ML22298A002

– Issuance of Amendment Nos. 293 and 289 Transition to GNF3 Fuel (EPID L-2021-LLA-0159) (Public)

ML22298A002
Person / Time
Site:	Quad Cities
Issue date:	12/15/2022
From:	Robert Kuntz Plant Licensing Branch III
To:	Rhoades D Constellation Energy Generation
Kuntz R
References
EPID L-2021-LLA-0159
Download: ML22298A002 (1)

Text

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION to this letter contains sensitive unclassified non-safeguards information. When separated from Enclosure 3, this document is DECONTROLLED.

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:

QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 293 AND 289 RE: TRANSITION TO GNF3 FUEL (EPID L-2021-LLA-0159)

Dear Mr. Rhoades:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 293 to Renewed Facility Operating License No. DPR-29 and Amendment No. 289 to Renewed Facility Operating License No. DPR-30 for Quad Cities Nuclear Power Station, Units 1 and 2 (Quad Cities). The amendments consist of changes to the technical specifications (TSs) in response to your application dated September 14, 2021 (Agencywide Documents Access and Management System Accession No. ML21257A419), as supplemented by letters dated November 3, 2021 (ML21307A444), January 11, 2022 (ML22011A319), and April 11, 2022 (ML22101A145).

The amendments support the transition from Framatome (formerly AREVA) ATRIUM 10XM fuel to Global Nuclear Fuel - Americas, LLC (GNF) GNF3 fuel at Quad Cities. Specifically, the amendment revises TS 5.6.5, "Core Operating Limits Report (COLR)," paragraph b, to add a report that supports 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.

December 15, 2022

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

Sincerely,

/RA/

Robert Kuntz, Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-254 and 50-265

Enclosures:

1. Amendment No. 293 to DPR-29

2. Amendment No. 289 to DPR-30

3. Proprietary Safety Evaluation

4. Non-Proprietary Safety Evaluation cc: Listserv CONSTELLATION ENERGY GENERATION, LLC AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-254 QUAD CITIES NUCLEAR POWER STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 293 Renewed License No. DPR-29 1.

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

A.

The application for amendment by Constellation Energy Generation, LLC (the licensee) dated September 14, 2021, as supplemented by letters dated November 3, 2021, January 11, 2022, and April 11, 2022, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's 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-29 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 293, 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 within 60 days of the date of issuance. Implementation of the amendment shall also include revision of the Updated Final Safety Analysis Report as described in the licensees letter dated September 14, 2021.

FOR THE NUCLEAR REGULATORY COMMISSION Nancy L. Salgado, 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: December 15, 2022

/RA/

CONSTELLATION ENERGY GENERATION, LLC AND MIDAMERICAN ENERGY COMPANY DOCKET NO. 50-265 QUAD CITIES NUCLEAR POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 289 Renewed License No. DPR-30 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 September 14, 2021, as supplemented by letters dated November 3, 2021, January 11, 2022, and April 11, 2022, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's 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-30 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 289, 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 within 60 days of the date of issuance. Implementation of the amendment shall also include revision of the Updated Final Safety Analysis Report as described in the licensees letter dated September 14, 2021.

FOR THE NUCLEAR REGULATORY COMMISSION Nancy L. Salgado, 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: December 15, 2022

/RA/

ATTACHMENT TO LICENSE AMENDMENT NOS. 293 AND 289 RENEWED FACILITY OPERATING LICENSE NOS. DPR-29 AND DPR-30 QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265 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-29 License DPR-29 Page 4 Page 4 License DPR-30 License DPR-30 Page 4 Page 4 TSs TSs 5.6-3 5.6-3 5.6-4 5.6-4 5.6-5 5.6-5 5.6-6 Renewed License No. DPR-29 Amendment No. 293 B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 293, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

C.

The licensee shall maintain the commitments made in response to the March 14, 1983, NUREG-0737 Order, subject to the following provision:

The licensee may make changes to commitments made in response to the March 14, 1983, NUREG-0737 Order without prior approval of the Commission as long as the change would be permitted without NRC approval, pursuant to the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in an Unreviewed Safety Question, a license amendment shall be submitted to the NRC staff for review and approval prior to implementation of the change.

D.

Equalizer Valve Restriction Three of the four valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor operation with one bypass valve open to allow for thermal expansion of water.

E.

The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined sets of plans1, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: Quad Cities Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2, submitted by letter dated May 17, 2006.

Constellation Energy Generation, LLC shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The CSP was approved by License Amendment No. 249 as modified by License Amendment No. 259.

1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Renewed License No. DPR-30 Amendment No. 289 B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 289, are hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

C.

The licensee shall maintain the commitments made in response to the March 14, 1983, NUREG-0737 Order, subject to the following provision:

The licensee may make changes to commitments made in response to the March 14, 1983, NUREG-0737 Order without prior approval of the Commission as long as the change would be permitted without NRC approval, pursuant to the requirements of 10 CFR 50.59. Consistent with this regulation, if the change results in an Unreviewed Safety Question, a license amendment shall be submitted to the NRC staff for review and approval prior to implementation of the change.

