ML24109A003

Issuance of Amendment No. 289 to Revise Technical Specifications to Use Framatome Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limit (EPID L-2023-LLA-0065) (Non-Proprietary)

ML24109A003
Person / Time
Site:	Millstone
Issue date:	05/21/2024
From:	Richard Guzman NRC/NRR/DORL/LPL1
To:	Carr E Dominion Energy Nuclear Connecticut
Shared Package
ML24109A004	List: ML24109A003
References
EPID L-2023-LLA-0065
Download: ML24109A003 (40)

Text

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION May 21, 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. 289 TO USE FRAMATOME LOSS-OF-COOLANT ACCIDENT EVALUATION METHODOLOGIES FOR ESTABLISHING CORE OPERATING LIMITS (EPID L-2023-LLA-0065)

Dear Eric Carr:

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued the enclosed Amendment No. 289 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 May 2, 2023, as supplemented by letter dated April 1, 2024.

The amendment revises Technical Specification (TS) 6.9.1.6.b to support the transition to Framatome (FRM) GAIA fuel with M5TM cladding at Millstone, Unit 3, and the resulting application of the FRM Small Break and Realistic Large Break Loss-of-Coolant Accident (LOCA) methodologies and the associated use of the GALILEO fuel performance code within the LOCA methods. Specifically, FRM Topical Reports EMF-2328-P-A, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, EMF-2103-P-A, Realistic Large Break LOCA Methodology for Pressurized Water Reactors, and ANP-10349-P-A, GALILEO Implementation in LOCA Methods, are added to the list of methodologies for establishing core operating limits.

The revision to TS 6.9.1.6.b also deletes one methodology no longer used to define core operating limits at Millstone, Unit 3. Since the FRM GAIA fuel assemblies contain fuel rods fabricated with M5TM cladding material, the licensee included an exemption request in accordance with Title 10 of the Code of Federal Regulations (10 CFR) Section 50.46 and 10 CFR Part 50, Appendix K as Attachment 8 to the application. The staff reviewed the exemption request in a separate exemption technical evaluation dated May 21, 2024.

The NRC staff has determined that the related SE contains proprietary information pursuant to 10 CFR 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 Safety Evaluation 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. 289 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. 289 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 May 2, 2023, as supplemented by letter dated April 1, 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. 289 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: May 21, 2024

ATTACHMENT TO LICENSE AMENDMENT NO. 289 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 6-20a 6-20a 6-21 6-21 6-21a 6-21a 6-21b 6-21b 6-21c

(2)

Technical Specifications The Technical Specifications contained in Appendix A, revised through Amendment No. 289 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. 270-287, 289

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, and Appendix D, Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion 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 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.

STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.7 A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with TS 6.8.4.g, Steam Generator (SG)

Program. The report shall include:

a.

The scope of inspections performed on each SG, Amendment No. 24, 40, 50, 69, 104, 173, 212, 215, 229, 238, 245, 249. 252, 255, 256, 279, 6-21 289

MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS STEAM GENERATOR TUBE INSPECTION REPORT (Continued) b.

Degradation mechanisms found, c.

Nondestructive examination techniques utilized for each degradation mechanism, d.

Location, orientation (if linear), and measured sizes (if available) of service induced indications, e.

Number of tubes plugged during the inspection outage for each degradation mechanism, f.

The number and percentage of tubes plugged to date and the effective plugging percentage in each steam generator.

g.

The results of condition monitoring, including the results of tube pulls and in-situ

testing, h.

The primary to secondary LEAKAGE rate observed in each SG (if it is not practical to assign the LEAKAGE to an individual SG, the entire primary to secondary LEAKAGE should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report, i.

The calculated accident induced leakage rate from the portion of the tubes below 15.2 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.49 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined; and j.

The results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator Region I, and one copy to the NRC Resident Inspector, within the time period specified for each report.

Amendment No. 238, 245, 249, 252, 255, 256, 279, 6-21a 289

MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS 6.10 Deleted.

6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.

6.12 HIGH RADIATION AREA As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601 (a) and (b) of 10 CFR Part 20:

6.12.1 High Radiation Areas with Dose Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation a.

Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

b.

Access to, and activities in, each such area shall be controlled by means of a Radiation Work Permit (RWP) or equivalent; that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c.

Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following plant radiation protection procedures for entry to, exit from, and work in such areas.

d.

Each individual or group entering such an area shall possess:

1.

A radiation monitoring device that continuously displays radiation dose rates in the area, or Amendment No. 245, 279, 6-21b 289

MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS 6.12 HIGH RADIATION AREA (cont.)

2.

A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the devices dose alarm setpoint is reached, with an appropriate alarm setpoint, or 3.

A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or Amendment No. 245, 279, 6-21c 289

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. 289 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. 289 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 May 2, 2023 (Reference 1), as supplemented by letter dated April 1, 2024 (Reference 2), Dominion Energy Nuclear Connecticut, Inc. (the licensee), submitted a license amendment request (LAR) to revise the technical specifications (TS) for Millstone Power Station, Unit 3 (MPS3). Specifically, the proposed amendment would revise TS 6.9.1.6.b to support the transition to Framatome (FRM) GAIA fuel with M5TM cladding at MPS3 and the resulting application of the FRM Small Break and Realistic Large Break Loss-of-Coolant Accident (LOCA) methodologies and the associated use of the GALILEO fuel performance code within the LOCA methods. Specifically, FRM Topical Reports (TR) EMF-2328-P-A, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, EMF-2103-P-A, Realistic Large Break LOCA Methodology for Pressurized Water Reactors, and ANP-10349-P-A, GALILEO Implementation in LOCA Methods, would be added to the list of methodologies in TS 6.9.1.6.b for establishing core operating limits (COLs). The proposed revision to TS 6.9.1.6.b would also delete one methodology no longer used to define COLs at MPS3.

Since the FRM GAIA fuel assemblies contain fuel rods fabricated with M5TM cladding material, the licensee has included a Title 10 of the Code of Federal Regulations (10 CFR) Section 50.46, Acceptance criteria for emergency core cooling systems [ECCS] for light-water nuclear power reactors, and 10 CFR Part 50, Appendix K, ECCS Evaluation Models, exemption request as to the LAR. The U.S. Nuclear Regulatory Commission (NRC or Commission) staff reviewed the exemption request in a separate exemption technical evaluation dated May 21, 2024 (Reference 3).

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 August 31, 2023 (Reference 18), NRC issued an audit plan, which provided the list of requested information, documents, and other details pertaining to the audit. By letter dated April 22, 2024 (Reference 19), the NRC staff issued the Audit Summary.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The supplemental letter dated April 1, 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 August 1, 2023 (88 FR 50183).

2.0 REGULATORY EVALUATION

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

Regulatory Requirements Under 10 CFR 50.92(a), in determining whether an amendment to a license will be issued, the NRC staff is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. The common standards for licenses in 10 CFR 50.40(a), and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public, and that the applicant will comply with the Commissions regulations.

Accordingly, for this action, the NRC staff must conclude that there is reasonable assurance that the proposed changes to the technical specifications do not endanger public health and safety.

The regulations under 10 CFR 50.36, Technical specifications, provide regulatory requirements related to the content of TSs. Specifically, Paragraph 50.36(c)(5), Administrative controls, states, in part, that Administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in safe manner.

Requirements for analyzing the design-basis LOCA are provided in 10 CFR 50.46 and 10 CFR Part 50, Appendix K, and Appendix A, General Design Criteria for Nuclear Power Plants, General Design Criterion (GCC) 35, Emergency core cooling.

The regulations under 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactor, insofar as it requires that each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding1 must be provided with an ECCS that must be designed so that its calculated 1 To support the use of M5TM fuel rod cladding at MPS3, an exemption to the requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50 was submitted by the licensee as an attachment to the LAR. The NRC staff reviewed the exemption request and discussed its evaluation in a separate SE dated May 21, 2024 (Reference 3).

