-10338-P-A, AREA-ARCADIA Rod Ejection Accident (Reference 16). Specifically, the licensee proposed the limits provided in Table 1 of Attachment 1 to the LAR, as supplemented in Attachment 3 to the supplemental letter dated September 16, 2024 (Reference 2), as the acceptance criteria for the MPS3 AREA analysis for GAIA fuel. The licensees analysis was performed against the acceptance criteria defined in RG 1.236 (Reference 9).

2 The P-8 function is to block a single loop low reactor coolant system (RCS) flow reactor trip when three of four power range neuron flux signals are less than the permissive setpoint. When operating below the P-8 trip point, the loss of a single reactor coolant pump (RCP) will not result in a reactor trip.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION This SE documents the NRC staff's review of the licensee's proposal for revising the MPS3 TSs to support the use of Framatome GAIA fuel with M5' fuel cladding material. The licensee submitted separate LARs to address other aspects of its planned transition to Framatome GAIA fuel.3 The sections below provide the NRC staffs evaluation of the following areas:

Application of VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID SDLs DNBR Penalties Supporting the MPS3 Transition to GAIA Fuel P-8 Power Range Neutron Flux Nominal Trip Setpoint and Allowable Value MPS3-specific Application of ANP-10338-P-A for Rod Ejection Accident Technical Specification Changes 3.2 VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID SDLs for MPS3 Section 3 of Attachment 3 to the LAR described the derivation procedure of SDLs for MPS3 cores containing Framatome 17x17 GAIA fuel being analyzed with the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs.

SDLs are obtained through the application of a Monte Carlo analysis. The main steps of the SDL derivation analysis may be summarized as follows:

Nine nominal statepoints covering the full range of normal operation and anticipated transient conditions were selected for both the ORFEO-GAIA and ORFEO-NMGRID.

The selected nominal statepoints are listed in Tables 3.6-1 and 3.6-2 of Attachment 3 to the LAR.

Two thousand random statepoints are generated for each of the nine nominal statepoints for each CHF correlation.

To calculate minimum DNBR (MDNBR) for each statepoint, the random statepoints are used as the input to VIPRE-D code.

Each MDNBR is randomized by a code/correlation uncertainty factor (Reference 11) using the upper 95% confidence limit on the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pair measured-to-predicted (M/P) CHF ratio standard deviation (Reference 12).

Total DNBR standard deviation (stotal) is then acquired by increasing the standard deviation of the resultant randomized DNBR by a small sample correction factor to obtain a 95% upper confidence limit and is then combined Root-Sum-Square with code and model uncertainties.

3 Other licensing actions supporting the use of GAIA fuel at MPS3 submitted for NRC approval include the application of Framatome loss-of-coolant accident methodologies (Reference 17) and relevant changes to the mechanical, thermal-mechanical, and thermal-hydraulic design features of the reactor core fuel assemblies associated with the transition to GAIA fuel at MPS3 (Reference 18).

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION SDL is then calculated based on Equation 1 in accordance with Reference 11:

Equation 1 where 1.645 multiplier is the z-value for the one-sided 95% probability of a normal distribution. This SDL provides peak fuel rod DNB protection at greater than 95/95.

A peak pin DNBR limit of 1.243 for VIPRE-D/ORFEO-GAIA and 1.261 for VIPRE-D/ORFEO-NMGRID with at least 95% probability at a 95% confidence level is obtained by applying total DNBR standard deviation in Equation 1.

The total DNBR standard deviation was then used to obtain 99.9% DNB protection in the full core DNBR limit of 1.251 for VIPRE-D/ORFEO-GAIA and 1.298 for VIPRE-D/ORFEO-NMGRID.

According to the LAR supplement (Reference 2), for MPS3 Framatome 17x17 GAIA Fuel, the SDL values are then rounded up and set to 1.26 for the VIPRE-D/ORFEO-GAIA code/correlation pair and 1.31 for the VIPRE-D/ORFEO-NMGRID code/correlation pair as seen in Table 1.

The licensee requests NRC approval of the Statistical Design Limits (SDLs) documented in to the subject LAR per 10 CFR 50.59(c)(2)(vii) as they constitute DBLFPBs.

Table 1: DNBR Limits for ORFEO-GAIA and ORFEO-NMGRID ORFEO-GAIA DDL 1.13 SDL 1.26 SALDET 1.45 SALSTAT 1.45 ORFEO-NMGRID DDL 1.21 SDL 1.31 SALDET 1.45 SALSTAT 1.45 Per information provided in supplement to the LAR (Reference 2), the safety analysis limits for deterministic and statistical DNB analyses (SALDET and SALSTAT, respectively) for ORFEO-GAIA and ORFEO-NMGRID are self-imposed, licensee-controlled limits and are set to a DNBR value that is higher than the design limits and provides margin to the design limits.

The retained DNBR margin is then calculated based on the difference between the safety analysis (self-imposed) limit (SAL) and the design limit (the DBLFPB) as shown in Equation 2.

Equation 2

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The calculated available retained DNBR margins are listed in Tables 2 and 3.

Table 2: DNBR Limits and Retained DNBR Margin for Deterministic DNB Applications DETERMINISTIC DNB APPLICATIONS DNB CORRELATION DDL SALDET RETAINED DNBR MARGIN

[%]

ORFEO-GAIA 1.13 1.45 22.0 ORFEO-NMGRID 1.21 1.45 16.5 Table 3: DNBR Limits and Retained DNBR Margin for Statistical DNB Applications STATISTICAL DNB APPLICATIONS DNB CORRELATION SDL SALSTAT RETAINED DNBR MARGIN

[%]

ORFEO-GAIA 1.26 1.45 13.1 ORFEO-NMGRID 1.31 1.45 9.6 The values of the retained DNBR margin listed above may be used to offset generic DNBR penalties for instance the transition core penalty for mixed cores.

The NRC staff considers the calculated SDL values for VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs for Framatome GAIA fuel at MPS3 to be acceptable as the SDL values were correctly calculated based on approved methods in Appendix F of Fleet Report DOM-NAF-2-P-A for the Framatome GAIA fuel which was approved by the NRC staff on December 20, 2023 (Reference 15).

3.3 DNBR Penalties Supporting the MPS3 Transition to GAIA Fuel To investigate the thermal hydraulic effects from the physical differences between the resident Westinghouse fuel and the prospective Framatome GAIA fuel in mixed-core configurations, the licensee used the VIPRE-D code along with the applicable CHF correlations for each fuel type according to the USNRC-approved DOM-NAF-2-P-A methodology in support of mixed-core analysis.

The licensee requests NRC approval for a TCP of 2.4% for DNBR analysis calculated using the VIPRE-D/ORFEO-GAIA code/correlation pair and a penalty of 2.7% using the VIPRE-D/ORFEO-NMGRID code/correlation pair above the GAIA first mixing vane spacer grid.

For the above-mentioned TCP calculations, a mixed-core configuration that consisted of a single GAIA assembly placed in a core full of the resident Westinghouse fuel was used.

Therefore, the penalties are applicable to both the first and second fuel transition cycles and may be removed once a full core of GAIA is attained.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION For the resident Westinghouse fuel, for all applicable code correlation pairs, the mixed-core CHF calculations presented improved DNBR performance in comparison with full-core CHF calculations of the resident fuel. Similarly, it was noted by the licensee that CHF performance was improved below the GAIA first mixing spacer grid. Therefore, the licensee did not apply any TCP to DNBR analysis results for the resident Westinghouse fuel or below the GAIA first mixing vane grid in GAIA transition cycles.

In Section 3.2 of Attachment 1 to the LAR, the licensee described their mixed-core DNBR analysis approach, the subchannel code modelling considerations, and mixed-core analysis performed for the MPS3 GAIA fuel transition. Moreover, the LAR supplement in Reference 2, provided additional information that clarified DENCs mixed-core DNBR evaluation supporting the GAIA fuel transition.

Methodology In its LAR, as supplemented (Reference 1 and 2), the licensee determined the DNBR penalties from a subchannel code analysis of the mixed-core configuration using VIPRE-D. Two sets of VIPRE-D models were developed in accordance with Dominion Energys NRC-approved DOM-NAF-2-P-A methodology to perform the MPS3 GAIA fuel transition mixed-core evaluation:

(1) Framatome GAIA fuel model pair:

a. Full-core VIPRE-D model of GAIA

b. Mixed-core VIPRE-D model; one GAIA fuel assembly is placed in a core of resident Westinghouse RFA-2 fuel with the single GAIA assembly modeled as the power limiting bundle (PLB). Minimum DNBR (MDNBR) is calculated within the PLB.

(2) Westinghouse RFA-2 fuel model pair:

a. Full-core VIPRE-D model of RFA-2

b. Mixed-core VIPRE-D model; one RFA-2 fuel assembly is placed in a core of GAIA fuel with the single RFA-2 assembly modeled as the PLB. MDNBR is calculated within the PLB.

