ML23144A314

Issuance of Amendment Nos. 281 and 274 Transition to GNF3 Fuel (EPID L-2022-LLA-0121) - Nonproprietary

ML23144A314
Person / Time
Site:	Dresden
Issue date:	07/06/2023
From:	Shilpa Arora Plant Licensing Branch III
To:	Rhoades D Constellation Energy Generation, Constellation Nuclear
Arora, S
References
EPID L-2022-LLA-0121
Download: ML23144A314 (1)

Text

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION July 6, 2023 Mr. David P. Rhoades Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 - ISSUANCE OF AMENDMENT NOS. 281 AND 274 RE: TRANSITION TO GNF3 FUEL (EPID L-2022-LLA-0121)

Dear Mr. Rhoades:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 281 to Renewed Facility Operating License No. DPR-19 and Amendment No. 274 to Renewed Facility Operating License No. DPR-25 for Dresden Nuclear Power Station, Units 2 and 3 (Dresden). The amendments consist of changes to the technical specifications (TSs) in response to your application dated August 18, 2022 (Agencywide Documents Access and Management System Accession No. ML22230C927), as supplemented by letters dated January 23, 2023 (ML23023A128) and April 17, 2023 (ML23107A228).

The amendments support the transition from Framatome, Inc. ATRIUM 10XM fuel to Global Nuclear Fuel - Americas, LLC (GNF/GNF-A) GNF3 fuel at Dresden, Units 2 and 3. Specifically, the amendment revises TS 5.6.5, Core Operating Limits Report (COLR), paragraph b, to remove the eight Westinghouse topical reports that will no longer be used and to add the two reports that support the General Electric Standard Application for Reactor Fuel analysis methodology to the list of approved methods to be used in determining the core operating limits in the COLR. The amendments approve the use of Framatome RODEX2A methodology with an additional thermal conductivity degradation penalty in mixed core thermal-mechanical calculations. Additionally, the amendments revise the alternative source term loss-of-coolant accident analysis to use a bounding core inventory. to this letter contains sensitive unclassified non-safeguards information. When separated from Enclosure 3, this document is DECONTROLLED.

OFFICIAL USE ONLY PROPRIETARY INFORMATION D. Rhoades OFFICIAL USE ONLY PROPRIETARY INFORMATION A copy of the safety evaluation is also enclosed. The Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Surinder S. Arora, Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-237 and 50-249

Enclosures:

1. Amendment No. 281 to DPR-19

2. Amendment No. 274 to DPR-25

3. Proprietary Safety Evaluation

4. Non-Proprietary Safety Evaluation cc: Listserv CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 281 Renewed License No. DPR-19 1.

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

A.

The application for amendment by the Constellation Energy Generation, LLC (the licensee), dated August 18, 2022, as supplemented by letters dated January 23, 2023, and April 17, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-19 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 281, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented during the Unit 2 fall 2023 refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Jeffrey A. Whited, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: July 6, 2023 CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-249 DRESDEN NUCLEAR POWER STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 274 Renewed License No. DPR-25 1.

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

A.

The application for amendment by the Constellation Energy Generation, LLC (the licensee), dated August 18, 2022, as supplemented by letter dated January 23, 2023, and April17, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B. of Renewed Facility Operating License No. DPR-25 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 274, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented during the Unit 2 fall 2023 refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Jeffrey A. Whited, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: July 6, 2023 ATTACHMENT TO LICENSE AMENDMENT NOS. 281 AND 274 RENEWED FACILITY OPERATING LICENSE NOS. DPR-19 AND DPR-25 DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 DOCKET NOS. 50-237 AND 50-249 Replace the following pages of the Renewed Facility Operating Licenses and Appendix A, Technical Specifications, with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.

Remove Insert License DPR-19 License DPR-19 Page 3 Page 3 License DPR-25 License DPR-25 Page 4 Page 4 TSs TSs 5.6-3 5.6-3 5.6-4 5.6-4 5.6-5 5.6-5 Renewed License No. DPR-19 Amendment No. 281 (2)

Constellation Energy Generation, LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear materials as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (3)

Constellation Energy Generation, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

Constellation Energy Generation, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Constellation Energy Generation, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct special nuclear materials as may be produced by the operation of the facility.

C.

This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2957 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 281, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3)

Operation in the coastdown mode is permitted to 40% power.

Renewed License No. DPR-25 Amendment No. 274 f.

Surveillance Requirement 4.9.A.10 - Diesel Storage Tank Cleaning (Unit 3 and Unit 2/3 only)

Each of the above Surveillance Requirements shall be successfully demonstrated prior to entering into MODE 2 on the first plant startup following the fourteenth refueling outage (D3R14).

3.

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

A.

Maximum Power Level The licensee is authorized to operate the facility at steady state power levels not in excess of 2957 megawatts (thermal), except that the licensee shall not operate the facility at power levels in excess of five (5) megawatts (thermal), until satisfactory completion of modifications and final testing of the station output transformer, the auto-depressurization interlock, and the feedwater system, as described in the licensees telegrams; dated February 26, 1971, have been verified in writing by the Commission.

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 274, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

C.

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

D.

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

E.

Restrictions Operation in the coastdown mode is permitted to 40% power.

Reporting Requirements 5.6 Dresden 2 and 3 5.6-3 Amendment No. 281/274 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 3.

The LHGR for Specification 3.2.3.

4.

Control Rod Block Instrumentation Setpoint for the Rod Block MonitorUpscale Function Allowable Value for Specification 3.3.2.1.

5.

The OPRM setpoints for the trip function for SR 3.3.1.3.3 b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel."

2.

NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications," August 1996.

3.

XN-NF-81-58(P)(A) Revision 2 and Supplements 1 and 2, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company, March 1984.

4.

ANF-89-98(P)(A) Revision 1 and Supplement 1, "Generic Mechanical Design Criteria for BWR Fuel Designs,"

Advanced Nuclear Fuels Corporation, May 1995.

5.