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION cooling performance following a postulated LOCA conforms to the criteria set forth in 10 CFR 50.46(b)(1) through (b)(4):

(1)

Peak cladding temperature. The calculated maximum fuel element cladding temperature shall not exceed 2200 °F.

(2)

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

(3)

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

(4)

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

Appendix K to 10 CFR Part 50, sets for the documentation requirements for each evaluation model (EM), and establishes required and acceptable features of EMs for heat removal by the ECCS.

GDC 35 requires abundant core cooling sufficient to (1) prevent fuel and cladding damage that could interfere with effective core cooling and (2) limit the metal-water reaction on the fuel cladding to negligible amounts. GDC 35 further requires suitable redundancy of the ECCS, such that it can accomplish its design functions, assuming a single failure, irrespective of whether its electrical power is supplied from offsite or onsite sources.

The MPS3 Final Safety Analysis Report (FSAR) section 3.1 (Reference 4), discusses the extent to which the design criteria for MPS3 structures, systems, and components important to safety comply with 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants. 2 Regulatory Guidance Generic Letter (GL) 88-16, Removal of Cycle-Specific Parameter Limits from Technical Specifications, insofar as it provides guidance for modifying TSs to remove cycle-specific parameter limits from the TSs to a licensee-controlled core operating limits report (COLR).

(Reference 5)

The NRC staff reviewed the licensees submittal to evaluate the applicability of Framatome methodology for MPS3 to confirm that the use of the methodologies is within the NRC-approved ranges of applicability and the results of the analyses are in compliance with the applicable requirements of the regulatory requirements listed above.

2 Conformance with GDC 35 is described in Section 3.1.2.35 of the MPS3 FSAR and is unaffected by the proposed TS changes. The proposed TS changes does not require relief from any other regulatory requirements and does not affect compliance with the GDC differently than described in the MPS3 FSAR.

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

3.0 TECHNICAL EVALUATION

3.1 Summary of the Proposed TS Changes The licensee proposed to use the methodologies documented in FRM TR EMF-2328-P-A (Reference 8 and 10), PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, EMF-2103-P-A (Reference 12), Realistic Large Break LOCA Methodology for Pressurized Water Reactors," and ANP-10349-P-A (Reference 14), GALILEO Implementation in LOCA Methods, for the LOCA analysis in support of the transition of the FRM GAIA fuel with M5 cladding at MPS3. In accordance with 10 CFR 50.46, the licensee performed a break spectrum analysis of the postulated small break LOCA (SBLOCA) and realistic large break LOCA events and submitted in the LAR two analytical reports documenting its plant specific LOCA analyses for MPS3: ANP-4031 (Reference 6), Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design, and ANP-4032 (Reference 7), Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design.

The NRC staff reviewed the SBLOCA analysis in ANP-4031, RLBLOCA analysis in ANP-4032, and the licensees request for additional information (RAI) response (Reference 2) for MPS3 and discusses its evaluation in the sections below.

3.2 SBLOCA Analysis The NRC staff reviewed the SBLOCA analysis based on the information documented in ANP-4031 (Reference 6) and the RAI response (Reference 2), and focused its review on the following subjects: (1) analytical methods used for the SBLOCA analysis; (2) compliance with the limitations and restrictions imposed in NRC safety evaluation report (SER) for the methods; (3) values of plant parameters used as input to the analytical models; (4) compliance of the analytical results with the 10 CFR 50.46(b) acceptance criteria; and (5) conclusion for the SBLOCA analysis. The NRC staffs evaluation of the SBLOCA analysis is discussed in sections 3.1.1 through 3.1.5, as follows.

3.2.1 SBLOCA Analytical Methods As discussed in Section 3.3 of ANP-4031P, the SBLOCA analysis was performed based on the SBLOCA evaluation model (EM) in the NRC-approved FRM TRs, EMF-2328-P-A, Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, (Reference 8), including Supplement 1-P-A, Revision 0 (Reference 10), and ANP-10349-P-A, Revision 0, GALILEO Implementation in LOCA Methods (Reference 14). The EM consisted of two computer codes:

(1) S-RELAP5 for predicting the primary and secondary system thermal-hydraulic and hot rod transient response and (2) GALILEO for determining the burnup dependent initial fuel rod conditions used as input to S-RELAP5 for the system response calculations. The SBLOCA EM was a conservative methodology that met the requirements of Appendix K to 10 CFR 50. The NRC staffs generic approval of the FRM SBLOCA EM was documented in References 9 and 11 for use of the S-RELAP5 code and References 14 and 15 for use of the GALILEO code.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.2.2 SBLOCA Methods - Compliance with NRC SER Limitations 3.2.2.1 Limitations in the NRC SER for TR EMF-2328-P-A, Revision 0 The NRC SER (Reference 9) for the TR, EMF-2328-P-A Revision 0, limited the use of the EM to the LOCA analysis for break sizes up to 10 percent of the reactor coolant system (RCS) cold-leg flow area.

The NRC staff reviewed the LAR and found in Section 3.5 of ANP-4031P that the SBLOCA analysis was performed for a break spectrum of cold-leg breaks ranging from 1.00-inch diameter to 8.70-inch diameter, which is 10 percent of cold-leg piping area. Therefore, the NRC staff finds that the SBLOCA analysis satisfies the limitation for the break sizes.

3.2.2.2 Requirements in the NRC SER for TR Supplement 1-P-A to EMF-2328-P-A, Rev. 0 The NRC SER (Reference 11) for the TR, Supplement 1-P-A to EMF-2328-P-A Revision 0 had no limitations and conditions. However, Section 4 of the SER imposed modelling requirements on the following eight areas.

1. Spectrum of break sizes,

2. Breaks in the attached piping,

3. Delayed reactor coolant pump (RCP) trip,

4. Maximum accumulator/safety injection tank/refueling water storge tank fluid temperature,

5. Core bypass-flow path in the reactor vessel,

6. Reactivity feedback,

7. Loop seal clearing and cross-over leg modelling, and

8. Core nodalization.

The licensees discussion of its compliance with model requirements (MRs) 1 through 4 in Sections 4.1, and 4.3 through 4.6 of ANP-4031P, and MRs 5 through 8 in the RAI-1 response (Reference 2) demonstrated that the licensee satisfied each of the MRs for use of the methodologies in Supplement 1-P-A to EMF-2328-P-A, Revision 0 in the SBLOCA analysis. The NRC reviewed the licensees disposition of the MRs as follows:

MR1 - Spectrum of Break Sizes MR1 in Section 4.1 of the SER (Reference 11) for TR Supplement 1-P-A to EMF-2328-P-A, Revision 0, required that: (1) the SBLOCA analyses include ((

)) up to, and including, the break that represented 10 percent of the RCS cold-leg flow area; and (2) the limiting break be identified using a finer break spectrum resolution near the potential worst size displaying the highest peak cladding temperature (PCT), such as the

((

)) break incremental change, to capture the limiting break.

The licensee presented the results of the SBLOCA analysis in Table 4-1, Table 4-2, and Figure 4-1 of ANP-4031P. The break spectrum analysis shown in Tables 4-1 and 4-2 was performed for RCS cold-leg breaks with sizes ranging from ((

)) using the resolution of ((

)) equivalent diameter increments. For the break sizes from ((

)), an ((

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

)) was used. The break spectrum analysis showed that the limiting PCT of 1622 degrees Fahrenheit (°F), which was within the 10 CFR 50.46(b)(1) criterion, would occur for the 8.60-inch break. Based on the above, the NRC staff finds that the range of the break sizes and break size resolutions for the break spectrum analysis satisfies MR1.

MR2 - Breaks in the Attached Piping MR2 required that the SBLOCA analysis be performed for breaks in the attached piping, including safety injection lines.