In Section 3.2.2 of Attachment 1 to the LAR, the licensee stated the importance of accurate modeling of the flow redistribution between adjacent hydraulically dissimilar fuel assemblies which induces a lateral flow component to the more dominant axial fluid velocity. This lateral flow component, which is commonly called fuel assembly crossflow, is a key requirement of the Dominion Energy mixed-core DNB evaluation.

To ensure accurate predictions of crossflow from higher pressure drop fuel regions to lower pressure drop fuel regions, in addition to the considerations listed in NRC-approved DOM-NAF-2-P-A topical report, other modeling considerations such as Radial Nodalization, Axial Nodalization, Axial Hydraulic Form Losses, and Critical Heat Flux correlations were included by the licensee in the VIPRE-D mixed core models.

The TCP values are then calculated based on Equation 3:

Equation 3

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION To obtain the MDNBRFull-Core and MDNBRTransition-Core values in Equation 3, the licensee used a pair of VIPRE-D models as explained above to run two DNB cases at the same statepoint conditions (power, RCS pressure and temperature, RCS flow rate, radial peaking, and axial shape). This procedure was repeated for a range of statepoints for GAIA fuel transition which included representative core thermal limits (CTLs) and FSAR Chapter 15 events.

The licensee proposed a TCP value of 2.4% for application to DNBR analysis results calculated using the VIPRE-D/ORFEO-GAIA code/correlation pair, and a TCP value of 2.7% when using the VIPRE-D/ORFEO-NMGRID code/correlation pair above the GAIA first mixing vane spacer grid.

The NRC staff considers these TCP values to be acceptable as they were obtained through approved methodologies and correctly applied to the Framatome GAIA fuel. The NRC staffs approval of the TCP values proposed for the Framatome GAIA fuel is based on its prior review and approval received for Appendix F to the Fleet Report DOM-NAF-2-P-A (Reference 15).

3.4 Analytical Limit for the Proposed P-8 Power Range Neutron Flux Nominal Trip Setpoint and Allowable Value Section 3.3 of Attachment 1 to the LAR (Reference 1) discussed proposed changes to the P-8 nominal trip setpoint (NTS) and allowable value (AV) listed as item 18.c in MPS3 TS Table 2.2-1. Specifically, the proposed changes reduced the P-8 NTS from 50.0% rated thermal power (RTP) to 35% RTP and AV from 50.6% RTP to 35.6% RTP. The proposed lower power values of the P-8 trip setpoint would result in an earlier actuation of the reactor trip and limit the plant operations to more restrictive conditions. This section discusses the NRC staffs review on application of the proposed analytical limit (AL) of 45% RTP for supporting the proposed P-8 NTS and the associated AV in accordance with the GDC 10 requirements.

On July 31, 2024, the NRC staff requested additional information (RAI) from the licensee (Reference 33) to support the completion of its evaluation of the proposed LAR. In RAI-1, the staff requested the licensee, in part, to describe the methodology for establishing the P-8 AV. In its response to RAI-1 (Reference 2), the licensee stated that the P-8 analysis was performed with the NRC-approved RETRAN code at MPS3 (Reference 19) for RCS transient analyses and the VIPRE-D/ORFEO-GAIA and VIPRE-D/ORFEO-NMGRID code/correlation pairs for DNBR calculations. The NRC reviewed and accepted the proposed code/correction pairs (Reference 15) and discussed the basis of acceptance in Section 3.1 of this SE.

Section 2.3.2 of Attachment 1 to the LAR indicated that the P-8 analysis applied mixed-core (or transition core) DNBR penalties supporting the transition from MPS3s resident Westinghouse fuel product to Framatome GAIA fuel. Specifically, a TCP (equivalent to the mixed core penalty referred in Section 3.3) of 2.4% was used for application to DNBR analysis results calculated using VIPRE-D/ORFEO-GAIA code/correlation pair, and a TCP of 2.7% was used when using VIPRE-D/ORFEO-NMGRID code/correlation pair above the GAIA first mixing van space grid.

The NRC reviewed and accepted the proposed MCPs and discussed the basis of acceptance in Section 3.3 of this SE.

The licensee used a low flow condition consistent with the P-8 interlock logic (loss of an RCP) was used to establish the proposed P-8 trip setpoint. The licensees RAI-2 response clarified that the VIPRE-D DNBR calculations assumed values for key parameters that would result in a

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION minimum margin to the DNBR limits. The key parameters included: (1) the lowest anticipated core flow rate based on RETRAN for conditions with three reactor coolant pumps in operation; (2) a low RCS pressure and high RCS temperature, which were the limiting directions for DNBR analysis; and (3) radial peaking factors reflecting COLR limits. A sensitivity study was performed by the licensee for a wide range of axial skewing to identify an appropriate axial power shape and peak. A sufficiently limiting power shape was selected to bound measured MPS3 flux map data at the power levels close to the proposed P-8 setpoint. The results of the P-8 analyses demonstrated that the AL of 45% RTP was sufficient to ensure that the P-8 function would maintain the calculated DNBR within the acceptable limits.

The NRC staff reviewed the licensees P-8 analysis and found that that the licensees analysis was performed using NRC-approved methods and conservative initial conditions that bounded the plant operating ranges or TS values, resulting in minimum calculated DNBRs.

Also, the results of the P-8 analysis demonstrated that the AL of 45% RTP for the P-8 function provided significant margin to the applicable DNBR limits. Therefore, the NRC staff determined that the analysis would not result in fuel failure, and thus satisfied the GDC 10 requirements.

As discussed in MPS3 TS bases 2.2.1, the determination of the NTS and associated AV was based on the AL with consideration of the instrumentation measurement uncertainties and the drift allowance between the NTS and the AL. Therefore, the NRC staff concluded that the AL of 45% RTP was acceptable for determination of the proposed item 18.c in MPS3 TS Table 2.2-1 related to the P-8 NTS of 35% RTP and AV of less than or equal to 35.6% RTP for application of the GAIA fuel transition at MPS3.

The NRC staff reviewed the instrumentation and control portions of the LAR and noted that the description of Section 3.3, Reduction of the P-8 Power Range Neutron Flux Nominal Trip Setpoint and Allowable Value, did not make any statement about instrument uncertainty. The staff requested licensee confirmation that the associated instrument uncertainty analysis/calculation remains bounding or applicable. The staff noted that while the amount of instrument uncertainty (delta 0.6%) did not change, the operating region where the uncertainty is considered did change. Therefore, the staff requested confirmation from the licensee that they reviewed or verified the associated instrument uncertainty and the reasoning that it remains bounding or valid. The licensee responded that as part of development of its LAR, they evaluated the uncertainty calculation with respect to the P-8 change (Reference 20). This evaluation indicated that the 0.6% uncertainty is a function of the bistable, and is not a function of the setting of the bistable. Based on the licensees response, the NRC finds that instrument uncertainty remains valid and applicable at the lower power level.

3.5 Rod Ejection Accident Analysis Section 3.4 of Attachment 1 to the LAR discussed the REA analysis for MPS3 with GAIA fuel.

The ejection of a rod cluster control assembly (RCCA) (i.e., rod ejection) would result from the mechanical failure of a control rod mechanism pressure housing. The consequences of the REA include a rapid reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The REA analysis for MPS3 applied the NRC-approved rod ejection accident methodology described in topical report (TR) ANP-10338P-A (Reference 16)4. Section 3.4 of the to the LAR indicated that a limitation and condition (L&C) for use of the methodologies of ANP-10342P-A (Reference 22), GAIA Fuel Assembly Mechanical Design, was applicable to the REA application. The L&C required that the most up-to-date guidance and analytical limits be considered when demonstrating acceptable performance of GAIA fuel under reactivity-initiated accident conditions and referenced the recently issued RG 1.236 (Reference 9) as the most up-to-date guidance for the REA analysis. To satisfy the L&C, the licensee performed the MPS3 REA analysis following the guidance in RG 1.236.

This section discusses the NRC staffs review on acceptability of the REA analysis in accordance with the applicable guidance and analytical limits in RG 1.236.

3.5.1 Initial Conditions Assumed in the REA Analysis for PWR Section C.2.2.1, PWR CRE [control rod ejection] Initial Conditions, of RG 1.236 provided guidance for selection of PWR REA initial conditions. Specific guidance important to the REA analysis was provided in Regulatory Positions (RPs) C.2.2.1.1 through C.2.2.1.13.

The licensee addressed how it follows the guidance of RG 1.236 in its RAI-6 response (Reference 2) and indicated that it satisfied each of the required initial conditions (ICs) in RPs C.2.2.1.1 through C.2.2.1.13 of RG 1.236 for the MPS3 REA analysis. The NRC reviewed the licensees usage of the relevant RPs and discussed its evaluation as follows.