EMF-85-74(P) Revision 0 Supplement 1 (P)(A) and Supplement 2 (P)(A), "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model," Siemens Power Corporation, February 1998.

6.

BAW-10247PA Revision 0, "Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors," AREVA NP, February 2008.

7.

XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors -

Neutronic Methods for Design and Analysis," Exxon Nuclear Company, March 1983.

(continued)

Reporting Requirements 5.6 Dresden 2 and 3 5.6-4 Amendment No. 281/274 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

8. XN-NF-80-19(P)(A) Volume 4 Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Company, June 1986.

9. XN-NF-80-19(P)(A) Volume 3 Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," Exxon Nuclear Company, January 1987.

10. EMF-2158(P)(A) Revision 0, "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," Siemens Power Corporation, October 1999.

11. EMF-2245(P)(A) Revision 0, "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel," Siemens Power Corporation, August 2000.

12. EMF-2209(P)(A) Revision 3, "SPCB Critical Power Correlation," AREVA NP, September 2009.

13. ANP-10298P-A Revision 1, "ACE/ATRIUM 10XM Critical Power Correlation," AREVA, March 2014.

14. ANP-10307PA Revision 0, "AREVA MCPR Safety Limit Methodology for Boiling Water Reactors," AREVA NP, June 2011.

15. XN-NF-84-105(P)(A) Volume 1 and Volume 1 Supplements 1 and 2, "XCOBRA-T: A Computer Code for BWR Transient ThermalHydraulic Core Analysis," Exxon Nuclear Company, February 1987.

16. ANF-913(P)(A) Volume 1 Revision 1 and Volume 1 Supplements 2, 3, and 4, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," Advanced Nuclear Fuels Corporation, August 1990.

17. EMF-2361(P)(A) Revision 0, "EXEM BWR-2000 ECCS Evaluation Model," Framatome ANP, May 2001.

(continued)

Reporting Requirements 5.6 Dresden 2 and 3 5.6-5 Amendment No. 281/274 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

18. EMF-2292 (P)(A) Revision 0, "ATRIUMTM-10: Appendix K Spray Heat Transfer Coefficients," Siemens Power Corporation, September 2000.

19. ANF-1358(P)(A) Revision 3, "The Loss of Feedwater Heating Transient in Boiling Water Reactors," Framatome ANP, September 2005.

20. EMF-CC-074(P)(A) Volume 4 Revision 0, "BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2," Siemens Power Corporation, August 2000.

21. NEDC-33930P, Revision 0, "GEXL98 Correlation for ATRIUM 10XM Fuel," Global Nuclear Fuels, February 2021, as approved by the NRC Staff SE dated XXX XX, 20XX.

22. 006N8642-P, Revision 1, "Justification of PRIME Methodologies for Evaluating TOP and MOP Compliance for non GNF Fuels, " Global Nuclear Fuels, January 2022, as approved by the NRC Staff SE dated XXX XX, 20XX.

The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements).

(continued)

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION ENCLOSURE 4 NON-PROPRIETARY SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 281 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-19 AND AMENDMENT NO. 274 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-25 CONSTELLATION ENERGY GENERATION, LLC DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 DOCKET NOS. 50-237 AND 50-249

OFFICIAL USE ONLY PROPRIETARY INFORMATION NON-PROPRIETARY SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 281 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-19 AND AMENDMENT NO. 274 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-25 CONSTELLATION ENERGY GENERATION, LLC DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 DOCKET NOS. 50-237 AND 50-249

1.0 INTRODUCTION

By letter dated August 18, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22230C927) (Reference 1), as supplemented by letters dated January 23, 2023 (ML23023A128) (Reference 2), April 17, 2023 (ML23107A228) (Reference 21), and May 16, 2023 (ML23136B189) (Reference 22), Constellation Energy Generation, LLC (CEG, the licensee), submitted a license amendment request (LAR) for the Renewed Facility Operating License Nos. DPR-19 and DPR-25 for Dresden Nuclear Power Station (DNPS), Units 2 and 3, respectively. The proposed change supports the transition from Framatome, Inc.

ATRIUM 10XM fuel to Global Nuclear Fuel - Americas, LLC (GNF/GNF-A) GNF3 fuel at DNPS, Units 2 and 3.

By letter dated September 14, 2021 (Reference 3), supplemented by letter dated April 11, 2022 (Reference 4), the licensee had submitted a similar LAR for transition from ATRIUM 10XM fuel to GNF3 for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. In addition, by letter dated January 20, 2022 (Reference 5), supplemented by letters dated March 16, 2022 (Reference 6), and August 10, 2022 (Reference 7), the licensee submitted a request to expand the applicability of GNF PRIME methodology (Reference 5) for evaluating the fuel centerline melt and cladding strain compliance for the ATRIUM 10XM fuel at QCNPS, Units 1 and 2. The DNPS, Units 2 and 3, LAR (Reference 1) also includes the application of the expansion of the same GNF PRIME methodology to analyze the ATRIUM 10XM fuel. The licensee proposes to use the GNF PRIME methodology (Reference 5) submitted for QCNPS, Units 1 and 2 to evaluate the fuel centerline melt temperature and cladding strain for the ATRIUM 10XM fuel at DNPS, Units 2 and 3.

The supplement dated January 23, 2023 (Reference 2) provided additional information that clarified the application and did not expand the scope of the application as originally submitted.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The proprietary information in this safety evaluation (SE) input is identified by text in bold font and enclosed within double square brackets. ((This is an example)).

2.0 REGULATORY EVALUATION

2.1 Proposed TS Changes

The licensee proposes to revise the DNPS, Units 2 and 3, Technical Specification (TS) 5.6.5.b, Core Operating Limits Report (COLR), by adding the following two reports as items 21 and 22 to support the General Electric Standard Application for Reactor Fuel (GESTAR) analysis methodology to the list of approved methods to be used in determining the core operating limits in the COLR:

21.