In Section 4.3 of ANP-4031P, the licensee discussed the required analysis of the attached piping breaks. The ECCS at MPS3 consisted of the accumulator, intermediate head safety injection (IHSI), low head SI (LHSI) and high head SI (HHSI) systems. At MPS3, the IHSI and LHSI flow was injected to the accumulator injection line which was connected to each RCS cold-leg. HHSI flow was injected through a separate line that was connected to each RCS cold-leg. Therefore, the licensee analyzed LOCAs at two locations: accumulator line and HHSI line. The cases analyzed assumed a double-ended guillotine of the accumulator line and HHSI line. The results demonstrated that the accumulator line break analysis was less limiting than that of the break spectrum analysis. For the HHSI line break case, the calculated PCT was bounded by that of the break spectrum analysis, while the maximum local oxidation (MLO) and core wide oxidations (CWO) were higher than those of the break spectrum analysis. The HHSI line break analysis showed the limiting total MLO and limiting CWO values were 4.28 percent and 0.08 percent, respectively, which are within the 10 CFR 50.46(b)(2-3) criteria. Also, Section 3.5 of ANP-4031P indicated that the accumulator line break analysis was based on the size of 0.4176 feet squared (ft2) (8.75-inch) accumulator line inside diameter, which was slightly greater than 10 percent of the cold-leg flow area with a break area of 0.4128 ft2 (8.70-inch diameter). However, the flow area limitation on the EM in TR EMF-2328, Revision 0 was imposed on the analysis of the cold-leg break, instead of accumulator line break or HHSI locations. The limitation was not applicable to the analysis of the accumulator or HHSI line breaks. In addition, a small increase (1.16 percent, (0.4176 ft2/0.4128 ft2) - 1) in the break size for the accumulator line break analysis above the flow area limitation would not change the thermal-hydraulic phenomena of a SBLOCA significantly. Based on the above, the NRC staff finds that the application of the EM in EMF-2328 for the analysis of accumulator line or HHSI line break is acceptable, and the attached piping break analysis satisfies MR2.

MR3 - Delayed RCP Trip Analysis MR3 required that SBLOCA analysis include a spectrum of hot-and cold-leg breaks to support the RCP trip procedure and determine/verify the trip timing consistent with Emergency Operating Procedures.

In Section 4.4 of ANP-4031P, the licensee discussed the delayed RCP trip analysis. The analysis assumed a delayed time of 5 minutes after reaching the RCP trip criteria for the operator to trip all four RCPs. RCP trip criteria were based on the MPS3 plant procedures and were modeled as the occurrence of the RCS pressure trip with concurrence of at least one charging pump running or one safety injection (SI) pump capable of injecting. The delayed time of 5 minutes for RCP trip assumed in the analysis was consistent with the current MPS3 time critical operator action program. The analysis was performed for the RCS cold-and hot-leg breaks for break sizes ranging 1.00 to 8.70 inches. The licensee also performed a second RCP trip analysis with the delayed time of 1-minute (RAI-2 response in Reference 2). Other

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION assumptions and inputs to the analysis remained consistent with the requirements of Appendix K to 10 CFR 50. The delayed RCP trip time of 1-minute was based on an average of operator training results for MPS3 that confirmed the time critical operator actions were accomplished within the time credited by the safety analyses. The results of the RCP trip analysis for both cases of 5-minute and 1-minute demonstrated compliance with 10 CFR 50.46(b) acceptance criteria regarding the PCT, MLO, and CWO. Based on the above, the NRC staff finds that the delayed RCP trip analyses satisfies MR3.

MR4 - Temperature Sensitivity Study MR4 required that the SBLOCA analyses include a ((

)).

In Section 4.5 of ANP-4031P, the licensee discussed the ECCS temperature sensitivity study.

((

)).

In Section 4.6 of ANP-4031P, the licensee discussed the RWST drain-down analysis to determine the impact of the ((

)) the SBLOCA results. During an LOCA, ((

)). Based on the above, the NRC staff finds that the ECCS temperature sensitivity study satisfies MR4.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION MR5 - Core Bypass-Flow Path in the Reactor Vessel MR5 required that: (1) the SBLOCA analyses should use the vessel nodal scheme for S-RELAP5 including ((

)); and (2) the SBLOCA analyses for Westinghouse-designed plants, such as MPS3, should include the model representing ((

)).

The core flow paths used in SBLOCA reactor vessel model for MPS3 were shown in Figure 3-3 of ANP-4031P. The licensee confirmed (RAI-1 response in Reference 2) that the reactor vessel nodal scheme did not consider ((

)). The ((

)) flow path was represented by ((

)). Based on the above, the NRC staff finds that reactor vessel nodal scheme used for the SBLOCA analysis satisfies MR5.

MR6 - Reactivity Feedback MR6 required that the SBLOCA analyses include moderator reactivity feedback when the moderator temperature coefficient (MTC) became positive. The MTC could become positive at beginning of core (BOC) conditions and depressurization during an SBLOCA event could cause positive reactivity feedback, resulting in an increase in core power prior to reactor trip. The maximum plausible value of the MTC should be incorporated based on the maximum positive MTC in the TS.

The licensee confirmed (RAI-1 response in Reference 2) that the moderator reactivity feedback at BOC conditions is included in the MPS3 SBLOCA analysis. The simulated reactivity defects were biased to be representative of a core with a BOC hot full-power conditions MTC at the MPS3 TS 3.1.1.3 limit. The MTC represented the most positive (least negative) reactivity feedback achievable as the moderator density decreased, which would ensure that the negative reactivity was minimized as the core voided. The calculations were performed at all rods out conditions, which produced a more positive MTC than would be achieved by modeling the control rod insertion and resulted in less negative reactivity feedback as the core voided. Based on the above, the NRC staff finds that the licensees use of the most positive MTC in the TS for the SBLOCA analysis satisfies MR6.

MR7 - Cross-over Leg Modelling and Loop Seal Clearing MR7 required that: (1) for the SBLOCA analyses, the RCS nodal scheme for S-RELAP5 should increase the number of ((

)). The analysis also should determine the number of loop seals cleared for all other break sizes.

Figure 3-1 of ANP-4031P showed the cross-over leg nodalization for loop-1 RCS noding and indicated that the MPS3 SBLOCA model included two nodes in each of the horizontal and riser segments in the cross-over leg nodalization, which satisfied the above Item (1) of MR7.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The licensee confirmed (RAI-1 response in Reference 2) that the MPS3 SBLOCA model implemented the loop seal biasing ((

)). For MPS3, Note 8 in Table 4-1 and Note 11 in Table 4-2 of ANP-4031P identified that the ((

)), thus satisfying Item (2) of MR7.

Also, Table 4-2 of ANP-4031P provided the loop seal clearing times for all break sizes, which showed the loop seal clearing phenomena were consistent with that described in the methodology, thus satisfying Item (3) of MR7. Based on the above, the NRC staff finds that the cross-over leg nodalization scheme and loops seal clearing model satisfies MR7.

MR8 - Core Nodalization MR8 required that: (1) the ((

] and (2) ((

))

The licensee confirmed (RAI-1 response in Reference 2) that a ((

)). Based on the above, the NRC staff finds that the licensees core model and flow loss coefficient used in the SBLOCA analysis satisfies MR8.

3.2.2.3 Limitations and Conditions in the NRC SER for TR ANP-10349-P-A, Revision 0 The GALILEO methodology in ANP-10349-P-A, Revision 0, GALILEO Implementation in LOCA Methods (Reference 14) was used in the LOCA analysis to determine the initial fuel stored energy, fission gap release, and the transient fuel-cladding gap conductance as an input to S-RELAP5 for the system response calculation.

The limitations and conditions in the SER for the GALILEO code were specified in Section 4.6 of the TR, ANP-10323P-A, Revision 1, GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors (Reference 15). The licensee provided a discussion addressing compliance with each of the limitations in Section 3.5 and Table 3-5 of ANP-4031 (Reference 6) and showed that it satisfied the applicable limitations. The NRC staff reviewed the licensees disposition of the limitations below.

Limitation 1 (L1) - Reactor and Fuel Types L1 restricted the use of the GALILEO code to pressurized water reactor (PWR) designs using the low-enriched Uranium (LEU) fuel loading.