D.

Equalizer Valve Restriction Three of the four valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor operation with one bypass valve open to allow for thermal expansion of water.

E.

The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans1, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: Quad Cities Nuclear Power Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 2, submitted by letter dated May 17, 2006.

Constellation Energy Generation, LLC shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The CSP was approved by License Amendment No. 244 and modified by License Amendment No. 254.

1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Reporting Requirements 5.6 5.6 Reporting Requirements Quad Cities 1 and 2 5.6-3 Amendment No. 293/289 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 5.6 Reporting Requirements Quad Cities 1 and 2 5.6-4 Amendment No. 26293/2894/259 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 5.6 Reporting Requirements Quad Cities 1 and 2 5.6-5 Amendment No. 26293/289259 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, February 2021, as approved by NRC Staff SE dated December 15, 2022. December 10, 2022.

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).

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

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. 293 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-29 AND AMENDMENT NO. 289 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-30 CONSTELLATION ENERGY GENERATION, LLC AND MIDAMERICAN ENERGY COMPANY QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION PROPRIETARY SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 293 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-29 AND AMENDMENT NO. 289 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-30 CONSTELLATION ENERGY GENERATION, LLC AND MIDAMERICAN ENERGY COMPANY QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265

1.0 INTRODUCTION

By letter to the U.S. Nuclear Regulatory Commission (NRC, the Commission) dated September 14, 2021 (Agencywide Documents Access and Management System Accession No. ML21257A419), as supplemented by letters dated November 3, 2021 (ML21307A444),

January 11, 2022 (ML22011A319), and April 11, 2022 (ML22101A145), Constellation Energy Generation, LLC (the licensee) requested changes to the technical specifications (TSs),

renewed facility operating licenses, surveillance requirements (SRs) for the Quad Cities Nuclear Power Station, Units 1 and 2 (Quad Cities). The proposed changes support the transition from Framatome (formerly AREVA) ATRIUM 10XM fuel to Global Nuclear Fuel - Americas, LLC (GNF) GNF3 fuel at Quad Cities. Specifically, the amendments would revise TS 5.6.5, "Core Operating Limits Report (COLR)," paragraph b, to add a report that supports 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 request approval of the use of Framatome RODEX2A methodology with an additional thermal conductivity degradation penalty in mixed core thermal-mechanical calculations. Additionally, the amendments would revise the alternative source term (AST) loss-of-coolant accident (LOCA) analysis to use a bounding core inventory.

The supplemental letters date November 3, 2022, January 11, 2022, and April 11, 2022, 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 on January 4, 2022 (87 FR 256).

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION

2.0 REGULATORY EVALUATION

2.1 Description of Proposed Changes 2.1.1 GEXL98 Correlation As part of the license amendment request (LAR) to transition to GNF3 fuel, Constellation requested the NRC review and approve the topical report NEDC-33930P, Revision 0, "GEXL98 Correlation for ATRIUM 10XM Fuel" (NEDC-33930) (ML21257A420). The purpose of this report was to describe the GEXL98 critical power (CP) model that will be applied to ATRIUM 10XM fuel.

2.1.2 Fuel Thermal-Mechanical Analysis Methods To support unrestricted use of Framatome fuel, the licensee presently uses the RODEX4 fuel thermal-mechanical analysis methodology documented in BAW-10247(P/NP)(A), Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors (ML081340220). However, to support the transition to GNF fuel the licensee opted to use RODEX2A to support the transition core analysis because ((

)).

The analytic methods comprising RODEX2A are described in XN-NF-81-58(P)(A) Revision 2 (ML081340725, non-public) and Supplements 1 and 2, RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, and in EMF-85-74(P) Revision 0 Supplement 1(P)(A) and Supplement 2(P)(A), RODEX2A (BWR [boiling-water reactor]) Fuel Rod Thermal-Mechanical Evaluation Model, (ML20013K315, non-public) all of which have been reviewed and approved for use by the NRC staff. In addition, the licensee will use the acceptance criteria contained in ANF-89-98(P)(A), Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs (ML081350281, non-public). All of these documents are already referenced in Quad Cities TS 5.6.5, Core Operating Limits Report (COLR), such that no TS changes are necessary to implement the RODEX2A fuel thermal-mechanical analysis.

The focus of the present review, therefore, is on the technical adequacy of the transition fuel thermal-mechanical analysis. The analyses documented in ANP-3918P, ATRIUM 10XM Fuel Rod Thermal-Mechanical Evaluation with RODEX2A for Quad Cities and Dresden, (ANP-3918) provide a set of limits on the linear heat generation rate (LHGR) of the Framatome ATRIUM 10XM fuel. When the ATRIUM 10XM fuel is operated in accordance with these limits, the report demonstrates that the applicable design criteria are satisfied.