3.5.1.1 RP C.2.2.1.1 - Times in Life RP C.2.2.1.1 of RG 1.236 states that accident analyses consider the full range of cycle operation from beginning of cycle (BOC) to end of cycle (EOC).

The licensees RAI-6 response addressing compliance with RP C.2.2.1.1 confirmed that, following the NRC-approved methodology in Section 6.4 of ANP-10338P-A, the MPS3 REA analysis considered several times in life (TILs) and power levels. The TILs included BOC, EOC, and appropriate middle of cycle (MOC) conditions. For each case analyzed, the TIL was chosen to maximize power peaking factors and the associated power level was selected to maximize ejected rod worths to ensure a conservative analysis.

Since the licensees REA analysis was performed for TILs at different power levels from BOC, EOC, and appropriate MOC conditions with the associated maximum power peaking factors and at power levels maximizing ejected rod worths, the NRC staff determined that the analysis adequately considered the effects of TILs including the associated power peaking factor and power level effects, and thus, satisfied RP C.2.2.1.1.

3.5.1.2 RP C.2.2.1.2 - Xenon Effects on REA Analysis at Hot Zero Power, BOC Conditions RP C.2.2.1.2 stated that accident analyses at hot zero power (HZP) encompass both (1) BOC following core reload and (2) restart following recent power operation.

4 NRC staffs final safety evaluation for ANP-10338P was issued on December 20, 2017 (Reference 21)

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The licensees RAI-6 response addressing compliance with RP C.2.2.1.2 indicated that the consequences of the REA event would be more severe when skewed xenon conditions existed that increased local peaking by crediting top or bottom peak axial power distribution. For Case (2), the HZP REA analysis initiated at restart following recent power operation where xenon existed. The licensee performed an analysis with consideration of xenon effects and assumed conservative representation of xenon distribution with respect to restart following recent operation. This approach was consistent with the methods in Section 7.1.4.6 of ANP-10338P-A. ((

)).

The NRC staff found that the licensees analysis, consistent with the methods in ANP-10338P-A, adequately considered xenon effects on the REA case for BOC HZP conditions, and therefore, determined the analysis satisfied RP C.2.2.1.2.

3.5.1.3 RP C.2.2.1.3 - Initial Power Level Conditions RP C.2.2.1.3 stated that accident analyses consider the full range of power operation including intermediate power levels up to hot full-power conditions. These calculations should consider power-dependent core operating limits (e.g., control rod insertion limits, rod power peaking limits, axial and azimuthal power distribution limits). At conditions where certain core operating limits do not apply, the analysis should consider the potential for wider operating conditions resulting from xenon oscillations or plant maneuvering.

When properly justified, cycle-independent bounding evaluations that demonstrate that regions of power operation are less limiting are an acceptable analytical approach to reduce the number of cases analyzed. For example, during control rod ejection scenarios initiated from at-power conditions, credit for power-dependent insertion limits in the technical specifications may be used to demonstrate that these particular events are of less significance with respect to coolable geometry.

The licensees RAI-6 response addressing compliance with RP C.2.2.1.3 confirmed that the power levels assumed in the REA analysis covered HZP, hot full power (HFP) and intermediate power in accordance with the methods in Section 6.4 of ANP-10338P-A. Specifically, the TS core operating values, including but not limited to rod insertion limits, rod position uncertainty, and axial power distribution limits, were accounted for in the REA analysis. When conditions existed that were not covered by TS core operating limits, the methods used for the REA analysis explicitly included the potential for wider operating conditions, which were consistent with ((

)).

The licensee further indicated that the RP C.2.2.1.3 guidance for reducing number of cases analyzed was not applied in the NRC-approved ANP-10338P-A methodology. For the MPS3 REA analysis, the TIL was chosen to maximize power peaking factors and the power level was selected to maximize the rod ejection worth. The combination of the maximum ejected rod worth and high-power peaking factors would provide for the conservative analyses.

Since the MPS3 REA analysis was performed using the methods consistent with that in ANP-10338P-A, maximizing ejected rod worth and high-power peaking factors, the NRC staff

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION determined the analysis adequately considered the effects of the TILs and the initial power levels, and thus, satisfied, RP C.2.2.1.3.

3.5.1.4 RP C.2.2.1.4 - Ejected Rod Worths RP C.2.2.1.4 stated that uncontrolled worth of an ejected rod be calculated based on the following conditions: (1) the range of control rod positions allowed at a given power level and (2) additional fully or partially inserted misaligned or inoperable rod(s) if allowed. Applicants do not need to consider dropped or misaligned rods which are being recovered within technical specifications limiting conditions for operation (TS LCO) completion times.

Sufficient parametric studies should be performed to determine the worth of the limiting control rod for the allowed configurations highlighted above. The evaluation methodology should account for (1) calculation uncertainties in neutronic parameters (e.g., neutron cross sections) and (2) allowed power asymmetries.

The licensees RAI-6 response addressing compliance with RP C.2.2.1.4 indicated that, following the method discussed in Section 6.4 of ANP-10338P-A, the selection of rod to be ejected was the ((

)) for the range of allowed initial positions for the power level under consideration. The ejected rod worth calculation within the REA analysis included the established control rod position uncertainty from ((

)) as discussed in Section 7.1.4.1 of ANP-10338P-A. Cross section adjustment factors (Section 6.5.3 of ANP-10338P-A) ((

)).

The licensee also confirmed that for the rods in ((

)). Since all other rods were assumed at all-rod-out position, rod misalignment did not impact the MPS3 REA analysis.

Inoperable rods did not impact rod ejection but would impact the scram worth. The licensee determined that the impact was small and would have no consequence to the ability to reduce power following scram.

Further, the licensee clarified that the REA methodology used for MPS3 did not consider

((

)). In addition, the core biasing strategies, including ejected rod worths, used in the REA analysis were consistent with that shown in Table 7-2 of ANP-10338P-A.

Since, for the REA analysis, (1) the ((

)) was selected, which was consistent with the method discussed in Section 6.4 of ANP-10338P-A, (2) the cross section adjustment factors were applied to all other rods to reduce scram worth, which was consistent with the approach in Section 6.5.3 of ANP-10338P-A, (3) ((

)), and (4) ((

)), the NRC staff determined that the analysis used the appropriately selected ejected rod worths and power peaking factors and thus, satisfied RP C.2.2.1.4.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.5.1.5 RP C.2.2.1.5 - Burnup-Dependent and Corrosion-Dependent Factors RP C.2.2.1.5 stated that, because of burnup-dependent and corrosion-dependent factors that tend to reduce cladding failure thresholds and allowable limits on core coolability during fuel rod lifetime, the limiting initial conditions may involve locations other than the maximum uncontrolled rod worth defined in Regulatory Position C.2.2.1.4 (e.g., uncontrolled rod motion at a core location adjacent to higher burnup fuel assemblies). For this reason, a more comprehensive search for the limiting conditions may be necessary to ensure that the total number of fuel rod failures is not underestimated, and allowable limits are satisfied. Applicants needed to survey a larger population of PWR ejected rod core locations and exposure points to identify the limiting scenarios.

When properly justified, combining burnup-dependent parameters to create an artificial, composite worst time-in-life (e.g., end-of-life cladding hydrogen content combined with maximum ejected worth) is an acceptable analytical approach to reducing the number of cases analyzed.

The licensees RAI-6 response addressing compliance with RP C.2.2.1.5 indicated that the REA analysis used the same method in Section 6.4 of ANP-10338P-A by selecting the ((

)) in the analysis. Section 6.4 acknowledged that ((

)) The same approach was applied to other acceptance criteria via investigation of parameters such as ((

)) where condition existed that would have a ((

)).

The licensee indicated that a comprehensive search for the limiting condition was ((

)).

Since the MPS3 REA analysis was performed using the methods consistent with that in ANP-10338P-A, the ((

)) was selected as the rod to eject in the analysis, and the analysis demonstrated that the ((

)) for the limiting condition, the NRC staff determined that the analysis adequately considered the effects of burnup-and corrosion-dependent parameters, and thus, satisfied RP C.2.2.1.5.

3.5.1.6 RP C.2.2.1.6 - Reactivity Insertion Rate RP C.2.2.1.6 states that the reactivity insertion rate be determined from differential control rod worth curves and calculated transient rod position versus time curves.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The licensees RAI-6 response addressing compliance with RP C.2.2.1.6 clarified that the MPS3 REA analysis did not use the maximum differential rod worth or calculated transient position versus time curve to calculate the reactivity insertion rate, because the analysis used an NRC-approved 3D model code (ARTEMIS) documented in ANP-10297P-A, Revision 0 (Reference 23) and ANP-10297P-A, Supplement 1P-A, Revision 1 (Reference 24).

The code calculated the core reactivity as the rod was ejected rather than relying on a differential control rod worth.