Since the licensee confirmed in its response to L1 in Table 3-5 of ANP-4031 that the SBLOCA analysis was analyzed for MPS3, which is a Westinghouse PWR using LEU fuel, the NRC staff finds that L1 is met.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION L2 - Rod Average Burnup L2 restricted the use of the methodology for rod average burnups to ((

)) for the Zircaloy-4 cladding and ((

)) for the M5 cladding.

The licensee confirmed in its response to L2 that the burnups applied to the analysis in the LAR (Reference 1) for the GAIA fuel with the M5 cladding did not exceed the rod average burnup of

((

)). Therefore, the NRC staff finds that L2 is met.

L3 - Cladding Types L3 restricted the use of the methodology to the cladding types of Zircaloy-4 and M5.

The licensee confirmed in its response to L3 that the analysis in the LAR would support operation for the GAIA fuel with the M5 cladding. Therefore, the NRC staff finds that L3 is met.

L4 - Rod Diameter L4 restricted the use of the methodology for the fuel rod diameter to a range of ((

)) mm to

((

)) mm.

The licensee confirmed in its response to L4 that the analysis in the LAR was performed using fuel with a rod outside diameter of 9.5 mm, which is within the applicable range of ((

)) mm to ((

)) mm. Therefore, the NRC staff finds that L4 is met.

L5 - Pellet Material L5 restricted the use of the methodology to pellet material with Uranium (235U) enrichments up to 5 weight percent.

The licensee confirmed in its response to L5 that the 235U enrichments applied in the analysis of the LAR did not exceed 5 weight percent. Therefore, the NRC staff finds that L5 is met.

L6 - Gadolinia Concentrations L6 restricted the use of the methodology to fuel with Gadolinia concentrations up to 10 weight percent. The licensee confirmed in its response to L6 that the Gadolinia fuel was not analyzed as part of the SBLOCA methodology. Therefore, the NRC staff finds that the L6 for Gadolinia concentrations is not applicable to the MPS3 SBLOCA analysis.

L7 - Initial Fuel Pellet Density L7 restricted the use of the methodology to fuel with initial fuel pellet density for a range of

((

)) of the theorical density (TD) of UO2.

The licensee confirmed in its response to L7 that the analysis in the LAR was performed using the initial pellet density of ((

)) of the TD of UO2, which was within the range of

((

)) of the TD of UO2. Therefore, the NRC staff finds that L7 is met.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION L8 - Fuel Grain Size L8 restricted the use of the methodology to fuel with grain sizes for a range of ((

))

microns.

The licensee confirmed in its response to L8 that the analysis in the LAR was performed using fuel pellets with a grain size of ((

)), which was within the allowable range of ((

)) microns. Therefore, the NRC staff finds that L8 is met.

L9 - Pellet Manufacturing Process L9 restricted the use of the methodology to fuel with pellet manufacturing process using dry conversion (DC) and ammonium diuranate (ADU).

The licensee confirmed in its response to L9 that the fuel pellet manufacturing process for the fuel design considered in the analysis of the LAR was the DC and ADU. Therefore, the NRC staff finds that L9 is met.

The remaining L10 through L13 items specified the limitations for fuel temperature, cladding strain, internal rod pressure, and fuel rod power, respectively. Since those limitations were related to thermo-mechanical methods, the NRC staff finds that L10 through L13 are not applicable to the limitations for the subject LOCA analysis in the LAR.

3.2.3 SBLOCA Analysis - Plant Initial Conditions and Assumptions 3.2.3.1 Acceptance Criteria As stated in Section 3.1 of ANP-4031P, the SBLOCA analysis was performed to meet the acceptance criteria of 10 CFR 50.46(b)(1-4) related to the limits of PCT, MLO, CWO, and coolable core geometry, respectively. For the SBLOCA analysis, the licensee indicated that once the analysis met the 10 CFR 50.46(b)(1-3) criteria, the analysis would satisfy the criterion of 10 CFR 50.46(b)(4) related to the integrity of the core geometry for cooling. The NRC staff finds that the licensees application of the 10 CFR 50.46(b)(1-4) criteria is consistent with that used in the SBLOCA methodology of EMF-2328-P-A, Revision 0. The SBLOCA methodologies indicated that demonstrating compliance with criteria (b)(1), (b)(2), and (b)(3) would ensure satisfaction of criterion (b)(4). Therefore, the NRC staff finds that the licensees application of 10 CFR 50.46(b)(1-4) criteria is acceptable.

3.2.3.2 Break Spectrum Analysis The SBLOCA break spectrum analysis (Section 3.2 of ANP-4031P) was based on break sizes less than or equal to 10 percent of the RCS cold-leg pipe area, which was within the limit allowed for the use of the NRC-approved SBLOCA methodology in EMF-2328-P-A, Revision 0 and therefore, is considered acceptable by the NRC staff. The analysis also assumed that the breaks would occur in the RCS cold-leg pipe on discharge side of the RCP, which was identified in the EMF-2328-P-A, Revision 0 as the limiting break location, resulting in a minimum margin to the limits of 10 CFR 50.46(b)(1-4). Therefore, the NRC staff finds that the break sizes and location considered in the SBLOCA analysis is acceptable.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.2.3.3 LOOP Consideration The analysis (Section 3.4 of ANP-4031P) considered the effect of a loss-of-offsite power (LOOP) on the SBLOCA analysis. The LOOP was assumed to occur concurrently with the reactor scram, which was actuated on the low pressurizer pressure reactor trip signal. The assumption of LOOP concurrent with reactor scram would result in an RCP trip. As discussed in section 3.1.2.2 of this SE, the licensee also performed a delayed RCP trip analysis. The analysis included two cases with assumed delayed times of 1 minute and 5 minutes after reaching the RCP trip criteria for the operator to trip all four RCPs. The results of the analysis for both cases demonstrated compliance with 10 CFR 50.46(b) acceptance criteria regarding PCT, MLO, and CWO. Therefore, the NRC staff finds that the assumption of LOOP is acceptable.

3.2.3.4 Single Failure Considerations To satisfy the single failure (SF) criterion required by Appendix K to 10 CFR 50, the licensee assumed in the SBLOCA analysis (Section 3.4 of ANP-4031P) that one emergency diesel generator (EDG) would fail during a LOCA event. The assumed loss of one EDG would disable one of the two ECCS trains, resulting in the unavailability of one motor-driven auxiliary feedwater (AFW) pump, one high head safety injection (HHSI) pump, one intermediate safety injection (IHSI) pump, and one low head safety injection (LHSI) pump. This assumption would minimize the AFW and SI injection flow for mitigation of the SBLOCA consequences and result in a minimum margin to the acceptance criteria of 10 CFR 50.46(b). Therefore, the NRC staff finds that the assumed loss of one EDG met the worst SF criterion required by 10 CFR 50 Appendix K and is acceptable.

3.2.3.5 Plant Initial Conditions Table 3-1 of ANP-4031P in the LAR listed initial plant conditions used as input in the SBLOCA analysis. During regulatory audits conducted from November 2023 to January 2024, the licensee provided in its portal, documentation for developing the FRM input parameters that were used to support its bases for selection of the initial conditions for the LOCA analysis. The NRC staff performed an audit on the documentation including responses to audit questions, MPS3 plant data book, MPS3 LOCA inputs, GAIA transition calculation, and ECCS surveillance requirements. The NRC staff found that the bases used by the licensee to select the numerical values in Table 3-1 had provided reasonable assurance that the inputs for the SBLOCA analysis would adequately represent the MPS3 system performance capacity and/or TS requirements, which would lead to a minimum margin to the 10 CFR 50.46 (b)(1-4) criteria. Therefore, the values for the inputs were found to be conservative and is considered acceptable by the NRC staff.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The following inputs were based on the ((

)). Their impacts on the results of the SBLOCA analysis were discussed as follows.

1. The pressurizer pressure was assumed at the nominal RCS operating pressure without measurement uncertainty. The licensee indicated in RAI-4 response that since ((

)).

2. The accumulator water volume per accumulator was assumed to be the ((

)). The SBLOCA break spectrum analysis in Table 4-2 of ANP-4031P showed that for all small-break cases analyzed, the ((

)). This analysis demonstrated that the assumption of the ((

)) would not affect the PCT during an LOCA.