2.1.3 Core Inventory Update and Resulting Dose Consequences The current analysis applicable to Framatome reload cores with the ATRIUM 10XM fuel is described in section 15.4.10 of the Quad Cities Updated Final Safety Analysis Report (UFSAR)

(ML22049A114). According to section 15.4.10.5.4 of the UFSAR, the Framatome analysis demonstrates that the total number of failed rods is fewer than 1000, which is the basis for the currently postulated radiological consequences. The LAR stated, For the post-CRDA [control rod drop accident] dose consequence analysis, consistent with section S.2.2.3.1.4 of Reference 6.2, the number of fuel rods that would reach [a deposited fuel enthalpy of] 170 cal/gm [calories per gram] is used. The in-text reference is to a section of the United States supplement to

OFFICIAL USE ONLY PROPRIETARY INFORMATION NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel (GESTAR-II, ML20330A195) The 170 cal/gm enthalpy deposition is used within both the Quad Cities licensing basis and that for the GNF3 fuel product line as an acceptance criterion, above, which the fuel cladding for that rod is assumed to fail and release radionuclides into the coolant following the postulated accident. For GNF3, a different number of fuel rods are assumed to fail, which necessitates an update to the radiological consequences analysis.

The NRC staff reviewed the CRDA analysis applicable to GNF3 fuel that supports the estimated number of failed fuel rods to ensure that the analytic methods were conservative, meaning the number is not underestimated, and the existing licensing basis criterion for fuel damage, 170 cal/gm, was adequately justified for the new fuel design. These considerations assure that the estimated radiological consequences have an adequate analytic basis.

2.1.4 Environmental Qualification (EQ) Impacts Section 3.4, Environmental Qualification Impacts, of attachment 1 of the LAR, stated that the change in core inventory has no impact on normal or post-accident temperature, pressure, or humidity. Further, the LAR stated that the GNF3 core inventory does not affect the normal operating doses for EQ and a detailed review of EQ equipment was conducted using the GNF3 scaled total integrated dose, i.e., accident plus the normal radiation dose.

2.1.5 TSs Change Per the supplemental letter, dated April 11, 2022, the proposed change would update the COLR listing of documents in TS 5.6.5.b with the addition of a new Topical Report, NEDC-33930P, "GEXL98 Correlation for ATRIUM 10XM Fuel," and the renumbered listing would refer the Report in COLR listing as # 21 as follows:

NEDC-33930P, Revision 0, "GEXL98 Correlation for ATRIUM 10 XM Fuel,"

dated February 2021, as approved by NRC Staff SE [safety evaluation]

dated XXX XX, 20XX.

2.2 Applicable Regulatory Requirements and Guidance 2.2.1 Regulatory Requirements The following NRC regulatory requirements are applicable to the NRC staff's review for this proposed license amendment:

Per Title 10 to the Code of Federal Regulations (10 CFR) 50.36(c)(5) 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 a safe manner.

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, aging, be included in the qualification program.

Regulation 10 CFR 50.67, Accident Source Term, establishes radiation dose limits for individuals at the boundary of the exclusion area and at the outer boundary of the low OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION population zone.10 CFR 50.67(b)(2) states that the NRC may issue the amendment only if the applicant's analysis demonstrates with reasonable assurance that:

(iv)

An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(v)

An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(vi)

Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

General Design Criteria The construction permits for Quad Cities predate the formal issuance of the current 10 CFR part 50, appendix A, General Design Criteria (GDC). During the construction permit licensing process, Quad Cities was evaluated against the Comment Draft of 70 Criteria, which was issued on July 10, 1967. The design bases of each Quad Cities unit were reevaluated at the time of initial Final Safety Analysis Report preparation against the draft of the 70 criteria current at the time of operating license application.

As stated in section 3.1 of the UFSAR, based on the understanding of the intent of the proposed criteria current at the time of operating license application, Quad Cities conforms with the intent of the Atomic Energy Commission GDC for nuclear Power Plant Construction Permits. As the GDC were finalized, the requirements were placed in Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR part 50, Domestic Licensing of Production and Utilization Facilities.

GDC, Criterion 10, Thermal Hydraulic System Design, contains the regulatory requirements applicable to fuel thermal-mechanical limits. Criterion 6 of the Quad Cities design basis is analogous to GDC, Criterion 10, and states:

The reactor core shall be designed to function throughout its design lifetime, without exceeding acceptable fuel damage limits which have been stipulated and justified. The core design, together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of a turbine generator set, isolation of the reactor from its primary heat sink, and loss of all offsite power.

The regulatory requirements applicable to the CRDA are found in GDC 13 Instrumentation and Control, and GDC 28 Reactivity Limits. The analogous criteria to GDC 13, in UFSAR section 3.1.2 are Criterion 12, Instrumentation and Control Systems, Criterion 13, Fission Process

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Monitors and Controls, and Criterion 14, Core Protection Systems. These criteria require the availability of instrumentation to monitor variables and systems over their anticipated ranges to assure safety, and of appropriate controls to maintain these variables and systems within prescribed operating ranges. These criteria apply because the sequence of events associated with the CRDA includes automatic actuations of protection systems, and potentially manual actions, and the sequence of events must be justified based on the expected values of the relevant monitored parameters and instrument indications.