For the reactor trip, the calculated reactivity insertion rates were based upon control rod scram position versus time provided by the MPS3 licensee (as discussed in Section 6.5.1 of ANP-10338P-A).

The NRC staff found that the determination of the reactivity insertion rates in the REA analysis was not based on the methods specified in the RG 1.236, but it was consistent with that of ANP-10297P-A, Revision 0, ANP-10297P-A, Supplement 1P-A, Revision 1, and ANP-10338P-A. Therefore, the NRC staff determined that the calculation of the reactivity insertion rates met the intent of RP C.2.2.1.6.

3.5.1.7 RP C.2.2.1.7 - Rate of Ejection RP C.2.2.1.7 stated that the rate of ejection be calculated based on the maximum pressure differential and the weight and cross-sectional area of the control rod and drive shaft, assuming no pressure barrier restriction.

The licensees RAI-6 response addressing compliance with RP C.2.2.1.7 indicated that the ejection rate used in the REA analysis was not determined based on the method of using the maximum pressure differential and weight and cross-sectional area of the control rod and drive shaft as specified in RG 1.236. The MPS3 REA analysis assumed that a fully inserted rod was ejected within 0.1 seconds, which was consistent with the approved methodology in Section 6.5 of ANP-10338P-A. The value of 0.1 seconds (in MPS3 FSAR Section 15.4.8.2, Reference 3) was also a typical value used for the acceptable REA analysis in the industry, as shown in Callaway FSAR Section 15.4.8.2 (Reference 35), and Byron and Braidwood FSAR Section 15.4.8.2 (Reference 36).

The NRC staff found that the value of 0.1 seconds for the ejection rate used in the REA analysis was not based on the methods specified in RG 1.236, but it was consistent with the value used in Section 6.5 of ANP-10338P-A and the value typically used for the acceptable REA analyses in the industry. Therefore, the NRC staff determined that the use of 0.1 seconds for the ejection rate met the intent of RP C.2.2.1.7.

3.5.1.8 RP C.2.2.1.8 - Initial Reactor System Coolant Conditions RP C.2.2.1.8 stated that the initial reactor coolant pressure, core inlet temperature, and flow rate used in the analysis be conservatively chosen, depending on the transient phenomenon being investigated. The range of values should encompass the allowable operating range and monitoring uncertainties.

The licensees RAI-6 response addressing compliance with RP C.2.2.1.8 indicated that the conditions of pressure, flow rate, and core inlet temperatures used in the REA analysis would

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION bound the operating range and monitoring uncertainty. The plant operating ranges were within the range of applicability of the thermal hydraulic methods and codes. These initial conditions were conservatively biased (including monitoring uncertainties) in accordance with the requirements in Section 7.1.4 of ANP-10338P-A to generate conservative results (with minimum margin to the applicable fuel and cladding limits).

The NRC staff found that the approach of selection of initial conditions for pressure, flow rate, and core inlet temperatures was consistent with the methodology in ANP-10338P-A. The selected initial conditions bounded the allowable operating range and monitoring uncertainties and were conservative, resulting in minimum margin to the applicable fuel and cladding limits.

Therefore, the NRC staff determined that the approach adequately considered the effects of the key initial reactor coolant system conditions on the REA response, and thus, satisfied RP C.2.2.1.8.

3.5.1.9 RP C.2.2.1.9 - Fuel Thermal Properties RP C.2.2.1.9 stated that fuel thermal properties (e.g., fuel-clad gap thermal conductivity, fuel thermal conductivity) cover the full range over the fuel rods lifetime and be conservatively selected based on the transient phenomenon being investigated. Time-in-life specific fuel properties could be used for a given burnup-specific state-point analysis.

The licensees RAI-6 response addressing compliance with RP C.2.2.1.9 confirmed that the REA analysis used fuel properties that were within the range of applicability of the ARTEMIS fuel rod model. The fuel properties (gap conductivity, fuel conductivity, and heat capacity) were considered and conservatively biased in accordance with the requirements in Section 7.1.4 of ANP-10338P-A to generate conservative results. TIL specific properties were considered for the burnup specific TIL analyzed.

The NRC staff found that the approach of selection of initial conditions for the fuel thermal properties was consistent with the methodology in Section 7.1.4 of ANP-10338P-A and was conservative, resulting in minimum margin to the applicable fuel and cladding limits. Therefore, the NRC staff determined that the approach adequately considered the effects of the key initial fuel thermal properties on the REA response, and thus, satisfied RP C.2.2.1.9.

3.5.1.10 RP C.2.2.1.10 - Moderator Reactivity Feedback RP C.2.2.1.10 stated that the moderator reactivity feedback resulting from voids, coolant pressure changes, and coolant temperature changes be calculated based on the various assumed conditions of the fuel and moderator using standard transport and diffusion theory codes. If boric acid shim was used in the moderator, the highest boron concentration corresponding to the initial reactor state should be assumed. If applicable, the range of values should encompass the allowable operating range (i.e., technical specifications in the core operating limits report) and any applicable analytical uncertainties.

The licensees RAI-6 response addressing compliance with RP C.2.2.1.10 confirmed that, in the REA analysis, the ARCADIA 3D code (ARTEMIS) was used to perform an explicit moderator temperature coefficient (MTC) calculation that calculated the effects of voids, coolant pressure, and temperature changes based on the conditions of the transient. The REA analysis assumed MTCs that encompassed the operating range. The MTC analytical uncertainties and biases applied to the moderator reactivity feedback were consistent with the methods in Section 7.1.4.4

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION of ANP-10338P-A. The REA analysis also assumed boron concentration that was consistent with the TIL and power level analyzed. The ((

)) applied in the calculation.

The NRC staff found that the MTC calculation was performed with the NRC-approved ARTEMIS code and MTCs assumed in the REA analysis bounded the operating range. Therefore, the NRC staff determined that the MTC calculation for the REA analysis adequately considered the effects of the voids, coolant pressure and temperature changes, and boron concentration, and thus, satisfied RP C.2.2.1.10.

3.5.1.11 RP C.2.2.1.11 - Doppler Reactivity Feedback RP C.2.2.1.11 stated that calculations of the Doppler reactivity feedback be based on and compared with available experimental data. Since the Doppler feedback reflected the change in reactivity as a function of fuel temperature, uncertainties in predicting the coefficient, as well as in predicting fuel temperatures at different power levels, should be reflected by conservative application of Doppler feedback.

The licensees RAI-6 response addressing compliance with RP C.2.2.1.11 indicated that Doppler reactivity feedback was considered during the evaluation of the ARCADIA TR described in Section 11.2.2 of ANP-10297P-A, Revision 0 (Reference 23) and Section 3.5.3 of NRC SE for ANP-10297P-A Supplement 1, Revision 1 (Reference 24). The benchmarks demonstrated that the ARCADIA 3D code (ARTEMIS) adequately predicted the Doppler temperature coefficient (DTC). The uncertainty applied to the Doppler temperature feedback followed the methods discussed in Section 7.1.4.3 and Table 7-2 of ANP-10338P-A.

Since the REA analysis was performed using the NRC-approved ARCADIA method with inclusion of the acceptable Doppler temperature feedback uncertainty in ANP-10338P-A, the NRC staff concluded that the REA analysis adequately considered the effects of the fuel temperature change and the Doppler temperature feedback uncertainty, and thus satisfied RP C.2.2.1.11.

3.5.1.12 RP C.2.2.1.12 - Control Rod Reactivity Insertion during Trip RP C.2.2.1.12 stated that control rod reactivity insertion during trip versus time should be obtained by combining the differential rod worth curve with a rod velocity curve based on maximum design limit values for scram insertion times. Alternatively, reactivity may be calculated using control rod velocity during trip based on maximum design limit values for scram insertion times. Any loss of available scram reactivity resulting from allowable rod insertion should be quantified.

The licensees RAI-6 response addressing compliance with RP C.2.2.1.12 indicated that a 3D code (ARTEMIS) of the ARCADIA method in ANP-10338P-A was used in the REA analysis for calculation of the negative reactivity insertion during the trip based on a scram curve consistent with the plant licensing basis. Rod position with respect to time was used in the core simulation (Section 6.5.1 of ANP-10338P-A).

The scram curve input to the MPS3 REA analysis was given as a function of control rod distance versus time (control rod velocity). The TS maximum allowed rod time from the fully withdrawn position to dashpot entry was modelled in the curve, providing the maximum scram

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION time. The loss of available scram reactivity due to partial inserted control was also explicitly handled by the 3D neutronic code (ARTEMIS).

The NRC staff found that the negative reactivity insertion during the trip was calculated using the 3D ARTEMIS code with input of a scram curve consistent with the plant licensing basis, which would result in the slowest addition of negative reactivity feedback to the core after the reactor trip and was conservative. Therefore, the NRC staff concluded that the calculation used for the control rod reactivity insertion during the trip satisfied RP C.2.2.1.12.