3. The main feedwater (MFW) temperature was based on the nominal operating temperature. The licensee indicated in the RAI-4 response that the ((

)) In the SBLOCA analysis, the MFW would be tripped at the time of the breaks and therefore, it had no function other than setting plant model initial conditions.

4. The steam generator (SG) secondary pressure was assumed at the ((

)). The licensee indicated in its response to RAI-4 (Reference 2) that the SG secondary pressure was related to the SS initialization inputs. The SS initialization inputs also included the initial RCS fluid temperature, ((

)). Following reactor/turbine trip during an SBLOCA event, the SG secondary pressure would rise rapidly to the main steam safety valve (MSSV) setpoint pressure, ((

)). The transient RCS temperatures ((

)), particularly at the time of PCT. The licensee also estimated that based on the thermal-hydraulic response in SGs for SBLOCA cases with and without actuation of the MSSVs, the use of a biased initial SG with the goal of minimizing the margin to the 10 CFR 50.46(b) would have 0 °F impact on the PCT results of the SBLOCA break spectrum for MPS3.

5. The assumed AFW temperature was based on the average of the maximum and minimum operating temperatures. The licensee indicated in the RAI-4 response that the AFW temperature was initiated to maintain liquid level and mass in the operable SG.

Since the heat sink temperature was a function of the SG secondary pressure and temperature, both parameters would change rapidly as SG pressure increased to the MSSV setpoint pressure. ((

)). Also,((

)). In addition, the motor driven AFW injection was assumed to begin at a conservative low SG level (0 percent narrow

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION range), reducing the SG cooling function. The licensee estimated that based on the thermal-hydraulic response in SGs for SBLOCA cases with and without actuation of the MSSVs, that the use of a biased initial AFW temperature with the goal of minimizing the margin to the 10 CFR 50.46(b) would have 0 °F impact on the PCT results of the SBLOCA break spectrum for MPS3.

Based on the above, the NRC staff determined that: (1) the values for a majority of the key parameters used in the SBLOCA adequately represented the MPS3 system performance capacity and/or TS requirements, leading to a minimum margin to the 10 CFR 50.46(b) criteria and were therefore conservative; (2) the use of ((

)) for some parameters (including pressurizer pressure, accumulator water volume, MFW temperature, SG secondary pressure and AFW temperature) would have negligibly small effects on PCT during an SBLOCA event; and (3) the limiting SBLOCA case with the PCT of 1622 °F, which was an 8.6-inch diameter RCS cold-leg pump discharge break, showed that a margin of 578 °F (2200 - 1622) to the PCT limit of 2200 °F was available. This result provides reasonable assurance that the available margin is significant to adequately compensate for PCT uncertainties associated with use of ((

)) for above item (2) plant parameters and thus, the limiting SBLOCA case would remain valid.

Therefore, the NRC staff finds that the values of the initial plant parameters in Table 3-1 of ANP-4031P are acceptable for use in the SBLOCA analysis.

3.2.4 SBLOCA Analysis - Results The SBLOCA break spectrum study in Section 4.2 and the attached piping break analysis in Section 4.3 of ANP-4031P showed that: (1) the limiting PCT break from the RCS cold-leg discharge break spectrum was the 8.60-inch diameter break with a PCT of 1622 °F, and (2) the HHSI line break analysis produced the limiting total MLO and limiting CWO values of 4.28 percent and 0.08 percent, respectively. Since the calculated limiting PCT, MLO and CWO meets the applicable criteria for 10 CFR 50.46(b)(1-3), respectively, the integrity of the core geometry required by 10 CFR 50.46(b)(4) is maintained. Therefore, the NRC staff finds the SBLOCA analysis acceptable for MPS3 licensing applications.

3.2.5 Conclusion for the SBLOCA Analysis Based on its evaluation in sections 3.1.1 through 3.1.4 above, the NRC staff finds that: (1) the SBLOCA analysis appropriately used the NRC-approved EM; (2) the analysis satisfactorily meets the limitations or requirements in the SERs for the EM; (3) the analysis appropriately used the values of key input parameters and assumptions that adequately reflected the limiting TS values or represented plant operating data; (4) for those assumed initial conditions that had not included measurement uncertainties, the effects of adding measurement uncertainties to those conditions in the SBLOCA analysis would be negligibly small; and (5) the results of the analysis are within the limits of the applicable 10 CFR 50.46(b)(1-4) requirements with significant margins. Therefore, the NRC staff finds that the SBLOCA analysis provides reasonable assurance that the use of the NRC-approved SBLOCA EM in TRs, EMF-2328-P-A, Revision 0 (Reference 8), including Supplement 1-P-A, Revision 0 (Reference 10), and ANP-10349-P-A, Revision 0 (Reference 14), is acceptable to support the fuel transition to GAIA fuel with M5 cladding at MPS3.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.3 RLBLOCA Analysis The NRC staff reviewed the RLBLOCA analysis based on the information documented in ANP-4032P (Reference 7), and focused the review on the following subjects: (1) analytical methods used for the RLBLOCA analysis; (2) compliance with the limitations imposed in NRC SER for use of the methods; (3) acceptable values of plant parameters used as input to the analytical models; (4) compliance of the analytical results with the 10 CFR 50.46(b) acceptance criteria; and (5) conclusion for the RLBLOCA analysis. The NRC staffs evaluation of the RLBLOCA analysis are discussed in sections 3.2.1 through 3.2.5, as follows.

3.3.1 RLBLOCA Analytical Methods As discussed in Section 3.3 of ANP-4032P (Reference 7), the RLBLOCA analysis was performed based on the methods in the NRC-approved FRM TRs, EMF-2103-P-A, Revision 3, Realistic Large Break LOCA Methodology for Pressurized Water Reactors (Reference 12),

and ANP-10349-P-A, Revision 0, GALILEO Implementation in LOCA Methods (Reference 14).

The RLBLOCA model consisted of two computer codes (Section 3.3 of ANP-4032P): S-RELAP5 (in EMF-2103-P-A, Revision 3) for the thermal-hydraulic calculations and GALILEO (in ANP-10349-P-A, Revision 0) for determining the initial fuel stored energy, fission gap release and, the transient fuel-cladding gap conductance as an input to S-RELAP5 for the system response calculation. The RLBLOCA model was a best estimate (BE) model that followed the code scaling, applicability and uncertainty (CSAU) evaluation methodology (Reference 16) and the requirements of the evaluation model development and assessment process (Reference 17). The CSAU method discussed an approach for defining and qualifying a BE thermal-hydraulic code and quantified the uncertainties in a LOCA analysis. The NRC staffs generic approval of the FRM RLBLOCA model is documented in References 12 and 13 for use of the S-RELAP5 code and References 14 and 15 for use of the GALILEO code.

3.3.1.1 Mixed Core Simulation Methodology The MPS3 RLBLOCA analysis ((

)) for use in the MPS3 RLBLOCA analysis.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.3.1.2 Nuclear Fuel Thermal Conductivity Degradation Effects NRC Information Notice 2009-23 (Reference 20), Nuclear Fuel Thermal Conductivity Degradation (TCD), discussed an issue related to the ability of thermal-mechanical fuel modeling codes to predict the exposure-dependent degradation of fuel thermal conductivity accurately. A safety concern with TCD in a LOCA would be that fuel temperatures modeled incorrectly would affect the initial stored energy, resulting in the LOCA evaluation model to underpredict PCTs. The NRC staff found that the models implemented in GALILEO explicitly considered the effects of fuel pellet TCD. Specifically, Section 6.0 of the NRC SER for TR, ANP-10323-P-A (Reference 15) indicated that the NRC staff had reviewed the thermal model, including TCD effects, and concluded that the GALILEO TR was acceptable for referencing in licensing application for FRM PWR fuel design. Based on the above, the NRC staff determined that the TCD issue identified in NRC Information Notice 2009-23 was satisfactorily resolved for the MPS3 RLBLOCA analysis.