The analogous criterion to GDC 28, in UFSAR section 3.1.2 is Criterion 32, Maximum Reactivity Worth of Control Rods. This criterion requires that the potential effects of a sudden or large change of reactivity, such as a dropped control rod, cannot (a) rupture the reactor coolant boundary or (b) disrupt the core, its support structures, or other vessel internals sufficiently to impair the effectiveness of emergency core cooling. This criterion applies because (1) the transient fuel enthalpy must be assessed for whether cladding failure or fuel rupture is predicted and to what extent; (2) the coolability of the core following the event must be established; and (3) the maximum reactor coolant pressure must be predicted to demonstrate stress limits for the reactor pressure vessel are not exceeded. These requirements provide assurance that fuel damage and reactor vessel pressure will not be excessive in the CRDA.

2.2.2 Guidance The NRC staff review of ANP-3918P was based on guidance contained in chapter 4.2, Fuel System Design, of NUREG-0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [light-water reactor] Edition (ML070740002).

The fuel system safety review, including the fuel thermal-mechanical analysis, provides reasonable assurance that the fuel system is not damaged as a result of normal operation and AOOs. According to chapter 1 of ANP-3918, the RODEX2A code is used to 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). 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 in the fuel pellets. Information Notice (IN) 2009-23, Nuclear Fuel Thermal Conductivity Degradation, 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 (ML091550527).

In order to ensure that a Critical Boiling Transition (CBT) 5 does not occur, two SAFDLs (specified acceptable fuel design limit) have been developed, as described in SRP, section 4.4, Thermal and Hydraulic Design (ML070550060):

5 CBT is the name given to the phenomena which occurs when a flow regime that has a higher heat transfer rate transitions to a flow regime that has a significantly lower heat transfer rate. Historically, terms such as critical heat flux, departure from nucleate boiling, and critical power have been used. However, the NRC staff needed a way to separate the general phenomena occurring (i.e., CBT) from a specific type of phenomena which may occur (e.g., departure from nucleate boiling, dryout) and from the specific values of certain parameters which are often used to signify such a transition has occurred (e.g., critical heat flux, critical power).

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION (c) 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 [departure from nucleate boiling] or boiling crisis condition during normal operation or AOOs, or (d) at least 99.9 percent of the fuel rods in the core will not experience a DNB or boiling crisis during normal operation or AOOs.

Typically, SAFDL (a) is associated with PWRs (pressurized-water reactors) and SAFDL (b) is associated with BWRs. CBT models such as GEXL98 which will be used on BWR fuel are necessary to ensure that the above SAFDLs can be satisfied. The main objective of this review was to determine GEXL98 could result in accurate predictions, such that at least 99.9 percent of the fuels rods in the core will not experience CBT during normal operation or AOOs.

Quad Cities UFSAR, Section 3.11, Environmental Qualification of Electrical Equipment, states:

The environmental qualification (EQ) of electrical equipment is performed in accordance with the guidelines of NRC Office of Inspection and Enforcement Bulletin (IEB)79-01B, [Environmental Qualification of Class IE Equipment, available on the NRC public website at https://www.nrc.gov/reading-rm/doc-collections/gen-comm/gen-letters/1980/gl80005.html] and the requirements of 10 CFR 50.49.

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.

Section 3.11 of NUREG-800, provides guidance on EQ of mechanical and electrical equipment for complying with 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants.

NUREG-0800, section 15.0.1, Radiological Consequence Analyses Using Alternative Source Terms, Revision 0, dated July 2000 (ML003734190), provides guidance on determining an AST to meet the requirements in 10 CFR 50.67.

Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision 0, dated July 2000 (ML003716792),

provides guidance to licensees of operating power reactors on acceptable applications of alternative source terms; the scope, nature, and documentation of associated analyses and evaluations; consideration of impacts on analyzed risk; and content of submittals.

3.0 TECHNICAL EVALUATION

3.1 GEXL98 Correlation NEDC-33930P describes how GEXL98 was developed using the GEXL correlation form and trained using data from the NRC approved ATRIUM 10XM CPR correlation (ML14183A734).

The development of GEXL98 follows the process that has been used in previously approved version of GEXL which has been applied to non-GE fuel, including GEXL96 which was applied

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION to ATRIUM-9B (ML003755953), GEXL80 that was applied to SVEA96+ (ML043210058), and GEXL97 that was applied to ATRIUM-10 (ML082070088).

The NRC staffs technical evaluation is focused on determining if GEXL98 is acceptable for use in reactor safety license calculations (i.e., that the model can be trusted). To perform this review, the NRC staff used the review framework described in NUREG/KM-0013, Credibility Assessment Framework for Critical Boiling Transition Models (ML19073A249).