3.5.1.13 RP C.2.2.1.13 - Reactor Trip Delay Time RP C.2.2.1.13 stated that the reactor trip delay time, or the amount of time that elapsed between the instant the sensed parameter (e.g., pressure, neutron flux) reached the level for which protective action was required and the onset of negative reactivity insertion, be based on maximum values of the following: (1) time required for the instrument channel to produce a signal, (2) time for the trip breaker to open, (3) time for the control rod motion to initiate, and (4) time required before control rods entered the core if the tips lied outside the core. The response of the reactor protection system should allow for inoperable or out-of-service components and single failures.

The licensees RAI-6 response addressing compliance with RP C.2.2.1.13 indicated that, in accordance with the methodology in Section 6.5.1 of ANP-10338P-A, the excore detector model, the signal processing (based on the excore flux signal), and control scram model were implemented in the 3D code (ARTEMIS) trip function. The reactor trip delay times were used as input to the MPS3 REA analysis. Specifically, the input included the (1) time required for instrumentation channel to provide a signal, (2) time for the trip breaker to open, and (3) time for the control rod insertion to initiate. The times input reflected the acceptance criteria (limiting values) contained in the MPS3 licensee-controlled technical requirements manual (TRM). Also, the licensee confirmed that for the required Item (4), the time required before control rods entering the core for cases with the control rod tips lying outside the core, was handled by the neutronic code (ARTEMIS in ANP-10338P-A), which tracked the position of the control rods explicitly.

Since the MPS3 REA analysis used the limiting values specified in the licensees TRM for reactor trip delay times required in items 1 through 3 and used the 3D code simulation (ARTEMIS) consistent with the method in ANP-10338P-A for treatment of delay time required in Item 4, the NRC staff concluded that the approach used for the reactor trip delay times satisfied RP C.2.2.1.13.

Based on the evaluations discussed in above Sections 3.5.1.1 through 3.5.1.13, the NRC staff concluded that the MPS3 REA analysis satisfied RP Section C.2.2.1 of RG 1.236 for initial conditions assumed in the REA analysis.

3.5.2 Methodologies and Acceptance Criteria for the Rod Ejection Analysis Section 3.4 of Attachment 1 to the LAR stated that the REA analysis for GAIA fuel at MPS3 was performed using the methodologies in topical report (TR) ANP-10338-P-A (Reference 16 and

21) and following the guidance in RG 1.236 (Reference 9). The licensee proposed the limits in Table 1 of Attachment 1 to the LAR, as supplemented, as the acceptance criteria for the MPS3

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION REA analysis. The NRC staff reviewed the proposed acceptance criteria and discussed its evaluation below.

3.5.2.1 Acceptance Criterion 1 (AC1) - High-Temperature Cladding Failure Threshold The licensee proposed in Table 1 of Attachment 1 to the LAR the limits in Figure 1 of RG 1.236 as the acceptance criteria for the high-temperature cladding failure threshold. RP C.3.1 of RG 1.236 restricted the applicability of the Figure 1 limits to events with prompt critical excursions (i.e., ejected rod worth greater than or equal to $1.00). In RAI-3, the staff requested the licensee to provide a discussion to address the compliance with RP C.3.1 for use of the Figure 1 limit. In its response to RAI-3 (Reference 2), the licensee confirmed that the proposed Figure 1 limits would be applied to the MPS3 REA analysis for events with prompt critical response, when the ejected rod worth was greater than or equal to $1.00. The NRC staff found that the licensees use of the Figure 1 limit met the required restriction of RP C.3.1 of RG 1.236, and thus, determined that AC1 (the proposed Figure 1 limit) was acceptable for use in the MPS3 REA.

3.5.2.2 AC2 - DNBR Design Limit of 1.12 for COBRA-FLX/ORFEO-GAIA The licensee proposed in Table 1 of Attachment 1 to the LAR a limit of 1.12 for the COBRA-FLX/ORFEO-GAIA code/correlation pair as the acceptance criterion for the DNBR design limit used in the REA analysis. The NRC staff found that the proposed limit of 1.12 was discussed in TR ANP-10341P-A, Revision 0, The ORFEO-GAIA and ORFEO-NMGRID Critical Heat Flux Correlations (Reference 25) and was previously reviewed and approved by NRC.

Specifically, Section 4.2 of the NRC SE for ANP-10341P-A, Revision 0 stated that ORFEO-GAIA is approved for use in predicting the critical heat flux (CHF) downstream of GAIA and intermediate GAIA mixing grid in GAIA fuel. This prediction must be made in the subchannel code COBRA-FLX with the modeling options as specified in Table 5-1 of the TR with the design limit of 1.12 over the application domain specified in in Table 2-2 of the initial submittal of the TR. The approved design limit contains a bias of 0.01 which NRC staff believed was necessary to account for the variations between the test fuel assembly and production fuel assembly which will used in the in the reactor. Since the proposed limit of 1.12 for COBRA-FLX/ORFEO-GAIA code/correction pair was previously approved by NRC as the DNBR design limit for use in the analysis of transients and accidents, including the REA event, the NRC staff determined that the DNBR design limit of 1.12 was acceptable for the REA analysis at MPS3 with GAIA fuel.

3.5.2.3 AC3 - PCMI limits for Peak Average Fuel Enthalpy Rise The licensee proposed in Table 1 of Attachment 1 to the LAR the limits in Figures 2 and 4 of RG 1.236 as the acceptance criteria for the pellet-cladding mechanical interaction (PCMI) threshold.

RP C.3.2 of RG 1.236 restricted the applicability of Figures 2 and 4 to recrystallized annealed (RXA) cladding type. In addressing compliance with the restriction, the licensee indicated in its RAI-4 response that the GAIA fuel design used M5 cladding. M5 cladding was a RXA cladding type as shown in the NRC-approved TR, BAW-10227P-A, Revision 1, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel (Reference 26). Specifically, Page A-3 of the TR stated that all components, (cladding, guide tubes, and grids) are specified in the fuel recrystallized (RXA) condition. Since the GAIA fuel design used M5 cladding, and NRC staff confirmed from the NRC-approved TR BAW-10227P-A Revision 1 that M5 cladding was a RXA cladding type, the NRC staff determined that the restriction in RP C.3.2 of RG 1.236 for

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION use of Figures 2 and 4 limits was met, and thus, concluded that the proposed PCMI limits were acceptable for use in the MPS3 REA analysis.

3.5.2.4 AC4 - Fuel Centerline Melt Temperature Limits The licensee proposed in Table 1 of Attachment 1 to the LAR the limits, less than 4754°F, decreasing linearly by 13.7°F per 10,000 MWD/MTU of burnup; rim melt is precluded, as the acceptance criteria for the fuel centerline melt (FCM) temperature limits.

In its RAI-5 response, the licensee provided the calculation of the FCM temperature limits for UO2 as a function of fuel burnup. The calculation was based on three components, including (1) the best estimate (BE) FCM temperature relationship, (2) a melt temperature uncertainty, and (3) a temperature prediction uncertainty. The result (Eq. 5) in Section 2 of the RAI-5 response showed that the FCM temperature limits were less than 4754.7oF (rounded down to 4754oF), decreasing linearly 12.8 oF per 10,000 MWd/MTU of burnup. The burnup dependent term of 12.8 oF was different from the value of 13.7°F in row 5 in Table 1 of Attachment 1 to the LAR. The licensee corrected the FCM limit burnup dependent term from 13.7°F to 12.8 oF as shown in Table 1 of Attachment 3 to the RAI response (Reference 2).

Based on the licensees RAI-5 response, the NRC staff found that the first two of the three components used to calculate the acceptance criteria for FCM temperature limits were included in the following licensing documents: (1) the BE FCM temperature limits (in Celsius) in Section 4.2.4.7.1 of TR ANP-10339P-A, Revision 0 (Reference 27) and (2) the melt temperature

((

)) in Section 6.8.3 of TR ANP-10338P-A, Revision 0 (Reference 16) and Section 4.2.4.7.1 of TR ANP-10339P-A, Revision 0. Since the referenced TRs were previously approved by the NRC, the NRC staff determined that the first two components were acceptable for determination of the acceptance criteria for FCM temperature limits.

The third component, the temperature prediction uncertainty, was originally included in Section 6.8.3 of TR ANP-10338P-A, Revision 0 (Reference 16) as ((

)), which was based on the database used with ((

)) The NRC previously approved (pages 74-75 of Reference 34) the analysis for Callaway, which was the same as MPS3, a Westinghouse 4-loop plant. Therefore, the NRC staff determined that the third component was acceptable for determination of the acceptance criteria for FCM temperature limits.

Because the calculation of the revised FCM temperature limits was based on the NRC-approved methods and uncertainties, the NRC staff concluded that the proposed revised limits shown in Table 1 of Attachment 3 to the RAI response (Reference 2) were acceptable for the MPS3 REA analysis.