3.3.2 RLBLOCA Methods - Compliance with NRC SER Limitations 3.3.2.1 Limitations in the NRC SE for TR EMF-2103-P-A, Revision 3 The NRC SER (Reference 13) for the TR, EMF-2103-P-A, Revision 3. imposed in its Section 4.0 eleven limitations on use of the RLBLOCA methods. The licensees discussion of its compliance with each of the limitations was discussed in Section 3.7 and Table 3-1 of ANP-4032P (Reference 7), showing that the licensee fully satisfied each of the limitations for use of the methodologies in EMF-2103(P)(A), Revision 3 for the RLBLOCA analysis. The NRCs staffs evaluation of the compliance discussion is provided as follows.

L1 - Acceptance Criteria L1 allowed the use of the EM in the TR EMF-2103-P-A, Revision 3 for determining whether plant-specific results complied with the acceptance criteria the ECCS performance set forth in 10 CFR 50.46(b)(1-3) only, and not for determining whether the acceptance criteria of 10 CFR 50.46(b)(4-5) were met.

The NRC staff reviewed ANP-4032P (Reference 7) in the LAR and found in Sections 3.1 and 4.2 of the TR, the RLBLOCA analysis was performed using the EM methods to meet the criteria of 10 CFR 50.46(b)(1-3). Based on the above, the NRC staff finds that the RLBLOCA analysis meets L1.

L2 - Plant Data Applicability L2 restricted the use of the EM in the TR EMF-2103-P-A, Revision 3, to application for 3-loop and 4-loop Westinghouse designed nuclear steam supply systems (NSSSs), and to Combustion Engineering-designed NSSSs with cold-leg ECCS injection, only. The RLBLOCA analysis in ANP-4032P was performed for MPS3. Since MPS3 is a 4-loop Westinghouse designed NSSS with cold-leg ECCS injection, the NRC staff finds that the RLBLOCA analysis for MPS3 satisfies L2.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION L3 - Fuel Cladding L3 restricted the use of the EM in the TR EMF-2103-P-A, Revision 3, to FRM M5 fuel cladding.

The licensee confirmed in its response to L3 in Table 3-1 of ANP-4032P that the RLBLOCA analysis was performed using the M5 cladding data as input. Therefore, the NRC finds that the RLBLOCA analysis satisfies L3.

L4 - Modeling Guidelines L4 required that the plant-specific RLBLOCA analyses follow the modeling guidelines contained in Appendix A to EMF-2103, Revision 3, and include a statement summarizing the extent to which the guidelines were followed and a justification for any deviations.

The licensee confirmed in its response to L4 that the RLBLOCA analysis presented in ANP-4032P completely followed the modelling guidelines contained in Appendix A of EMF-2103-P-A, which contained Table A-6 listing key parameters and associated probability distribution functions (PDFs) for consideration in the RLBLOCA analysis and Table A-7 listing the model parameter uncertainty ranges and associated PDFs. In addition, Figure 4-1 in ANP-4032P, showed linear scatter plots of the key parameters sampled for all RLBLOCA cases, demonstrated that the spread and coverage of all the values used adequately reflected that the PDFs were consistent with that in Table A-6 of EMF-2103-P-A, and that the parameter ranges were within the uncertainty ranges listed in Table 4-2 in ANP-4032P. Based on the above, the NRC staff finds that the RLBLOCA analysis satisfies L4.

L5 - Burnup Limitation L5 restricted the application of the EM in TR EMF-2103, Revision 3 to currently licensed fuel burnup limits of the rod average burnup of ((

)).

The licensee confirmed in its response to L5 that the RLBLOCA analysis was performed for fuel burnups within the rod average burnup limit of ((

)). Therefore, the NRC staff finds that the analysis satisfies L5.

L6 - Pellet Relocation Packing Factor Data Set The fuel pellet relocation packing factor was derived from currently available data. L6 stated that should new data become available to suggest that fuel pellet fragmentation behavior was other than that suggested by the currently available database, the NRC would request FRM-AREVA to update its model to reflect such new data.

The NRC had not sent a letter to FRM (formally AREVA) requesting an update to the application for new data and the licensee confirmed in its response to L6 that the RLBLOCA analysis was performed using the relocation packing factor in the approved TR EMF-2103, Revision 3. Based on the above, the NRC staff finds that the analysis satisfies L6.

L7 Percent Cathcart-Pawel (C-P) Equivalent Cladding Reacted (ECR)

The acceptance criterion of 10 CFR 50.46(b)(2), requiring cladding oxidation not to exceed 17 percent of the initial cladding thickness prior to oxidation, was based on the use of the Baker-Just oxidation correlation. To account for the use of the C-P correlation, L7 required a

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION reduction in the cladding oxidation limit from 17 percent to 13 percent, inclusive of pre-transient oxide layer thickness.

Table 4-4 of ANP-4032P included the calculated MLO, which was 6.47 percent and was within the limit of 13 percent. Therefore, the NRC staff finds that the RLBLOCA analysis satisfies L7.

L8 Percent C-P ECR In conjunction with L7 above, L8 stated that C-P oxidation results would be considered acceptable, provided plant-specific ((

)).

The licensee confirmed in its response to L8 that for all ((

)). Therefore, the NRC staff finds that the analysis satisfies L8.

L9 - Uncertainty Treatment for Plant Parameters L9 stated that the uncertainty treatment for plant parameters would be considered acceptable if plant parameters were ((

)), as appropriate. Alternative approaches could be used, provided they were supported with appropriate justification.

The licensee confirmed in its response to L9 that ((

)). Based on the above, the NRC staff finds that the analysis satisfies L9.

L10 - ((

))

L10 stated that ((

)).

The licensee confirmed in its response to L10 that the RLBLOCA analysis did not use

((

)). Based on the above, the NRC staff finds that L10 is not applicable to RLBLOCA analysis for MPS3.

L11-Reanalysis L11 required that any plant submittal to the NRC using EMF-2103-P-A, Revision 3, which was not based on the first statistical calculation intended to be the analysis of record, would have to state that a re-analysis had been performed and identified the changes that were made to the EM and/or input to obtain the results in the submitted analysis.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The licensee confirmed in its response to L11 that the present RLBLOCA analysis was based on the first statistical application of TR EMF-2103-P-A, Revision 3. Based on the above, the NRC staff finds that L11 is not applicable to the analysis for MPS3.

3.3.2.2 Limitations in the NRC SER for TR ANP-10349-P-A, Revision 0 The GALILEO methodology in ANP-10349-P-A, Revision 0 (Reference 14) was used in the LOCA analysis to determine the initial fuel stored energy, fission gap release and the transient fuel-cladding gap conductance as an input to S-RELAP5 for the system response calculation.

The limitations for the use of the GALILEO code were specified in Section 4.6 of the TR, ANP-10323P-A, Revision 1, GALILEO Fuel Rod Thermal-Mechanical Methodology for Pressurized Water Reactors (Reference 15). The licensee provided its discussion of compliance with each of the limitations of the SER for the use of GALILEO methodology in Section 3.7 and Table 3-2 of ANP-4032P (Reference 7) for the MPS3 RLBLOCA analysis. Since the same GALILEO methodology was used for both SBLOCA and RLBLOCA analysis, the SER limitations were identical. The licensees discussion of its compliance with the SER limitations, except for Limitation 6 regarding Gadolinia concentration limits, in Section 3.7 and Table 3-2 of ANP-4032P for the RLBLOCA analysis was the same as that discussed in Section 3.5 and Table 3-5 of ANP-4031P for the SBLOCA analysis. The NRC staff determined that SE section 3.1.2.3, concluding that the licensee satisfied each of the limitations, except for Limitation 6, for use of GALILEO in the SBLOCA analysis, was applicable to the RLBLOCA analysis.

Regarding Limitation 6 that restricted the Gadolinia concentrations up to 10 weight percent, the licensee confirmed in its response to Limitation 6 in Table 3-2 of ANP-4032P that the Gadolinia concentration analyzed for the RLBLOCA analysis did not exceed 10 weight percent. Based on the above, the NRC staff finds that Limitation 6 is met for the RLBLOCA analysis.