NEDC-33930P, Revision 0, GEXL98 Correlation for ATRIUM 10XM Fuel, Global Nuclear Fuels, February 2021, as approved by the NRC staff SE dated July 5, 2023.

22.

006N8642-P, Revision 1, Justification of PRIME Methodologies for Evaluating TOP and MOP Compliance for non-GNF Fuels, Global Nuclear Fuels, January 2022, as approved by the NRC staff SE dated July 5, 2023.

The licensee proposes to delete the following eight Westinghouse topical reports (TRs) from TS 5.6.5.b that will no longer be used to support the COLR evaluations after the fall outage in 2023.

3.

CENPD-300-P-A, Reference Safety Report for Boiling Water Reactor [BWR] Reload Fuel.

4.

WCAP-16081-P-A, 10x10 SVEA Fuel Critical Power Experiments and CPR [critical power ratio] Correlation: SVEA-96 Optima2.

5.

WCAP-15682-P-A, Westinghouse BWR ECCS [emergency core cooling system]

Evaluation Model: Supplement 2 to Code Description, Qualification and Application.

6.

WCAP-16078-P-A, Westinghouse BWR ECCS Evaluation Model: Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel.

7.

WCAP-15836-P-A, Fuel Rod Design Methods for Boiling Water Reactors -

Supplement 1.

8.

WCAP-15942-P-A, Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors, Supplement 1 to CENPD-287.

9.

CENPD-390-P-A, The Advanced PHOENIX and POLCA Codes for Nuclear Design of Boiling Water Reactors.

10. WCAP-16865-P-A, Westinghouse BWR ECCS Evaluation Model Updates:

Supplement 4 to Code Description, Qualification and Application, Revision 1, October 2011.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 2.2 Plant Type DNPS, Units 2 and 3, are General Electric (GE) BWRs of BWR/3 type. Each unit has a separate primary containment of Mark I type but the units share a common secondary containment (reactor building).

2.3 Regulations and Guidance Documents Regulations The NRC staff considered the following regulations in its review of the proposed changes.

Title 10 of the Code of Federal Regulations (10 CFR) 50.36(c)(5), Administrative controls, requires that provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner must be included in a licensees TS.

10 CFR Part 50, Appendix A:

Updated Final Safety Analysis Report (UFSAR), section 3.1.1, contains the 1967 Atomic Energy Commission (AEC) proposed draft General Design Criteria (GDC) which were used to evaluate compliance of the original design of DNPS, Units 2 and 3. UFSAR, section 3.1.2, contains an evaluation of the design basis of the DNPS, Unit 2 and 3, against the appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, effective May 21, 1971, and subsequently amended July 7, 1971. These GDCs are intended to establish minimum requirements for the design of nuclear power plants.

The regulatory requirements applicable to the fuel design limits are based on 10 CFR part 50, Appendix A, criterion 10, as follows:

Criterion 10- Reactor Design. The reactor core and associated coolant, control and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

DNPS, Units 2 and 3, UFSAR, section 3.1.2.2.1, provides an evaluation that shows that the current fuel design limits meet the intent of the Criterion 10 requirements.

Guidance NUREG-0800, Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition

Section 4.2, Fuel System Design, (ML070740002)

The fuel system safety review, including the fuel thermal-mechanical analysis, provides (in part) reasonable assurance that the fuel system is not damaged as a result of normal operation and anticipated operational occurrences (AOOs) []. According to ANP-3918P (Reference 10), section 1.0, the RODEX2A (References 19, 20, and 21) code is used to

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION analyze the fuel rod for fuel centerline temperature, cladding strain, rod internal pressure, cladding collapse, cladding fatigue, and external oxidation. Each of these parameters has a corresponding limit established in ANF-89-98(P)(A), Revision 1 and Supplement 1 (Reference 20). When satisfied, these limits provide reasonable assurance that the fuel will not be damaged.

Since the RODEX2A methodology was developed in the 1980s, the licensee made some modeling adjustments, which are described in section 3.4 of ANP-3918P, to account for a phenomenon known as thermal conductivity degradation (TCD) in the fuel pellets.

Information Notice (IN) 2009-23, Nuclear Fuel Thermal Conductivity Degradation (ML091550527), discusses how historical fuel thermal mechanical codes like RODEX2A may overpredict fuel rod thermal conductivity at higher burnups based on more recently obtained experimental data.

Section 4.4, Thermal and Hydraulic Design, (ML070550060).

The SRP acceptance criteria lists the following acceptable approaches to ensure that hot fuel rods in the reactor core will not experience a departure from nucleate boiling (DNB) during normal operation or AOOs:

A. For departure from nucleate boiling ratio (DNBR), CHFR [critical heat flux ratio] or CPR correlations, there should be a 95-percent probability at the 95-percent confidence level that the hot rod in the core does not experience a DNB or boiling transition condition during normal operation or AOOs.

B. The limiting (minimum) value of DNBR, CHFR, or CPR correlations is to be established such that at least 99.9 percent of the fuel rods in the core will not experience a DNB or boiling transition during normal operation or AOOs.

2.4 Environmental Qualifications Regulatory Review DNPS, Units 2 and 3, UFSAR, section 3.11, Environmental Qualification of Electrical Equipment, states:

The environmental qualification (EQ) of electrical equipment for Dresden Station is performed per the guidelines of [NRC Office of Inspection and Enforcement Bulletin (IEB)] 79-01B, [Environmental Qualification of Class IE Equipment,] and the requirements of 10 CFR 50.49 [Environmental qualification of electric equipment important to safety for nuclear power plants.].

IEB 79-01B required the licensee to perform a detailed review of the EQ of Class-1E electrical equipment to ensure that the equipment will function under (i.e., during and following) postulated accident conditions.

Regulation 10 CFR 50.49 identifies requirements for establishing a program for qualifying electric equipment that is important to safety as defined in 10 CFR 50.49(b). Section 50.49(e) of 10 CFR requires, in part, that the effects of temperature, pressure, humidity, chemical effects, radiation, and aging be included in the qualification program.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Section 3.11 of NUREG-800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, provides guidance on EQ of mechanical and electrical equipment for complying with 10 CFR 50.49.