3.5.2.5 AC5 - Peak Radial Average Enthalpy Limit

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The licensee proposed in Table 1 of Attachment 1 to the LAR the limits, less than 230 cal/g as the acceptance criteria for the peak radial average enthalpy. The NRC staff found that the proposed limit of 230 cal/g was consistent with RP C.6.a of RG 1.236, stating that peak radial average enthalpy should remain below 230 cal/g. Therefore, the NRC staff determined that the proposed limits were acceptable for use in the MPS3 REA analysis as the limits on damaged core coolability.

Based on the evaluations discussed in above Sections 3.5.2.1 through 3.5.2.5, the NRC staff concluded that the proposed DBLFPBs in Table 1 of Attachment 1 to the LAR, as supplemented by Attachment 3 to the RAI response (Reference 2) for the FCM temperature limits, were acceptable for the MPS3 REA analysis, because (1) the DNBR limit was previously approved by the NRC for the DNBR calculation, (2) the FCM temperature limits were derived based on the NRC approved methods, (3) fuel rod cladding failure thresholds satisfied the applicable RP C.3.1 and C.3.2, and (4) allowable limits on damaged core coolability satisfied RP C.6.a of RG 1.236.

3.5.3 DNBR Mixed-Core Penalties for MPS3 REA Analysis On page 16 of 25 of Attachment 1 to the LAR, the licensee stated that the MPS3 REA analysis incorporates a conservative thermal-hydraulic penalty that accommodates mixed-core changes in flow distribution. In RAI-10, the staff requested the licensee to provide the value of the penalty factor, how it was determined, and a justification for its use in the REA analysis.

In its response to RAI-10 (Reference 2), the licensee indicated that the GAIA transition core penalty (TCP) of 2.4% for the VIPRE-D/ORFEO-GAIA code/correlation pair was applicable to the REA analysis, which was performed with the COBRA-FLX/ORFEO-GAIA code/correlation pair. As discussed in Section 3.3 of this SE, the NRC staff reviewed and approved the TCP of 2.4% for the VIPRE-D/ORFEO-GAIA code/correlation pair. To support the application of the TCP of 2.4% for the COBRA-FLX/ORFEO-GAIA code/correlation pair used for the REA analysis, the licensee did a comparison between the results of the VIPRE-D data analysis and the equivalent data analysis performed with COBRA-FLX. The licensee presented the results of the comparison in Tables 1 and 2 of the RAI-10 response for overall statistics of the combined data set for the ORFEO-GAIA correlation. The results demonstrated good agreement in ORFEO-GAIA critical heat flux (CHF) predictions between the VIPRE-D and COBRA-FLX codes. The difference in the average measured to predicted (M/P) CHF was ((

)) for ORFEO-GAIA. The standard deviations demonstrated comparable data spreads were obtained.

The good agreement in ORFEO-GAIA test results between VIPRE-D and COBRA-FLX indicated that code-to-code differences were negligible for these correlations. Therefore, the NRC staff determined that the application of the VIPRE-D calculated DNBR TCP to COBRA-FLX DNBR results was acceptable for the REA analysis.

As described in the licensees RAI-8 response, the total number of DNB fuel failures in MPS3s REA analysis was calculated using a penalized COBRA-FLX/ORFEO-GAIA DNBR limit of 1.22.

When compared with the NRC-approved COBRA-FLX/ORFEO-GAIA DNBR design limit of 1.12 discussed in Section 3.5.2.2 of this SE, the MPS3 REA analysis retained 8.19% DNBR margin as calculated in Eq. 7 of the RAI-10 response. Therefore, the NRC staff determined that the retained 8.19% DNBR margin to the design limit of 1.12 was sufficient to bound the TCP of 2.4% and thus, concluded that the COBRA-FLX/ORFEO-GAIA DNBR limit of 1.22 used in the REA analysis was adequate, retaining sufficient margin to account for the TCP, and was therefore acceptable.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.5.4 Results of the MPS3 REA Analysis The licensees RAI-8 response showed the results of the MPS3 analysis in Figures 1 through 5 for the following most limiting cases.

1.

Maximum Fuel Temperature - the case with burnup (BU) of ((

))

2.

Fuel Rim Temperature - the case with BU of ((

))

3.

Maximum Enthalpy - the case with BU of ((

))

4.

Maximum Enthalpy Rise - the case with BU of ((

))

5.

Minimum DNBR - the case with BU of ((

))

The results demonstrated that the margin to the limits for Cases 1 through 4, including fuel temperature, fuel rim temperature, enthalpy, and enthalpy rise, was available. Therefore, the NRC staff determined that the limiting REA Cases 1 through 4 were acceptable.

Case 5 was the limiting DNBR case. The licensees RAI-8 response indicated that the MPS3-specific total number of DNB failed rods was determined using a penalized SAFDL of 1.22 (as compared to 1.12 discussed in above Section 3.5.2.2 of this SE) for the COBRA-FLX/ORFEO-GAIA code/correlation pair. The calculation was performed using the ((

)) in the NRC-approved XN-NF-82-06(P)(A), Revision 1 and Supplements 2, 4, and 5 (Reference 37) that would ((

)). The results showed that the DNB fuel cladding failure of 2.1% of all pins would occur in the core.

This result was within the limits of 10% of the rods experiencing cladding damage and 0.25%

experiencing core melt assumed in the current MPS3 REA radiological analysis in MPS3 FSAR Section 15.4.8.4. Therefore, the NRC staff concluded that the existing MPS3 REA radiological analysis remained valid.

3.5.5 REA Overpressure Analysis The licensee indicated on page 15 of 25 of Attachment 1 to the LAR that no aspect of the Framatome fuel would affect severity of the REA overpressure analysis and no overpressure reanalysis was performed.

In its RAI-9 response, the licensee indicated that an engineering assessment was performed to demonstrate the conservatism of the existing MPS3 analysis of record (AOR) methodology, WCAP-7588, Revision 1-A (Reference 30), for use of GAIA fuel at MPS3. The assessment compared key GAIA fuel and reactivity parameters calculated under the NRC-approved AREA TR ANP-10338P-A (Reference 16) to those considered in the WCAP-7588, Revision 1-A AOR.

WCAP-7588, Revision 1-A discussed analyses for both one-dimensional (1D) and 3D analysis and showed that the 1D kinetics method over-predicted the severity of the transient. This same expectation of conservatism applied to AREAs 3D analysis (Reference 16) relative to the 1D method of WCAP-7588, Revision 1-A. Application of 3D kinetics in the NRC-approved AREA topical report (Reference 16) would generate a significantly reduced core power response, and therefore a reduced system pressure response relative to the current MPS3 FSAR 1D methodology. Also, the calculated ejected rod worths from the AREA analysis for use of GAIA at MPS3 were much lower than those considered in WCAP-7588, Revision 1-A. The licensee stated that the primary system pressure analysis results using GAIA at MPS3 would be less severe than the results of WCAP-7588, Revision 1-A, because of (1) the margins provided by

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION AREAs lower predicted ejected rod worths and (2) the AREA 3D methodology. Therefore, the NRC staff concluded that reasonable assurance was provided that the existing FSAR overpressure analysis supported by WCAP-7588, Revision 1-A remained conservative and valid for GAIA fuel at MPS3.

3.6 Technical Evaluation of the TS Changes 3.6.1 Evaluation of Proposed Change to TS 2.1.1.1 GDC 10 requires that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and AOOs. This is accomplished by having a DNB design basis. The MPS3 TS 2.1.1.1 currently requires that DNB ratio (DNBR) for resident Westinghouse fuel be maintained greater than or equal to 1.14 using the NRC staffs approved WRB-2M DNB correlation. The licensee proposed the addition of the VIPRE-D/ORFEO-GAIA DDL of 1.13 to TS 2.1.1.1 which is obtained from DOM-NAF-2-P-A, Appendix F. As provided in the LAR, the licensee will retain the WRB-2M DNB limit to support the current Westinghouse fuel product during the transition to full cores of GAIA.

The NRC staff considers the proposed addition of VIPRE-D/ORFEO-GAIA DDL of 1.13 to the reactor core safety limit in TS 2.1.1.1 to be acceptable since the DDL was correctly derived based on the previously approved methods. The DDL value is applicable to the Framatome GAIA fuel which is scheduled for onload during the Spring 2025 refueling outage. Acceptance of the 1.13 value for VIPRE-D/ORFEO-GAIA DDL was appropriately based on approved methods in Appendix F of Fleet Report DOM-NAF-2-P-A for the Framatome GAIA fuel which was approved by the NRC staff on December 20, 2023 (Reference 15). Accordingly, the NRC staff finds that the proposed revision to TS 2.1.1.1 is acceptable and satisfies the requirements of 10 CFR 50.36(c)(1).