3.1.1 Review of Framework for Critical Boiling Transition Models This section discusses the review framework for CBT models. The framework is generated from a single main goal. That main goal is then logically decomposed into subgoals. Logical decomposition is the process of generating a set of subgoals which are logically equivalent (i.e.,

necessary and sufficient) to the main goal. This decomposition is expressed using Goal Structure Notation (GSN). Each subgoal can either be further logically decomposed into other subgoals or if no further decomposition is deemed useful, the subgoal is considered a base goal and evidence must be provided to demonstrate that the base goal is true.

For CBT models, the top goal is that the CBT model can be trusted in reactor safety analyses.

Based on the experience from multiple NRC technical staff members, a study of previous SEs, and multiple discussions with various industry experts, this goal is broken down into various subgoals as given in the figures below.

Figure 9: Breaking Down of G - Main Goal Experimental data Experimental data is the cornerstone of a CBT model. Not only is the data used to generate the coefficients of the model and validate the model, but previous data are often used to generate the models form. Therefore, it is essential that the experimental data are appropriate.

Demonstrating the experimental data is appropriate is accomplished using the three sub-goals given in Figure 2 below.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Figure 10: Breaking Down G1 - Experimental Data While the experimental data are generally the focus of a CBT model, for models such as GEXL98, no experimental data is used. Instead of trying to replicate the experimental behavior, the goal of GEXL98 is to replicate the behavior of the Framatomes approved ATRIUM 10XM CPR correlation, such that GEXL98 could be used to predict the CPR performance of ATRIUM 10XM fuel. Thus, instead of fitting the correlation to experimental data, the correlation is fitted to predictions from the ATRIUM 10XM correlation.

Because the NRC has previously approved the ATRIUM 10XM correlation, the NRC believes there is reasonable assurance that the results of the approved correlation can be trusted in supplying data which is used to train and validate GEXL98. Therefore, the NRC staff considers goal G1 met, as the data are from an approved CP correlation and the uncertainty of that data is the uncertainty of the correlation which has been previously quantified in the prior staff review.

Model Generation There are numerous ways to generate the CBT model. While some methods of model generation are based on first principle physics, CBT models are still largely empirical in nature.

Therefore, there is no single correct way to generate the model. The form of the model is often based on previous model forms, machine learning techniques, and engineering judgment.

Demonstrating the model generation is appropriate is accomplished using the two sub-goals given in Figure 3 below.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Figure 11: Breaking Down G2 - Model Generation The development of GEXL98 is consistent with the general GEXL form and with the previous application of that form to other non-GE fuel types. Because the NRC staff have previously reviewed and approved those prior developments, and because the NRC staff performing this review do not believe there is anything in the development of GEXL98 which would change the staffs conclusions or require additional review, the NRC staff have concluded that goal G2 has been previously reviewed and approved, and that no additional review is warranted.

Model Validation As defined by Oberkampf and Roy6, validation is the accumulation of evidence used to assess the claim that a model can predict a real physical quantity. Thus, validation is a never-ending process as more evidence can always be obtained to bolster this claim. However, at some point, when the accumulation of evidence is considered sufficient to make the judgment that the model can be trusted for its given purpose, the model is said to be validated. Demonstrating the model validation is appropriate is accomplished using the five sub-goals given in Figure 4 below.

6 Oberkampf, W.L., and C.J. Roy, Verification and Validation in Scientific Computing, Cambridge University Press, Cambridge, United Kingdom, 2010.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Figure 12: Breaking Down G3 - Model Validation Validation Error Validation Error The correct validation error has been calculated.

G3.1, Review Framework for CBT Models The validation error calculated is the ratio of the GEXL98 prediction of critical power to that of the ATRIUM 10XM prediction of critical power. This CPR is similar to the experimental critical power ratio (ECPR) where the correlations prediction of critical power is compared to the measured value of critical power. However, for GEXL98, there are no true measurements of critical power and the prediction from the approved ATRIUM 10XM correlation is used in its place.

Because the NRC staff has previously reviewed and approved the ATRIUM 10XM correlation, the goal of this topical report is to ensure that GEXL98 correlation matches the predictions from that approved correlation, and the ECPR as defined would allow for the quantification of the difference between the predictions of GEXL98 and the ATRIUM 10XM correlation, the NRC staff has concluded that this goal has been met.

Data Distribution The second sub-goal in demonstrating that the models validation was appropriate is to demonstrate that the data is appropriately distributed throughout the application domain. This is typically demonstrated using the six sub-goals as given in Figure 5 below.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Figure 13: Breaking Down G3.2 - Data Distribution No further break down of the sub-goals is deemed useful. Therefore, the evidence demonstrating the following goals were met are provided below.

Validation Data Validation Data The validation data (i.e., the data used to quantify the models error) should be identified.

G3.2.1, Review Framework for CBT Models The letter dated January 11, 2022, provide a response to the NRC staffs request for additional information (RAI) RAI-01. The response to RAI-01 provided the data set used to train and validate the GEXL98 CPR model. However, the response did not identify which of the data was used for training and which was used for validation. Therefore, the letter dated April 11, 2022, provided a response to SFNB-RAI-2 specified which data was used for training and which data was used for validation.