3.0 TECHNICAL EVALUATION

The GNF3 and ATRIUM 10XM fuel designs are dimensionally similar, to the extent that both are 10x10 orthogonal fuel pin matrices constrained by the DNPS, Units 2 and 3, fuel lattice design.

The licensee stated that for transition from ATRIUM 10XM fuel to GNF3 fuel at the DNPS, Units 2 and 3, the GE methodologies listed in the NRC-approved TR NEDC-24011-P-A, General Electric Standard Application for Reactor Fuel (GESTAR II, Main), Revision 31 (Reference 8) will be used to determine the core operating limits for cores containing GNF3 fuel.

The licensee submitted the following reports that are applicable to DNPS, Units 2 and 3:

NEDC-33930P, Revision 0 (Reference 9)

ANP-3918P, Revision 0, (Reference 10)

006N8642-P, Revision 1 (Reference 11)

The DNPS, Units 2 and 3, fuel transition to GNF3 consists of evaluation of the following:

GEXL Correlation for ATRIUM 10XM and GNF3 Fuels.

Mixed Core Thermal-Mechanical Analysis Using RODEX2A Methodology

Mixed Core Thermal-Mechanical Analysis Using PRIME Methodology

Containment Analysis

TS Changes 3.1 GEXL Correlation In BWRs, the thermal-hydraulic conditions resulting in an onset to transition boiling (OTB) in fuel have been used as limiting conditions to avoid entering a region where fuel damage could occur. Although it is recognized that a OTB would not necessarily result in damage to the BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a limiting condition. GE has developed critical steam quality versus boiling length correlations for various fuels, designated as the GEXL correlation which is used for accurately predicting the OTB in BWR fuel assemblies during both steady-state operation and reactor transient conditions. The use of the GEXL correlation is necessary for determining the minimum critical power ratio (MCPR) operating limits resulting from transient analysis, the MCPR safety limit analysis, and the core operating performance and design. The GEXL correlation is an integral part of the transient analysis methodology used by GNF. In the core reload design analysis these correlations are used in determining the thermal margin for the operating cycle. In the safety analysis, these correlations are used in determining the change in CPR during postulated transients and an acceptable MCPR safety limit. NEDC-32851P-A, Revision 5 (Reference 16),

Section 5.4 defines the following six input parameters to the GEXL correlation for the calculation of fuel bundle critical power: (a) boiling length, (b) thermal diameter, (c) mass flux, (d) system pressure, (e) R-factor, and (f) annular flow length. As a part of the GESTAR II methodology, the GEXL correlation for a specific fuel requires development of coefficients for the specific lattice geometry and peaking factors used in the fuel assembly design.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.2 GEXL Correlation for ATRIUM 10XM and GNF3 Fuels For the transition mixed core, a GEXL correlation is required to analyze the Framatome ATRIUM 10XM fuel with the GE methodology. Report NEDC-33930P, submitted as a part of the QCNPS, Units 1 and 2, fuel transition LAR (Reference 3) describes the development of the specific GEXL98 correlation for the ATRIUM 10XM fuel which has been approved by the NRC staff for application to QCNPS, Units 1 and 2, application (Reference 12). For the mixed core design and safety analysis, the licensee proposes to use the GEXL98 correlation to predict the expected critical power performance of the ATRIUM 10XM fuel. In a letter dated May 16, 2023 (Reference 22), the licensee stated:

The GNF GEXL98 correlation developed is applicable to ATRIUM 10XM and mixed cores of GNF3 and ATRIUM 10XM. DNPS [Units 2 and 3] is an equivalent BWR/3, designed and operated in the same manner as QCNPS [Units 1 and 2]. Both DNPS

[Units 2 and 3] and QCNPS [Units 1 and 2] share the same analyzed power to flow maps, thermal hydraulic characteristics, core size, and rated core thermal power. DNPS

[Units 2 and 3] and QCNPS [Units 1 and 2] use the same ATRIUM 10XM design, and therefore the results from GNF Report NEDO-33930P [NEDC-33930P] are applicable.

The NRC staff has reviewed the above statement and finds it acceptable. Since the report NEDC-33930P has been accepted by the NRC staff for application to QCNPS, Units 1 and 2, (Reference 12), it is also applicable to DNPS, Units 2 and 3. The GEXL98 correlation developed in NEDC-33930P is acceptable for predicting the critical power performance of ATRIUM 10XM fuel in the mixed core of DNPS, Units 2 and 3.

For the GNF3 fuel in the mixed core, the licensee intends to apply the appropriate NRC-approved GEXL correlation to determine the expected thermal margin and acceptable MCPR safety limit for the mixed core. NEDC-33880P (Reference 15) which is included in GESTAR II provides the GEXL correlation for the GNF3 fuel.

3.3 Mixed Core Thermal-Mechanical Analysis Using RODEX2A Methodology The NRC-approved Framatome RODEX4 (BAW-10247PA, Revision 0), (Reference 17),

RODEX2 (XN-NF-58(P)(A), Revision 2), (Reference 18), and RODEX2A (EMF-85-74P),

(References 19), thermal-mechanical analysis methodologies for ATRIUM 10XM fuel are currently included in DNPS, Units 2 and 3, TS 5.6.5.b. The licensee stated that RODEX4 is currently used for the DNPS, Units 2 and 3, fuel thermal-mechanical analysis. The licensee submitted ANP-3918P as attachment 10 to the QCNPS, Units 1 and 2, LAR (Reference 3) and is included as Reference 6.13 in the DNPS, Units 2 and 3, LAR. This report documents the fuel rod thermal-mechanical calculations performed for ATRIUM 10XM fuel that will be in the mixed core with GNF3 fuel in DNPS, Units 2 and 3, during the transition cycle. As described in ANP-3918P, using the RODEX2A code modified by including the TCD penalty, the licensee evaluated the following fuel rod design items:

Cladding Collapse

Overheating of Fuel Pellets

Stress and Strain Limits (pellet/cladding interaction and cladding stress)

Fuel Densification and Swelling, Fatigue

Cladding Oxidation

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Hydriding and Crud Buildup

Rod Internal Pressure The analysis results are evaluated according to the generic fuel rod thermal and mechanical design criteria contained in ANF-89-98(P)(A), Revision 1, and Supplement 1, (Reference 20).