3.6.2 Evaluation of proposed changes to TS Table 2.2-1 for Reactor Trip System Instrumentation Trip Setpoints The Reactor Trip System (RTS) initiates a unit shutdown, based on the values of selected unit parameters, to protect against violating the core fuel design limits and Reactor Coolant System (RCS) pressure boundary during AOOs and to assist the Engineered Safety Features (ESF)

Systems in mitigating accidents. TS 2.2-1 requires the RTS Instrumentation and Interlock Setpoints be set consistent with the NTS values shown in Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints. Item 18.c in the Table addresses the NTS and AV for the Power Range Neutron Flux permissive-8 (P-8). The proposed changes would reduce the P-8 NTS from 50.0% rated thermal power (RTP) to 35% RTP and AV from 50.6% RTP to 35.6% RTP.

According to the licensee, the lower setpoint would maintain the margin of safety for GAIA fuel.

The NRC staffs evaluation for the proposed changes in section 3.4 determined that lower power values of the P-8 trip setpoint would result in an earlier actuation (i.e., lower power) of the reactor trip and limit the plant operations to more restrictive conditions. The staff evaluation confirmed that the licensees P-8 analysis was performed with the NRC-approved code at MPS3 for RCS transient analyses as well as for the DNBR calculations. Section 50.36(b) of 10 CFR requires that each license authorizing the operation of a facility will include TSs and that the TSs will be derived from the safety analysis. The regulation under 10 CFR 50.36(c)(1)(ii)(A) requires that where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION abnormal situation before a safety limit is exceeded. Since the staffs review determined that the proposed reduction to the P-8 trip setpoints would result in more restrictive plant operations (i.e.,

the chosen values of the trip setpoint would result in an earlier actuation of the reactor trip and are necessary to meet the GAIA DNBR safety limits, the NRC staff finds the changes acceptable. Based on the above considerations, the NRC staff finds that the proposed revision to TS Table 2.2-1 is acceptable and satisfies the requirements of 10 CFR 50.36(c)(1)(ii)(A).

3.6.3 Evaluation of proposed change to TS 6.9.1.6.b TS 6.9.1.6.b, Core Operating Limits Report (COLR), identifies the NRC-approved analytical methodologies that are used to determine the core operating limits for MPS3. TS 6.9.1.6.b, Core Operating Limits Report (COLR), requires that the cycle-specific COLR contain a complete listing of TS referenced topical reports (i.e., report number, title, revision, date, and any supplements). The licensee proposed to add Report, Appendix F of DOM-NAF-2-P-A, which describes ORFEO-GAIA and ORFEO-NMGRID correlation application with VIPRE-D to item 23 of the TS section. Appendix C and D of DOM-NAF-2-P-A, will be retained to support the current Westinghouse fuel product during the transition to full cores of GAIA. VEP-NE-2-A is currently included as a COLR reference.

Section 50.36(b) of 10 CFR states, in part, that, the technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to Section 50.34. Section 50.36(c)(5), states, in part, that, administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure the operation of the facility in a safe manner. The staffs review in Section 3.2 of this SE determined that applications of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF correlations with VIPRE-D Computer Code support updating of specific TS requirements as addressed above. Based on the NRC review and approval of the Appendix F to the Fleet Report DOM-NAF-2-P-A (Reference 15), the methodologies for the CHF correlations are found to be acceptable and appropriate for inclusion in TS 6.9.1.6.b. Based on the above considerations, the NRC staff finds that the proposed revision to TS 6.9.1.6.b is acceptable and satisfies the requirements of 10 CFR 50.36(c)(5).

3.7 Technical Conclusion The NRC staff reviewed the licensees P-8 analysis and found that (1) the analysis was performed using the NRC-approved methods and conservative initial plant conditions, and (2) the results showed that the analytical limit of 45% RTP for the P-8 function provided significant margin to maintain the minimum DNBR within the applicable limits, assuring no fuel failure, as applicable to all relevant operational occurrences, and thus satisfying the GDC 10 requirements. Therefore, the NRC staff concluded that the provided adequate technical basis to support the proposed TS changes to TS Table 2.2-1.

The NRC staff reviewed the proposed DBLFPBs in Table 1 of Attachment 1 to the LAR, as supplemented, and determined the limits were acceptable for the MPS3 REA analysis, because the limits (1) were previously approved by the NRC for the DNBR calculation, (2) were derived based on the NRC approved methods for fuel melt temperature, (3) satisfied the applicable regulatory positions for fuel rod cladding failure thresholds, or (4) satisfied the applicable regulatory positions of RG 1.236 for allowable limits on damaged core coolability.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The NRC staff reviewed the MPS3 REA analysis and found that the analysis was performed using acceptable analytical models with assumptions in accordance with the guidance and acceptance criteria in RG 1.236 and that the analysis demonstrated that appropriate reactor protection and safety systems would prevent postulated REAs that could: (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding; or (2) cause sufficient damage that would significantly impair the capability to cool the core. Therefore, the NRC staff concluded that the plant would continue to meet the relevant requirements of GDC 28, as applicable to RG 1.236, following implementation of the GAIA fuel transition at MPS3 and that the REA analysis was acceptable.

The NRC staff has reviewed the licensees LAR, as supplemented to support the plant-specific application of the VIPRE-D thermal hydraulics code with the ORFEO-GAIA and ORFEO--NMGRID CHF correlations for use of Framatome GAIA fuel in MPS3 cores and finds that the licensee has provided adequate technical basis to support the proposed TS changes.

Specifically, the NRC staff has concluded that the following items are acceptable and in compliance with 10 CFR 50.36(c)(1) and (c)(5) and GDC 10:

The proposed addition of VIPRE-D/ORFEO-GAIA DDL of 1.13 to TS 2.1.1.1 The proposed addition of DOM-NAF-2-P-A, Appendix F, which describes ORFEO-GAIA and ORFEO-NMGRID correlation application with VIPRE-D to the TS 6.9.1.6.b: Core Operating Limits Report (COLR) reference list The SDL value of 1.26 for VIPRE-D/ORFEO-GAIA and the SDL value of 1.31 for VIPRE-D/ORFEO-NMGRID documented in Attachment 3 of the LAR.

A TCP of 2.4% for DNBR analysis calculated using the VIPRE-D/ORFEO-GAIA code/correlation pair and a TCP of 2.7% using the VIPRE-D/ORFEO-NMGRID code/correlation pair above the GAIA first mixing vane spacer grid.

The NRC staff notes that this safety evaluation and the subsequent conclusions presented herein are applicable to the LAR in support of full batch loading of Framatome GAIA fuel assemblies with M5 cladding to the MPS3 reactor as long as the licensee complies with the methodologies as described within the LAR.

Based on its review as described in detail above, the NRC staff concludes that the licensee has provided adequate technical basis to support the proposed TS changes. Specifically, the NRC staff finds the licensee has demonstrated that (1) the methods proposed by the licensee are applicable for the intended purpose, (2) the Framatome GAIA fuel assembly specific safety analyses results meet the applicable licensing criteria, and (3) the proposed TS changes are acceptable and satisfy the 10 CFR 50.36 requirements.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Connecticut State official was notified of the proposed issuance of the amendment on September 30, 2024. The State official had no comments.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, published in the Federal Register on January 2, 2024 (89 FR 108), and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

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

7.0 REFERENCES

1.

Holloway, J., Dominion Energy Nuclear Connecticut, Inc., letter to U. S. Nuclear Regulatory Commission, Millstone Power Station Unit 3, Proposed Amendment to Support Implementation of Framatome GAIA Fuel, October 30, 2023, ADAMS (ADAMS)

Accession No. ML23304A047 (Public), ML23304A046 (Proprietary, Non-Public).

2.

Holloway, J., Dominion Energy Nuclear Connecticut, Inc., letter to U. S. Nuclear Regulatory Commission, Millstone Power Station Unit 3, Response to Request for Additional Information Regarding Proposed Amendment to Support Implementation of Framatome GAIA Fuel, September 16, 2024, ML24260A195 (Public), ML24260A194 (Proprietary, Non-Public).

3.

Holloway, J., Dominion Energy Nuclear Connecticut, Inc., letter to U. S. Nuclear Regulatory Commission, Millstone Power Station, Units 1, 2, and 3 Updates to the Final Safety Analysis Reports, dated June 28, 2023, Package ML23193A862.

4.

Standard Review Plan, Section 4.2, Fuel System Design, NUREG-0800 Revision 3, U.S. Nuclear Regulatory Commission, March 2007, ML070740002.

5.

Standard Review Plan, Section 4.3, Nuclear Design, NUREG-0800 Revision 3, U.S.

Nuclear Regulatory Commission, March 2007, ML070740003.

6.

Standard Review Plan, Section 4.4, Thermal and Hydraulic Design, NUREG-0800 Revision 2, U.S. Nuclear Regulatory Commission, March 2007, ML070550060.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION

7.