3.3.3 RLBLOCA Analysis - Input Parameters 3.3.3.1 Acceptance Criteria The RLBLOCA analysis (Section 3.1 of ANP-4032P) was performed to meet the acceptance criteria of 10 CFR 50.46(b)(1-3) related to the limits of PCT, MLO, and CWO. EMF-2103-P-A, Revision 3 was not used to demonstrate compliance with criterion (b)(4). Criterion (b)(4) was addressed by evaluating the GAIA fuel assemblies response under postulated LOCA blowdown loads via a separate license amendment request (Reference 21). EMF-2103-P-A, Revision 3 was not used to demonstrate compliance with criterion (b)(5). The discussion of long-term-cooling (LTC) following a LOCA to meet criterion (b)(5) is included in Chapter 6 of FSAR for MPS3. The LTC analyses are not dependent on the specific fuel designs in the core.

3.3.3.2 LOOP Consideration To satisfy GDC 35, requiring that an ECCS be designed to provide abundant core cooling flow for conditions with and without a loss-of-offsite power (LOOP), the licensee (Section 3.4 of ANP-4032P) ((

)). For the RLBLOCA analysis, ((

)). The NRC staff finds the assumed

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION LOOP and no-LOOP conditions in the RLBLOCA analysis meets the GDC 35 requirements, and therefore, is acceptable.

3.3.3.3 Single Failure (SF) Considerations To satisfy the SF criterion, the licensee assumed in the RLBLOCA analysis that one train of pumped ECCS injection would fail during a LOCA event. The assumed loss results in one HHSI pump, one IHSI pump, and one LHSI pump being unavailable. The minimum flow rates were shown as items 3.n, 3.m, and 3.l, respectively, in Table 4-1 of ANP-4032P. The assumed minimum SI flow rates were lower bound flow rates reflective of a single train injection representing only one HHSI pump, one IHSI pump, and one LHSI pump. Those assumed SI flow rates were consistent with those in the FRM methodologies in ((

)), and would result in a minimum margin to the acceptance criteria of 10 CFR 50.46(b).

Therefore, the NRC staff finds that the assumed loss of one EDG met the worst SF criterion required by 10 CFR 50 Appendix K and is acceptable.

3.3.3.4 Initial Plant Conditions The RLBLOCA analysis for MPS3 applied the code input development guidelines in Appendix A of EMF-2103-P-A, Revision 3 (Reference 12) to assure that the model nodalization scheme was consistent with that used in the code validation. The statistical approach was used to create sampled cases. For every set of input created, each key LOCA parameter listed in Table A-6 of EMF-2103-P-A, Revision 3 was randomly sampled over a range established through code uncertainty assessment or expected operating limits provided by TS or plant data. Table A-6 included both parameters related to LOCA phenomena, based on phenomena identification and ranking table (PIRT) in EMF-2103-P-A, Revision 3, and to plant operational parameters. A statistical approach was used to determine that the first three criteria of 10 CFR 50.46(b)(1-3),

PCT, MLO, and CWO, were met with a probability higher than 95 percent with 95 percent confidence (95/95).

The initial plant conditions assumed for the RLBLOCA analysis were listed in Table 4-1 of ANP-4032P. The analysis assumed (Section 1.0 of ANP-4032P) the full-power operation at a core power level of 3723 mega-watt thermal (MWt) including measurement uncertainty, a maximum-allowed total peaking factor of 2.6, a radial peaking factor of 1.70 (including uncertainty) and SG tube plugging up to 10 percent per SG. As discussed in section 3.2.3 of this SE under the subject of single failure considerations, the assumed SI flow rates would result in a minimum margin to the acceptance criteria of 10 CFR 50.46(b) and is therefore, considered conservative and acceptable by the NRC staff.

Also, the analysis considered typical operational ranges or applicable for TS limits for key parameters such as ((

)). Figure 4-1 in ANP-4032P, showing linear scatter plots of the key parameters sampled for all RLBLOCA cases, demonstrated that the spread and coverage of all the values used adequately reflected the PDFs that were consistent with that in Table A-6 of EMF-2103-P-A and the plant parameter ranges were within the uncertainty ranges listed in Table 4-2 in ANP-4032P.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Therefore, the NRC staff finds that the selection of the initial values for the key parameters and associated PDFs used in the RLBLOCA analysis adequately followed the modelling guidelines contained in Appendix A of EMF-2103-P-A and is therefore, acceptable.

3.3.4 RLBLOCA Analysis - Results The RLBLOCA analysis in Section 3.4 of ANP-4032P indicated that ((

)). The results presented in Table 4-4 of ANP-4032P showed that: (1) the limiting PCT and CWO break was Case Number ((

)) of the ((

)) case set with a PCT of 1835 °F and CWO of 0.062 percent; and (2) the limiting MLO break was Case Number ((

)) of the ((

)) case set with an MLO of 6.47 percent. The acceptance criterion of 10 CFR 50.46(b)(2), requiring cladding oxidation not to exceed 17 percent of the initial cladding thickness prior to oxidation, is based on the use of the Baker-Just oxidation correlation. To account for the use of the Cathcart-Pawel (C-P) correlation, the limit of 17 percent is reduced to 13 percent, inclusive of pre-transient oxide layer thickness and ((

)).

3.3.5 Conclusion for the RLBLOCA Analysis Based on its evaluation in sections 3.2.1 through 3.2.4 above, the NRC staff finds that: (1) the RLBLOCA analysis used NRC-approved methods; (2) the analysis satisfactorily meets the limitations or requirements in the SEs for the methods; (3) the analysis used the values of input parameters and assumptions that adequately reflect the limiting TS values or represented plant operating data; and (4) the results of the analysis were within the limits of the applicable 10 CFR 50.46(b)(1-3) requirements. Therefore, the NRC staff finds the RLBLOCA analysis provides reasonable assurance that the associated methods in NRC-approved FRM TRs, EMF-2103-P-A, Revision 3 (Reference 12) and ANP-10349-P-A, Revision 0 (Reference 14) are acceptable for the RLBLOCA analysis to support the fuel transition to GAIA with M5 cladding at MPS3.

3.4 TS Changes 3.4.1 Current TS MPS3 TS 6.9.1.6 Requirement MPS3 TS 6.9.1.6 requires COLs be established for each reload cycle and contains references to the approved analytical methods used to determine the COLs. The TS 6.9.1.6.b COLR reference list includes documents that define the methods used to determine the core operating limits for MPS3. The existing TS 6.9.1.6.b references for LBLOCA and SBLOCA are:

5.

WCAP-16996-P-A, REALISTIC LOCA EVALUATION METHODOLOGY APPLIED TO THE FULL SPECTRUM OF BREAK SIZES (FULL SPECTRUM LOCA METHODOLOGY), (W Proprietary)

(Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel Factor.)

6.

WCAP-16009-P-A, REALISTIC LARGE-BREAK LOCA EVALUATION METHODOLOGY USING THE AUTOMATED STATISITICAL TREATMENT OF UNCERTAINTY METHOD (ASTRUM), (W Proprietary). (Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel Factor.)

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

8.

WCAP-10054-P-A, WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL USING THE NOTRUMP CODE, (W Proprietary). (Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel Factor.)

9.

WCAP-10079-P-A, NOTRUMP - A NODAL TRANSIENT SMALL BREAK AND GENERAL NETWORK CODE, (W Proprietary).

(Methodology for Specification 3.2.2.1 - Heat Flux Hot Channel Factor.)

16. WCAP-8301, LOCTA-IV Program: Loss-of-Coolant Transient Analysis. Methodology for Specification:

• 3.2.2.1 Heat Flux Hot Channel Factor

17. WCAP-10054-P-A, Addendum 2, Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code:

Safety Injection into the Broken Loop and COSI Condensation Model.