ANP-3918, section 3.3, summarizes the thermal and mechanical design criteria. The licensee stated that fuel rod internal hydriding is verified by quality control inspections. ANP-3918P, table 3-3 presents the evaluation results. The NRC staff find the results acceptable because the acceptance criteria stated in ANP-3918P, Table 3-3 are met by comparing with the analysis results.

3.4 Mixed Core Thermal-Mechanical Analysis Using PRIME Methodology The NRC-approved GNF PRIME model and computer program documented in TRs NEDC-33256P-A, NEDC-33257P-A, and NEDC-33258P-A, (Reference 13), is used to perform coupled thermal and mechanical interaction analyses for the BWR fuel rods. These TRs provide the technical basis, qualification, and application methodology respectively for the best-estimate predictions of the thermal and mechanical performance of fuel rods experiencing variable power histories. NEDC-33840P-A (Reference 14) describes the currently used thermal overpower (TOP) and the mechanical overpower (MOP) methodologies to evaluate and demonstrate compliance with the no fuel melt and the cladding strain criteria for BWR fuels.

Section 4.0 of the NRC staff SE (included in Reference 13) of PRIME TRs provides conditions and limitations for the NRC staffs approval, which are also included in NEDC-33840P-A in support of fuel transitions. The PRIME methodology has been incorporated into GESTAR II, which is referenced in the DNPS, Units 2 and 3, TSs and incorporated into their licensing bases.

Therefore, the scope of the present review is to satisfy condition and limitation 1.a for the acceptability of using PRIME to generate TOP and MOP limits for ATRIUM 10XM fuel, which would be inconsistent with condition and limitation 1.a of the PRIME SE.

The licensee submitted the GNF reports 006N8642-NP and 006N8642-P, Revision 1 (Reference 11), as attachments 4 and 6 respectively to letter dated January 30, 2022 (Reference 5). This report along with its supplements (References 6 and 7) provide the following information:

It establishes a technical basis to justify the use of PRIME methodology as an alternate to demonstrate compliance with no fuel melt and cladding strain criterion for non-GNF fuels.

A basis for the properties to use in PRIME to model CWSR zircaloy-2 cladding for material properties ((

)), per the Technical Evaluation of the Safety Evaluation Report (SER) included in NEDC-33257P-A.

Justification supporting the LARs for transition from ATRIUM 10XM fuel to GNF3 fuel for DNPS Units 2 and 3.

3.4.1 Thermal Overpower The GESTAR II criterion for no fuel melt compliance is that during AOOs, loss of fuel rod mechanical integrity shall not occur due to fuel melting (section 1.2.2.B.ix of GESTAR II). The TOP is the screening parameter used by GNF to show compliance to the no fuel centerline

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION melting criterion. The acceptance criteria for the fuel melt is the fuel rod centerline predicted temperature for the transient must be below the melt temperature. To ensure fuel melting is prevented, GNF performs a statistical analysis of a TOP transient ((

)). In accordance with NEDC-33258P-A, (Reference 13), section 2.2.2, the ((

)) accounts for changes in gap conductance and fuel temperature due to relaxation of pellet-cladding interaction stresses by cladding creep during a long transient. This time interval addresses the time allowed for the operator to take action to reduce power in case of an event requiring operator action. The licensee stated that the ((

)). In the ((

))

category, ((

)) uncertainties are a contributor. GNF ((

)). NRC staff finds the GNF approach acceptable because these uncertainties represent a small contributor to the overall uncertainty in GNFs approach.

The NRC staff finds the GNF approach acceptable because of the following:

((

)) analysis assumption is conservative because for any AOO, a reactor scram or operator intervention will terminate the overpower condition sooner than 10 minutes.

Using the fuel centerline melt temperature assumption is conservative because the way that fuel melt leads to cladding failure is that the melt process causes the pellet to expand to an extent that the fuel cladding fails. However, such a failure would not occur due to a limited amount of fuel melting at the centerline.

In consideration of the above conservatisms, and the ((

)) contributes to the overall uncertainty in the TOP evaluation method, the NRC staff finds the GNF approach acceptable for modeling the GNF3 fuel in the reactor cores. Therefore, the NRC staff determined that the PRIME TOP evaluation methods provide reasonable assurance that fuel melt is precluded during conditions of normal operation and AOOs, consistent with GDC 10 in the DNPS, Units 2 and 3, licensing basis.

3.4.2 Mechanical Overpower The GESTAR II criterion for the cladding strain compliance is that during normal operation and AOOs, loss of fuel rod mechanical integrity shall not occur due to pellet-cladding mechanical interaction (PCMI) (section 1.2.2.B.x of GESTAR II). This limit is referred to as the MOP limit which is the screening parameter used by GNF to show compliance to the cladding strain criterion.

In section 3.1 of report 006N8642-P, GNF provides a brief discussion of the phenomena associated with the MOP analysis, a justification to apply the current MOP acceptance criterion and a limit on cladding strain to GNF3 fuel, and finally a description of updates to the PRIME methodology and further justification to apply the methodology to ATRIUM 10XM fuel. In Section 3.2 of report 006N8642-P, consistent with NEDC-33258P-A, Table 2-1, GNF stated the following transient strain limit for GNF fuel designs in terms of peak pellet exposure (PPE):

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((

))

The above limits are applied as transient diametral strain limits, taken as the maximum strain during the AOO minus the strain at the onset of the AOO. The purpose of these limits is to ensure adequate ductility remains in the cladding.