Additionally, the response to SFNB-RAI-3, provided a table summarizing a random subspace analysis of the ECPR statistics from training a validation. That table described the results from refitting the correlation by randomly choosing 75 percent of the data set to use for training and then using the remaining 25 percent of the data set for validation purposes. This process was repeated three times resulting in four different versions of the GEXL98 correlation. The mean and variance of the results ECPR values were very consistent demonstrating that this correlation is relatively insensitive to the data used for training. Additionally, as demonstrated in table 3-2 from the topical report, there was not significant difference in the mean or variance from the validation data set compared to the training or total data set. Because the random subspace analysis demonstrates that there is little sensitivity to the training data chosen, because the statistics for the training, validation, and total data sets are nearly identical, and

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION because each data point was identified as either belonging to the training or validation set, the NRC has concluded that this goal has been met.

Application Domain Application Domain The application domain of the model should be mathematically defined.

G3.2.2, Review Framework for CBT Models The application domain for ACE/ATRIUM-10XM is given in table 2-1 of Reference 1 in NEDC-33930P, and the application domain for GEXL98 is given in table 4-3 in NEDC-33930P.

((

)) Because the application domain has been mathematical defined, the NRC has concluded that this goal has been met.

Expected Domain Expected Domain The expected domain of the model should be understood.

G3.2.3, Review Framework for CBT Models The expected domain of the GEXL98 correlation is given in Figures 2-2, 2-3, and 2-4 of NEDC-33930. Because these figures allow for a detailed understanding of the expected domain, the NRC staff concludes that this goal has been met.

Data Density Data Density There should be an appropriate data density throughout the expected domain.

G3.2.4, Review Framework for CBT Models

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The expected domain of the GEXL98 correlation is given in Figures 2-2, 2-3, and 2-4, of NEDC-33930. The NRC staff finds that the data density given in these figures is consistent with the data density of previously approved models. Because the data density is consistent with previous approval, the NRC staff concludes that this goal has been met.

Sparse Regions Sparse Regions Sparse regions (i.e., regions of low data density) in the expected domain should be identified and justified to be appropriate.

G3.2.5, Review Framework for CBT Models The expected domain of the GEXL98 correlation is given in Figures 2-2, 2-3, and 2-4, of NEDC-33930. As demonstrated in the figures, there are no sparse regions in the expected domain. Because there are no sparse regions in the expected domain, the NRC staff concludes that this goal has been met.

Restricted Domain Restricted Domain The model should be restricted to its application domain.

G3.2.6, Review Framework for CBT Models Consistent with the approval of many other CP models (as well as critical heat flux models),

approval of the proposed amendment would place a limitation on the use of the GEXL98 correlation which restricts its approval to its application domain. Because this limitation is in place, the NRC staff concludes that this goal has been met.

Consistent Model Error The third sub-goal in demonstrating that the models validation was appropriate is to demonstrate that the model error is consistent over the application domain. This is typically demonstrated using the three sub-goals as given in Figure 6 below.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Figure 14: Breaking Down G3.3 - Consistent Model Error No further decompositions of the sub-goals were deemed useful. Therefore, the evidence demonstrating the following goals were met are provided below.

Poolability Poolability The validation error should be investigated to determine if it contains any sub-groups which are obviously not from the same population (i.e., not poolable).

G3.3.1, Review Framework for CBT Models Table 3-3 of NEDC-33930 and the response to SFNB-RAI-3 provided analysis which demonstrated that there were no obvious sub-groups in the GEXL98 data set which could not be pooled into the same population. The NRC staff also confirmed this by performing their own analysis on the data and did not discover any obvious non-poolable sub-groups. Because there were no sub-groups discovered which were non-poolable, the NRC staff concludes that this goal has been met.

Non-Conservative Subregions Non-Conservative Subregions The expected domain should be investigated to determine if contains any non-conservative subregions which would impact the predictive capability of the model.

G3.3.2, Review Framework for CBT Models

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The NRC staff identified ((

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION

)) Because Constellation has taken appropriate steps to account for any non-conservative subregions, the NRC staff have concluded that this goal has been met.

Model Trends Model Trends The model is trending as expected in each of the various model parameters.

G3.3.3, Review Framework for CBT Models The ECPR trends versus model parameters are given in Figures 3-2 through 3-5 of NEDC-33930. The NRC staff found the trending was reasonable when compared with other CP models. The ECPR trends provided were for the entire data set and not just the validation data set, therefore the NRC staff created their own plots of solely the validation data using the data provided in response to SFNB-RAI-2. Based on those plots, the NRC staff confirmed that there was not significant difference in trending between only the validation data and the entire data set. Because there were no unexpected trends in the ECPR versus model parameters, the NRC staff has concluded that this goal has been met.

Quantified Model Error The fourth sub-goal in demonstrating that the models validation was appropriate is to demonstrate that the model error has been appropriately quantified over the application domain.

This is typically demonstrated using the three sub-goals as given in Figure 7 below.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Figure 15: Breaking Down G3.4 - Quantified Model Error No further decompositions of the sub-goals were deemed useful. Therefore, the evidence demonstrating the following goals were met are provided below.