Standard Review Plan, Section 15.4.8, Spectrum of Rod Ejection Accidents (PWR),NUREG-0800 Revision 3, U.S. Nuclear Regulatory Commission, March 2007, ML070550014.

8.

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10.

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11.

Dominion Energy Fleet Report, VEP-NE-2-A, Statistical DNBR Evaluation Methodology, June 1987, ML101330527 (Non-Public).

12.

Holloway, J., Dominion Energy, letter to U. S. Nuclear Regulatory Commission, Virginia Electric and Power Company (Dominion Energy Virginia), North Anna and Surry Power Stations Units 1 and 2, Dominion Energy Nuclear Connecticut, Inc. (DENC), Millstone Power Station Units 2 and 3, Request for Approval of Appendix F of Fleet Report DOM-NAF-2-P Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code, December 19, 2022, ML22353A620 (Public), ML22353A619 (Proprietary, Non-Public).

13.

Holloway, J., Dominion Energy, letter to U. S. Nuclear Regulatory Commission, Virginia Electric and Power Company (Dominion Energy Virginia), North Anna and Surry Power Stations Units 1 and 2, Dominion Energy Nuclear Connecticut, Inc. (DENC), Millstone Power Station Units 2 and 3, Request for Approval of Appendix F of Fleet Report DOM-NAF-2-P Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code, Response to Request for Additional Information, April 6, 2023, ML23096A298 (Public),

ML23096A297 (Proprietary, Non-Public).

14.

Holloway, J., Dominion Energy, letter to U. S. Nuclear Regulatory Commission, Virginia Electric and Power Company (Dominion Energy Virginia), North Anna and Surry Power Stations Units 1 and 2, Dominion Energy Nuclear Connecticut, Inc. (DENC), Millstone Power Station Units 2 and 3, Request for Approval of Appendix F of Fleet Report DOM-NAF-2-P Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code, Response to Second Request for Additional Information, July 26, 2023, ML23208A092 (Public),

ML23208A091 (Proprietary, Non-Public).

15.

Miller, G. E., U.S. Nuclear Regulatory Commission, letter to Holloway, J. E. Dominion Energy, North Anna Power Station, Unit Nos. 1 and 2, Surry Power Station Unit Nos. 1 and 2, and Millstone Power Station, Unit Nos. 2 and 3 - Review of Appendix F to DOM-NAF-2, Qualification of the Framatome ORFEO-GAIA and ORFEO-NMGRID CHF Correlations in the Dominion Energy VIPRE-D Computer Code, (EPID L-2022-LLT-

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 0003), December 20, 2023, ML23283A305 (Public), ML232083A304 (Proprietary, Non-Public).

16.

ANP-10338-P-A, Revision 0, AREA' - ARCADIA Rod Ejection Accident, December 2017, ML18059A782 (Public), ML18059A7823 (Proprietary, Non-Public).

17.

Holloway, J., Dominion Energy Nuclear Connecticut, Inc., letter to U. S. Nuclear Regulatory Commission, Millstone Power Station Unit 3, License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits and Exemption Request for use of M5TM Cladding, May 2, 2023, ML23123A279.

18.

Holloway, J., Dominion Energy Nuclear Connecticut, Inc., letter to U. S. Nuclear Regulatory Commission, Millstone Power Station Unit 3, Proposed Amendment to Revise Technical Specifications for Reactor Core Safety Limits, Fuel Assemblies, and Core Operating Limits Report Related to Framatome GAIA Fuel, May 23, 2023, ML23145A195.

19.

Guzman, R. V., U.S. Nuclear Regulatory Commission, letter to Heacock, D. A. Dominion Nuclear, Millstone Power Station, Unit No. 3 - Issuance of Amendment Adopting Dominion Core Design and Safety Analysis Methods for Addressing the Issues Identified in three Westinghouse Communication Documents, (CAC No. MF6251), July 28, 2016, ML16131A728.

20.

Sinha, S., Dominion Energy, email to Guzman, R. V., U.S. Nuclear Regulatory Commission, Millstone Power Station, Unit 3 - Acceptance Review Determination re:

LAR to Support the Implementation of Framatome GAIA Fuel (EPID L-2023-LLA-0150),

February 27, 2024, ML24058A232.

21.

Morey, D. C., U.S. Nuclear Regulatory Commission, letter to Peters, G. AREVA, Inc.,

Final Safety Evaluation for AREVA Inc. Topical Report ANP-10338P, Revision 0, AREATM - ARCADIA Rod Ejection Accident, (CAC No. MF7009/EPID: L-2015-TOP-004), December 20, 2017, ML17334A780 (Public), ML17334A777 (Proprietary, Non-Public).

22.

ANP-10342NP-A, Revision 0, GAIA Fuel Assembly Mechanical Design, Topical Report, Framatome, Inc., September 2019, ML19309D916.

23.

ANP-10297NP-A, Revision 0, The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results, Topical Report, AREVA NP, Inc.,

March 2010, ML14195A145.

24.

ANP-10297NP-A, Supplement 1NP-A, Revision 1, The ARCADIA Reactor Analysis System for PWRs Methodology Description and Benchmarking Results, Topical Report, Framatome, Inc., December 2020, ML21071A064.

25.

ANP-10341NP-A, Revision 0, The ORFEO-GAIA and ORFEO-NMGRID Critical Heat Flux Correlations, Topical Report, AREVA, Inc., August 2016, ML16238A078 (Public),

ML16238A076 (Proprietary, Non-Public).

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION

26.

BAW-10227P-A, Revision 1, Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel, June 2003, ML15162B047 (Public), ML15162B052 (Proprietary, Non-Public).

27.

ANP-10339P-A, Revision 0, ARITA - ARTEMIS/RELAP Integrated Transient Analysis Methodology, Topical Report, Framatome, Inc., October 2023, ML24026A140 (Public),

ML24026A141 (Proprietary, Non-Public).

28.

ANP-10323P-A, Revision 1, GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors, Topical Report, Framatome, Inc., November 2020, ML21005A028 (Package).

29.

Witt, T. A., Ameren Missouri to letter to U. S. Nuclear Regulatory Commission, Callaway Plant Unit 1, Supplement to License Amendment and Exemption Request Regarding Use of Framatome GAIA Fuel (LDCN 22-0002), August 3, 2023, ML23215A196 (Package).

30.

WCAP-7588, Revision 1-A, An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods, Westinghouse Electric Corporation, January 1975, ML20330A094 (Proprietary, Non-Public).

31.

Guzman, R. V., U.S. Nuclear Regulatory Commission, letter to Carr, E. S., Dominion Nuclear, Millstone Power Station, Unit 3 - Regulatory Audit Plan in Support of License Amendment Request to Implement Framatome GAIA Fuel (EPID L-2023-LLA-0150),

April 4, 2024, ML24088A330.

32.

Guzman, R, V., U.S. Nuclear Regulatory Commission, letter to Carr, E.S., Dominion Nuclear, Millstone Power Station, Unit 3 - Summary of Regulatory Audit Supporting the Review of License Amendment Request for Implementation of Framatome GAIA Fuel (EPID L-2023-LLA-0150), September 3, 2024, ML24240A153.

33.

Guzman, R, U.S. Nuclear Regulatory Commission, email to Sinha, S, Dominion Energy Nuclear Connecticut, Inc., Millstone Power Station, Unit 3 - Request for Additional Information Re; LAR to Support Implementation of Framatome GAIA Fuel (EPID L-2023-LLA-0150), July 31, 2024, ML24213A260.

34.

Chawla, M. L., U.S. Nuclear Regulatory Commission, letter to Diya, F., Ameren Missouri, Callaway Plant, Unit No. 1 - Issuance of Amendment No. 235 to Revise Technical Specifications to Use Framatome GAIA Fuel (EPID L-2022-LLA-0150), October 5, 2023, ML23240A369.

35.

Witt, T. A., Ameren Missouri, letter to U. S. Nuclear Regulatory Commission, Callaway Plant, Unit 1 - Final Safety Analysis Report Revision OL-27 and Technical Specification Bases Revision 25, Chapter 15, Accident Analysis, dated May 6, 2024, ML24149A267.

36.

Gullott, D. M., Exelon Generation Company, LLC, letter to U. S. Nuclear Regulatory Commission, Byron/Braidwood Stations, Revision 18 Updated Final Safety Analysis Report, Chapter 15, Accident Analysis, dated December 17, 2020, ML21008A400.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION

37.

XN-NF-82-06(P)(A), Revision 1, Supplements 2, 4, and 5, Qualification of Exxon Nuclear Fuel for Extended Burnup, Exxon Nuclear Company, November 1985, ML0881710707, ML081710199 (Proprietary, Non-Public).

Principal Contributors: S. Sun, NRR N. Amini, NRR R. Grover, NRR Date: November 19, 2024