For the transition cycles, compliance must be shown for the cladding strain criteria for both the newer GNF3 and the co-resident ATRIUM 10XM fuel bundles. As discussed above, for ATRIUM 10XM fuel the GNF PRIME methodology is an alternate method to evaluate the cladding strain criteria based on the MOP limit.

As described in section 3.1 of report 006N8642-P, the primary driver for cladding straining during an AOO is thermal expansion of the fuel pellet. When the pellet expands in an overpower condition, the cladding yields upon expansion of the UO2 [uranium dioxide] pellet. The PRIME methodology is used to analyze fuel in an overpower condition in both slow and fast transients to predict the cladding strain resulting from the transient.

In section 3.3 of report 006N8642-P, GNF provided a brief description of the worst-tolerance (WSTOL) approach to calculate the transient ((

)) cladding strain when ((

)). In this approach, all design and operating parameters that impact the calculated cladding strain are placed at their worst tolerance, which is most conservative for the purposes of the strain calculation limit. The specific parameters are listed in NEDC-33258P-A, table 2. GNF stated that for manufacturing tolerance inputs not available for non-GNF fuel, GNF fabrication parameters will be used. The NRC staff finds it acceptable because GNF fabrication tolerances are applied to the non-GNF fuel in the demonstration of WSTOL predictions for cladding diametral strain in the qualification plots shown in Figure 3-1 and Figure 3-2 of report 006N8642-P. In consideration of the design similarities between the fuel types (i.e., both bundle designs are a 10x10 orthogonal fuel matrix),

the NRC staff determined that applying a conservative representation based on the GNF fuel fabrication parameters is acceptable for application to the ATRIUM 10XM fuel.

In Section 3.3 of report 006N8642-P, GNF described that the WSTOL methodology conservatively biases ((

)) to reduce cladding strength and increase creep rates (via increased temperatures) and increased metal thinning due to oxidation. Since the PRIME model for crud is ((

)), the NRC staff find that it is acceptable for application to the GNF3 fuel.

GNF stated that an additional conservatism applied in WSTOL analyses is a power penalty, which accounts for power spiking and enrichment as well as uncertainties of non-GNF fuels.

This power penalty factor applied on the maximum linear heat generation rate (MLHGR) for WSTOL analysis ((

)).

Based on the above review, NRC staff determined that the report 006N4862-P provided sufficient justification and methodological changes, such that GNF could adequately model predicted cladding strain for the ATRIUM 10XM fuel to demonstrate that cladding failures due to PCMI are prevented. Therefore, the NRC staff finds that the licensee has addressed the

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION requirements of the GDC 10 licensing basis acceptably, with respect to fuel cladding failures associated with PCMI.

3.4.3 Demonstration Sensitivity Study For the ATRIUM 10XM fuel the cladding material is cold work stress relieved (CWSR) Zircaloy-2 (page 3-1 of ANP-3918P). ((

)).

In report 006N8642-P, section 4.3, the licensee described a compliance sensitivity study to demonstrate the effect of using ((

)) fuel in PRIME model to the updated CWSR Zircaloy-2 properties of ATRIUM 10XM fuel. For the study the licensee used the method number 1 approach described in section 5.1 of NEDC-33840P-A, which directly evaluates compliance to fuel centerline melting and cladding transient strain criteria using statistical and WSTOL methodology, respectively.

As stated in the March 16, 2022 letter from Constellation Energy Generation (Reference 6), the sensitivity analysis is based on the DNPS, Unit 3, Cycle 28, mock reload mixed core analysis with resident ATRIUM 10XM and fresh GNF3 fuel bundles. The licensee stated that a mock reload analysis is a study of a typical reload licensing process to identify potential problems and develop solutions prior to performing a licensed reload analysis. The transient for which results are shown in report 006N8642-P, figure 4-3, is an inadvertent start-up of the high-pressure coolant injection (HPCI) system determined from the mock reload analysis to be the most limiting AOO transient in terms of showing compliance to the no fuel melt and cladding strain criterion. The other AOOs analyzed by the licensee and shown to be less limiting included feedwater controller failures (FWCF), load rejection no bypass (LRNBP), and turbine trip no bypass (TTNBP).

The NRC staff finds i the sensitivity analysis acceptable because the results of the analysis showed that the most limiting AOO met the no fuel melt and the cladding strain criteria.

3.5 Containment Analysis Due to differences in the fuel decay heat and the stored sensible energy in the reactor internals (for example in fuel assemblies and other components), the fuel transition from the ATRIUM 10XM fuel to GNF3 fuel may impact the small break and the following design basis accident (DBA) loss-of-coolant accident (LOCA) containment analyses of record (AOR):

(i).

Small break and DBA mass and energy (M&E) release analysis, (ii).

Containment Short-Term Response to a DBA (UFSAR, section 6.2.1.3.2.1),

(iii).

Containment Long-Term Response to a DBA (UFSAR, section 6.2.1.3.2.2),

(iv).

Containment Response to a DBA-LOCA for minimum net positive suction head (NPSH)

(UFSAR, section 6.2.1.3.3),

(v).

LOCA transient loads (UFSAR, section 6.2.1.3.5.2).

For a mixed core (ATRIUM 10XM and GNF3) and full core transition from ATRIUM 10XM to GNF3 fuel, the licensee must confirm whether AORs (i) through (v) remain bounding. In a letter dated January 23, 2023 (Reference 2), in response to an NRC staff request for additional information (RAI), the licensee stated the following:

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The five AOR mentioned in the RAI above in items (i) through (v) are being addressed as part of the fuel transition project. Any changes to these AOR will be reviewed, as applicable, in accordance with 10 CFR 50.59.

The NRC staff finds the above response acceptable because the licensee will evaluate the changes in the containment AORs in accordance with 10 CFR 50.59.