Methodology for Specification:

• 3.2.2.1 Heat Flux Hot Channel Factor 3.4.2 Proposed TS Changes As stated in the LAR:

The proposed TS changes are needed to support the transition to FRM GAIA fuel with M5TM cladding at MPS3, which requires the application of the FRM SBLOCA and RLBLOCA methodologies. DENC and FRM have entered into an agreement for batch implementation of the GAIA fuel at MPS3. A full reload batch of GAIA fuel assemblies is planned for initial insertion in Cycle 24. This onload is currently scheduled for the spring 2025 refueling outage.

Additionally, the proposed changes amend MPS3 TS 6.9.1.6.b by removing a legacy COLR reference no longer use to establish core operating limits.

TS 6.9.1.6.b lists methodology documents used to determine the core operating limits for MPS3. 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, the TS 6.9.1.6.b list only identifies the topical report document identification number and title. The proposed revision to this list adds three documents in support of the FRM SBLOCA and RLBLOCA methods, as shown below.

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

• 3.2.2.1 Heat Flux Hot Channel Factor

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

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 The proposed revision to the list also deletes the WCAP methodology listed as document no. 6 under the current TS listing above. The proposed change would replace the reference with Deleted. The LAR proposed to retain the WCAP methodologies listed as document nos. 5, 8, 9, 16 and 17 to support the current MPS3 fuel product prior to and during the transition to GAIA.

3.4.3 NRC Staff Evaluation of TS changes As discussed in sections 3.1 and 3.2 of this SE, the NRC staff has determined the use of the methods in the above FRM TRs for the SBLOCA and RLBLOCA analysis is acceptable to support the fuel transition to GAIA with M5 cladding at MPS3. The NRC staff has confirmed that the use of the methodologies is within the NRC-approved ranges of applicability and the results of the analyses are in compliance with the applicable regulatory requirements.

The guidance in NRC GL 88-16, Removal of Cycle-Specific Parameter Limits from Technical Specifications, outlines a process that licensees can use to move cycle-specific parameters from the plant specific TSs to a licensee-controlled document entitled the COLR. The guidance also specifies that the analytical methods used to determine the operating limits are those previously reviewed and approved by the NRC and documented in the TR(s).

The NRC staff has determined that the FRM TRs would be appropriately referenced in the MPS3 Administrative Controls section (TS 6.9.1.6.b) for use as the applicable methods to determine the MPS3 core operating limits. The format of the added FRM TRs, containing the TR number, title, and applicable cycle-specific parameters, is consistent with that of other TRs in the current TS 6.9.1.6.b. The NRC staff also finds that the proposed deletion of the Westinghouse TR from the TS (listed as document no. 6 under TS 6.9.1.6.b) is acceptable since this TR is no longer used to define the core operating limits in the COLRs. Therefore, based on the above, the NRC staff finds that the proposed changes to the MPS3 TSs are consistent with the guidance in NRC GL 88-16 and the requirements of 10 CFR 50.36(c)(5), and are therefore, acceptable.

3.5 Technical Conclusion The NRC staff reviewed the LAR in conjunction with additional and supplemental information listed in various sections of this SE related to the proposed changes to TS 6.9.1.6.b in support of the transition to FRM GAIA fuel with M5TM cladding at MPS3 and the resulting application of the FRM LOCA methodologies.

Based on its review, as summarized in various sections of this SE, the NRC staff concludes that the licensee provided adequate technical basis to support the proposed TS changes.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Specifically, the NRC staff finds the licensee has demonstrated that (1) the analysis was performed using the NRC-approved FRM LOCA methods, (2) the analysis complies with staff limitations imposed for application of the TRs where applicable for the LOCA methods, (3) the results meet the applicable acceptance criteria in 10 CFR 50.46, and (4) 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 April 18, 2024. The State official had no comments.

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 August 1, 2023 (88 FR 50183), 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, 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, ADAMS (ADAMS) Accession No. ML23123A279.

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 License Amendment Request to Use Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits, April 1, 2024, ML24093A216.

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

3.

Guzman, R., U.S. Nuclear Regulatory Commission, letter to Carr, E., Dominion Energy Nuclear Connecticut, Inc., Millstone Power Station, Unit No. 3 - Exemption from the Requirements of 10 CFR Part 50, Section 50.46 and Appendix K Regarding Use of M5 Cladding Material (EPID L-2023-LLE-0013), May 21, 2024, Package ML24110A058.

4.

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.

5.

U.S. Nuclear Regulatory Commission, Removal of Cycle-Specific Parameter Limits from Technical Specifications, Generic Letter 1988-16, October 4, 1988, ML031200485.

6.

ANP-4031, Revision 0, Millstone Unit 3 Small Break LOCA Analysis with GAIA Fuel Design, Licensing Report, Framatome, Inc., May 2, 2023, ML23123A277 (Proprietary Version in Attachment 3 to the LAR), ML23123A279 (Non-Proprietary Version in to the LAR).

7.

ANP-4032, Revision 0, Millstone Unit 3 Realistic Large Break LOCA Analysis with GAIA Fuel Design, Licensing Report, Framatome, Inc., May 2, 2023, ML23123A277 (Proprietary Version in Attachment 5 to the LAR), ML23123A279 (Non-Proprietary Version in Attachment 6 to the LAR).

8.

EMF-2328-P-A, Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, Framatome ANP Richland, Inc., March 2001, ML011410417 (Proprietary),

ML011410383 (Non-Proprietary).

9.

Richards, S., U.S. Nuclear Regulatory Commission, letter to Mallay, J., Framatome ANP Richland, Inc., Acceptance for Referencing of Licensing Topical Report EMF-2328(P),

Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, March 15, 2001, ML010800365.

10.

EMF-2328-P-A, Revision 0, Supplement 1-P-A, Revision 0, PWR Small Break LOCA Evaluation Model S-RELAP5 Based, AREVA NP, Inc., December 2016, Package ML16356A396.

11.

U.S. Nuclear Regulatory Commission, Revised Final Safety Evaluation by the Office of Nuclear Reactor Regulation for Topical Report EMF-2328(P)(A), Revision 0, Supplement 1, Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, December 13, 2016, Package ML16313A327 (Proprietary).

12.

Peters, G., AREVA, Inc. to U.S. Nuclear Regulatory Commission, EMF-2103-P-A Revision 3, Realistic Large Break LOCA Methodology for Pressurized Water Reactors, June 2016, Package ML16286A579.

13.

Hsueh, K., U.S. Nuclear Regulatory Commission, letter to Peters, G., AREVA, Inc.,

Final Safety Evaluation for Topical Report EMF-2103(P), Revision 3, Realistic Large Break LOCA Methodology for Pressurized Water Reactors, June 17, 2016, ML16286A319 (Proprietary).

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

14.

ANP-10349-P-A, Revision 0, GALILEO Implementation in LOCA Methods, Framatome, Inc., Package ML20290A661.

15.

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

16.

NUREG/CR-5249, Quantifying Reactor Safety Margins, Application of Code Scaling, Applicability, and Uncertainty Evaluation Methodology to a Large Break, Loss-of-Coolant Accident, U.S. NRC, December 1989, ML030380473.

17.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.203, Transient and Accident Analysis Methods, December 2005, ML053500170.

18.

Guzman, R, U.S. Nuclear Regulatory Commission, Millstone Power Station, Unit 3 -

Setup of Online Reference Portal and Audit Plan for the NRC staff's review of LAR to Use Framatome Small Break and Realistic Large Break LOCA Methodologies, August 31, 2023, ML23258A055.

19.

Guzman, R, U.S. Nuclear Regulatory Commission, Millstone Power Station, Unit 3 -

Summary of Regulatory Audit in Support of License Amendment Request to Use Framatome Small Break and Realistic Large Break Loss of Coolant Accident Evaluation Methodologies for Establishing Core Operating Limits, April 22, 2024, ML24103A020.

20.

NRC Information Notice 2009-23: Nuclear Fuel Thermal Conductivity Degradation, Brach, E. W, McGinty, T. J., Tracy, G. M., October 8, 2009, ML091550527.

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

21.

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, ADAMS (ADAMS) Accession No. ML23145A195.

Principal Contributor: S. Sun, NRR Date: May 21, 2024