Error Database Error Data Base The validation error statistics should be calculated from an appropriate database.

G3.4.1, Review Framework for CBT Models In table 3.2 of NEDC-33930, the statistics for the error are given. This includes the ECPR mean and variance for the training data, the validation data, and the combined data set. Because there is no substantial difference between these three data sets, and because the uncertainty used for GEXL bounds the validation data set, the NRC staff has concluded that this goal has been met.

Statistical Method Statistical Method The validation error statistics should be calculated using an appropriate method.

G3.4.2, Review Framework for CBT Models In section 6.0 of NEDC-33930, the statical method used to generate the error is discussed. The process used to ensure that the acceptance criterion of less than 0.1% of fuel rods are suspectable to critical power assumes that the ECPR distribution has a normal shape. While the actual distribution of GEXL98s ECPR values is not a normal distribution, the conservative increase in the standard deviation ensures that the underlying population will be conservatively evaluated. Because the GEXL98 mean and standard deviation conservatively bound the validation data, the NRC staff has concluded that this goal has been met.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Appropriate Bias for Model Uncertainty Appropriate Bias The models error should be appropriately biased in generating the model uncertainty.

G3.4.3, Review Framework for CBT Models Section 6.0 of NEDC-33930 describes the method for determining the overall GEXL98 uncertainty. This uncertainty is generated by ((

)) Because the GEXL98 uncertainty includes the uncertainty in the ACE/ATRIUM-10XM correlations prediction of measured data, the uncertainty in the GEXL98 correlations prediction of the ACE/ACTIRUM-10XMs predicted values, and a conservative combination of both uncertainties, the NRC staff has concluded that this goal has been met.

Model Implementation The fifth sub-goal in demonstrating that the models validation was appropriate is to demonstrate that the model will be implemented in a manner consistent with its validation. This is typically demonstrated using the two sub-goals as given in Figure 8 below.

Figure 16: Breaking Down G3.5-Model Implementation

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION No further decompositions of the sub-goals were deemed useful. Therefore, the evidence demonstrating the following goals were met are provided below.

Same Computer Code Same Computer Code The model has been implemented in the same computer code which was used to generate the validation data.

G3.5.1, Review Framework for CBT Models As specified in response to SFNB-RAI-6, GEXL98 was implemented in GEXLM02, a computer module which is shared among multiple computer codes. Because the validation of GEXL98 was performed using this module and this module will be called in the application GEXL98, the NRC staff have concluded that this goal has been met.

Same Methodology Same Methodology The models prediction of the critical boiling transition is being applied using the same methodology as it was when predicting the validation data set for determining the validation error.

G3.5.2, Review Framework for CBT Models Unlike DNB models which are calculated in a subchannel code which has a number of complex closure relations, the application of a CPR model is relatively simple. A single channel is modeled with R-factors accounting for radial powers and additive constants accounting for thermal hydraulic effects. Given the simplicity of CPR models and the description of the procedures for implementation and validating such a model in a new computer code provided in response SFNB-RAI-6, the NRC staff has determined that this goal has been met.

Transient Prediction Transient Prediction The model results in an accurate or conservative prediction when it is used to predict transient behavior.

G3.5.3, Review Framework for CBT Models Generally, the application of a CP model to a transient will be justified by comparison to transient data. However, the goal of GEXL98 is to replicate the behavior of the ACE/ATRIUM-10XM correlation, therefore, the NRC staff has determined that this criterion does not apply.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 3.1.2 GEXL98 Correlation Conclusion Based on evidence provided in section 3.1.1 of this SE, the NRC staff concludes that the data supporting the GEXL98 correlation was previously reviewed and approved via the approval of ACE/ATRIUM 10XM. In section 3.1.1 of this SE, the NRC staff concludes that the generation of the GEXL98 correlation was previously reviewed and approved in past topical reports. Based on the evidence in section 3.1.1 of this SE, the NRC staff concludes that the GEXL98 has sufficient validation as demonstrated through appropriate quantification of its error. Therefore, based on the cumulative evidence, the NRC staff concludes that GEXL98 is appropriate to be used in reactor safety analysis for Quad Cities to analyze the fuel transition from ATRIUM 10XM fuel to GNF3.

3.2 Fuel Thermal-Mechanical Analysis Methods 3.2.1 RODEX2A-Based Fuel Thermal-Mechanical Limits The analyses documented in ANP-3918 provide a set of limits on the linear heat generation rate (LHGR) of the Framatome ATRIUM 10XM fuel. When the ATRIUM 10XM fuel is operated in accordance with these limits, the report demonstrates that the applicable design criteria are satisfied.

LHGR Limits Chapter 2 of ANP-3918 provides a set of LHGR limits. These curves, located in Figures 2-1 and 2-2, represent the maximum linear heat generation rate that was analyzed in RODEX2A to demonstrate that the applicable design limits were satisfied. These curves will form the basis for future core operating limits.