3.6 Evaluation of TS Changes The NRC staffs evaluation of the TS proposed changes listed in section 2.1 above is as follows:

Addition of NEDC-33930P in DNPS, Units 2 and 3, TS 5.6.5.b, as a reference to the use of GEXL98 correlation is acceptable for application to the ATRIUM 10XM fuel design and licensing analysis based on the NRC staff technical evaluation given in Section 3.2 above.

Addition of report 006N8642-P into the DNPS, Units 2 and 3, TS 5.6.5.b, as a reference to determine core operating limits is acceptable because it justifies the application of PRIME methodology for TOP and MOP analysis for the ATRIUM 10XM fuel. The analysis will provide reasonable assurance that the fuel cladding failures are precluded under conditions of normal operation and AOOs.

The NRC staff finds the deletion of the eight Westinghouse TRs in TS 5.6.5.b acceptable because the licensee does not intend to use them in determining the core operating limits in the COLR.

3.7 Environmental Qualification Review The purpose of this LAR is to amend Renewed Facility Operating License Nos. DPR-19 and DPR-25 for DNPS, Units 2 and 3, respectively. The proposed change would support the transition from Framatome, Inc. ATRIUM 10XM fuel to Global Nuclear Fuel-Americas, LLC (GNF) GNF3 fuel at DNPS. The NRC staff reviewed the submittal to determine the impact of the proposed change on the EQ of electrical equipment.

In section 3.4.2, Environmental Qualification Impacts, of attachment 1 of the LAR, the licensee stated that the change in core inventory has no impact on existing normal or post-accident temperature, pressure, or humidity. Further, the licensee noted that the GNF3 core inventory has no impact on the normal operating doses for EQ and that a detailed review of EQ equipment, and their supporting calculations, was conducted using the GNF3 total integrated dose (TID), i.e., accident plus the normal radiation dose.

In addition, the licensee stated that all EQ program equipments tested and/or analyzed radiation resistance envelopes the GNF3 fuels calculated accident radiation dose, calculated or measured normal operating radiation dose, and EQ program required margins. Since the change in TID will have no impact on the EQ zone radiation, thermal, pressure, or humidity conditions, the transition to GNF3 fuel will have no impact on non-safety-related equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions by the safety-related equipment. Additionally, the licensee

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION noted that no new components are being added to the existing DNPS EQ equipment list because of the proposed change.

While reviewing the LAR, the NRC staff determined that an audit in accordance with LIC-111, Regulatory Audits, Revision 1, was necessary to verify key assumptions, analyses, and test reports used to support the basis for the LAR. During the audit, the NRC staff audited documents to confirm that: (1) the change in core inventory has no impact on normal or post-accident temperature, pressure, humidity, or chemical effects, and (2) affected components remain qualified as a result of the revised accident dose. Details of the NRC staffs audit can be found in the audit report dated April 5, 2023, (ML23088A224).

In the LAR, the licensee noted that EQ documentation will be evaluated and updated as required, to support its compliance with the existing EQ program and prioritized to ensure timely revision.

3.7.1 Environmental Qualifications Evaluation Summary Based on its review of the information in the LAR, as confirmed by the results of the regulatory audit, the NRC staff finds that the proposed change should have no adverse impact on the DNPS EQ program or its ability to continue to meet the requirements of 10 CFR 50.49 and IEB 79-01B.

4.0 TECHNICAL AND REGULATORY CONCLUSIONS 4.1 Technical Conclusions Based on the above NRC staff review, the following technical conclusions are made:

NEDC-33930P provides an acceptable GEXL correlation (GEXL98) for the ATRIUM 10XM fuel approved by the NRC staff.

Thermal-mechanical evaluation using the RODEX2A code modified to account for TCD for the ATRIUM 10XM fuel provides acceptable results as documented in ANP-3918P.

The justification of using PRIME methodology, documented in report 006N8642-P, for compliance with the no fuel melt (or TOP) and cladding strain (or MOP) criteria is acceptable. The analysis provides reasonable assurance that the fuel cladding failures are precluded under conditions of normal operation and AOOs. Based on the NRC staffs review, licensees request to apply the PRIME methodology for MOP and TOP analysis for the legacy ATRIUM 10XM fuel is acceptable.

The GNF3 and ATRIUM 10XM fuel designs are dimensionally similar, to the extent that both are 10x10 orthogonal fuel pin matrices constrained by the DNPS, Units 2 and 3 fuel lattice design.

The NRC staff finds the fuel transition to a GNF3 and ATRIUM 10XM for DNPS, Units 2 and 3, acceptable.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 4.2 Regulatory Conclusions Based on the above NRC staff review, the following regulatory conclusions are made:

The 10 CFR 50.36(c)(5) requirement is met because the licensee added the documents NEDC-33930P, Revision 0, and report 006N8642-P, Revision 1, in the DNPS, Units 2 and 3, TS.

The 10 CFR Part 50, appendix A, GDC 10 requirement is met because by using the existing NRC-approved methodologies in conjunction with newly added documents to the DNPS, Units 2 and 3, TS, the licensee can confirm with reasonable assurance that the fuel design analysis results will not exceed the acceptable design limits with appropriate margins during any condition of normal operation and AOOs.

Based on 10 CFR 50.59 process, the licensee will perform containment analysis based on the mixed core and new fuel full core.

5.0 STATE CONSULTATION

In accordance with the Commissions regulations on, the Illinois State official was notified of the proposed issuance of the amendment on April 21, 2023. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change the requirements with respect to installation or use of a facilitys components located within the restricted area as defined in 10 CFR Part. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding November 1, 2022 (87 FR 65833). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

7.0 CONCLUSION

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

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8.0 REFERENCES

1.

Letter from Constellation Energy Generation, LLC to NRC, License Amendment Request Regarding Transition to GNF3 Fuel, August 18, 2022, 9Agencywide Documents Access Management System (ADAMS) Accession No. ML22230C927 (non-Proprietary).

2.

Letter from Constellation Energy Generation, LLC to NRC, Response to Request for Additional Information Regarding Transition to GNF3 Fuel License Amendment Request, January 23, 2023, (ML23023A128) (non-Proprietary).

3.

Letter from Exelon Generation Company, LLC to NRC, Request for Licensing Amendment Regarding Transition to GNF3 Fuel, September 14, 2021, (ML21257A419) (Proprietary) and (ML21257A420) (non-Proprietary).

4.

Letter from Constellation Energy Generation, LLC to NRC, Response to Request for Additional Information Related to the License Amendment Request to Transition to GNF3 Fuel, April 11, 2022, (ML22101A147) (Proprietary), and (ML22101A146) (non-Proprietary).

5.

Letter from Exelon Generation Company, LLC to NRC, Request to Expand Applicability of PRIME Methods to Evaluate Fuel Centerline Melt and Cladding Strain Compliance for Framatome Fuel at Quad Cities, January 20, 2022, (ML22020A400) (Proprietary), and (ML22020A399) (non-Proprietary).

6.

Letter from Constellation Energy Generation, LLC to NRC, Supplement to Request to Expand Applicability of PRIME Methods to Evaluate Fuel Centerline Melt and Cladding Strain Compliance for Framatome Fuel at Quad Cities, March 16, 2022, (ML22075A212)

(non-Proprietary).

7.

Letter from Constellation Energy Generation, LLC, to NRC, Response to Request for Additional Information Regarding Quad Cities Request to Expand Applicability of GNF Thermal Mechanical Analysis Methods to Framatome Fuel, August 10, 2022, (ML22222A102) (Proprietary), and (ML22222A101) (non-Proprietary).

8.

General Electric, NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel (GESTAR II, Main), Revision 31, November 2020, (ML20330A199) (non-Proprietary).

9.

Global Nuclear Fuel, NEDC-33930P, Revision 0, GEXL98 Correlation for ATRIUM 10XM Fuel, February 2021, (ML21257A421) (Proprietary), and (ML21257A420) (non-Proprietary).

10. Framatome, ANP-3918P, Revision 0, ATRIUM 10XM Fuel Rod Thermal-Mechanical Evaluation with RODEX2A for Quad Cities and Dresden, April 2021, (ML21257A421)

(Proprietary), and (ML21257A420) (non-Proprietary),

11. Global Nuclear Fuel, report 006N8642-P, Revision 1, Justification of PRIME Methodologies for Evaluating TOP and MOP Compliance for non-GNF Fuels, January 2022, (ML22020A400) (Proprietary), and (ML22020A399) (Non-Proprietary).

12. Letter from NRC to Constellation Energy Generation, LLC, Quad Cities Nuclear Power Station, Units 1 And 2 - Issuance of Amendment Nos. 293 and 289 Transition to GNF3 Fuel (EPID L-2021-LLA-0159), December 15, 2022, (ML22298A003) (Proprietary).

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13. Global Nuclear Fuel, The PRIME Model for Analysis of Fuel Rod Thermal - Mechanical Performance: Part 1 - Technical Bases, NEDC-33256P-A, Revision 2, October 2021, Part 2

- Qualification, NEDC-33257P-A, Revision 2, October 2021, Part 3 - Application Methodology, NEDC-33258P-A, Revision 2, October 2021, (ML21279A282) (Proprietary) and (ML21279A283) (non-Proprietary).

14. Global Nuclear Fuel, NEDC-33840P-A, Revision 1, The PRIME Model for Transient Analysis of Fuel Rod Thermal-Mechanical Performance., August 2017, (ML17230A011)

(Proprietary) and ML17230A012 (non-Proprietary).

15. Global Nuclear Fuel, NEDC-33880P/NEDO-33880, Revision 0, GEXL21 Correlation for GNF3 Fuel, March 2017, (ML17096A519) (Proprietary) and (ML17096A520) (non-Proprietary).

16. Global Nuclear Fuel, NEDC-32851P-A and NEDO-32851-A, Revision 5, GEXL14 Correlation for GE14 Fuel, April 2011, (ML111290535) (Proprietary), and (ML111290532)

(non-Proprietary).

17. AREVA, BAW-10247PA, Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP Inc., February 2008, (ML081340383) (Part 1, Proprietary), (ML081340385) (Part 2, Proprietary), and (ML081340208) (non-Proprietary).

18. Exxon Nuclear, XN-NF-81-58(P)(A), Revision 2, Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company, March 1984, (ML081340725) (Proprietary).

19. Siemens, EMF-85-74(P) Revision 0 Supplement 1(P)(A) and Supplement 2(P)(A),

RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, Siemens Power Corporation, February 1998, (ML20013K314) and (ML20248M079) (non-Proprietary).

20. AREVA, ANF-89-98(P)(A), Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, May 1995, (ML081350281) (Proprietary).

21. Letter from Constellation Energy Generation, LLC to NRC, providing clarification on the implementation schedules for Dresden, Units 2 and 3, dated April 17, 2023, (ML23107A228)

(non-Proprietary).

22. Letter from Constellation Energy Generation, LLC to NRC, confirming applicability of the Quad Cities SFNB RAI-5 to Dresden, Units 2 and 3, dated May 17, 2023, (ML23136B189)

(non-Proprietary).

Principal Contributors: A. Sallman, NRR Matthew McConnell, NRR K. Hartage, NRR J. Kaizer, NRR R. Grover, NRR Date of Issuance: July 6, 2023

Package: ML23144A330 Proprietary: ML23144A310 Nonproprietary: ML23144A314 NRR-058 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/STSB/BC NRR/DSS/SNSB/BC NAME SArora SRohrer VCusumano PSahd DATE 5/18/2023 5/25/2023 5/30/2023 5/30/2023 OFFIC NRR/DEX/ELTB/BC NRR/DSS/SFNB/BC NRR/DEX/EXHB/BC NRR/DRA/ARCB/BC NAME JPaige SKrepel BHayes KHsueh DATE 5/30/2023 5/30/2023 5/30/2023 5/30/2023 OFFICE OGC NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME BVaisey JWhited SArora DATE 6/27/2023 7/6/2023 7/6/2023