Evaluation of Fuel Design Limits Chapter 3 of ANP-3918 presents the fuel rod design evaluation. The ATRIUM 10XM fuel rod performance is evaluated for operation up to the linear heat generation rate limits provided in Chapter 2 of ANP-3918. The fuel rod evaluation is limited to a rod average exposure of 62 megawatt days per kilogram uranium (MWd/kgU).

Within this envelope, (i.e., peak allowable linear heat generation rate and peak fuel rod average burnup up to 62 MWd/kgU), the fuel rod performance is evaluated against the following thermal and mechanical design criteria: internal hydriding; cladding collapse; overheating of fuel pellets; stress and strain limits; fuel densification and swelling; fatigue; cladding oxidation, hydriding and crud buildup; rod internal pressure; and plenum spring design. The specific acceptance criteria are developed in ANF-89-98(P)(A), revision 1 and supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs. As noted in section 2.1.2, above, this topical report (TR) has been approved for use by the NRC staff and has already been incorporated into the Quad Cities TS COLR references. Since the specific values of the acceptance criteria are contained in an NRC-approved TR that is already referenced in the Quad Cities TS, the NRC staff determined that they are an acceptable way to demonstrate that the ATRIUM 10XM fuel will continue to remain intact under conditions of normal operation, with appropriate margin for AOOs, consistent with Criterion 6 of the Quad Cities design basis.

Section 3.3 of ANP-3918 presents a summary of the evaluation performed to establish that each design criterion is satisfied. The results are provided in table 3-3 of ANP-3918, which

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION demonstrates that the acceptance criterion for each fuel damage mechanism is satisfied. Since the licensee used NRC-approved methods and design acceptance criteria to demonstrate that the fuel design limits satisfy the applicable acceptance criteria, the NRC staff determined that the proposed fuel design limits contained in ANP-3918P are acceptable and may be applied to the legacy, Framatome fuel, in concert with the transition to the GNF3 fuel bundle design.

Thermal Conductivity Degradation The degradation in thermal conductivity of nuclear fuel as a function of burnup is described in NRC Information Notice (IN) 2009-23, Nuclear Fuel Thermal Conductivity Degradation (ML091550527 and supplement 1 at ML121730336). As nuclear fuel is burned through its lifetime, atomic and subatomic interactions within the fuel cause a reduction in the thermal conductivity of the fuel, causing the fuel pellets to retain additional heat at a given power level.

Some thermal mechanical codes that are used to predict fuel performance, which are largely based on empirical models, cannot accurately represent this phenomenon because they were developed before data existed to quantify the nature of this phenomenon.

The RODEX2A fuel performance code is among those that were developed before the effects of thermal conductivity degradation (TCD) were understood and quantified; as such, it does not model this phenomenon. When uncorrected, RODEX2A will underpredict fuel temperatures at increasing burnup levels. This condition can cause the code to underestimate the effects of fuel system damage from mechanisms like fuel centerline melt and cladding plastic strain, which reduces assurance that the analyses demonstrate compliance with Criterion 6 of the Quad Cities design basis, concerning the stipulation and justification of acceptable fuel damage limits.

The analyses described in ANP-3918P applied penalties to address the impact of TCD relative to fuel centerline melting and transient cladding strain and determined that the modeling for fuel rod internal pressure was adequately conservative. These penalties are based on analyses described in a July 14, 2009, assessment of the effects of the lack of TCD models in RODEX2A (ML092010160). The assessment used ((

)), to quantify the effects of TCD relative to the predicted results obtained from RODEX2A. The assessment used the same approach to demonstrate that the RODEX2A modeling for internal rod gas pressure was adequately conservative to account for the effects of TCD, despite not modeling the phenomenon explicitly. The 2009 assessment establishes a quantitative means to assure that the RODEX2A predictions for centerline melting, transient cladding strain, and fuel rod internal pressure include sufficient margin relative to the acceptance criteria to account for the effects of TCD.

The NRC staff reviewed both ANP-3918P and the 2009 assessment. Since the 2009 assessment was based on both empirical data and a code-to-code comparison, the NRC staff determined that the assessment provides an acceptable means to quantify the effects of TCD relative to RODEX2A predictions of fuel centerline melting, transient cladding strain, and fuel rod internal pressure. The NRC staff also notes that the assessment was performed using the ATRIUM 10XM fuel bundle design, meaning it is applicable to the legacy fuel at Quad Cities.

Based on these considerations, the NRC staff determined that the application of penalties to account for TCD effects on fuel centerline melt and transient cladding strain, and the demonstration that a penalty is unnecessary for fuel rod internal pressure, provide assurance that the fuel rod thermal-mechanical evaluation continues to demonstrate compliance with Criterion 6 of the Quad Cities design basis, even though RODEX2A does not explicitly model

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION TCD. Therefore, the NRC staff determined that the treatment of TCD within the Quad Cities fuel thermal-mechanical analysis was acceptable.

Fuel Thermal-Mechanical Limits Conclusion Based on the review described above, the NRC staff determined that the thermal-mechanical limits provided in ANP-3918 are acceptable for application to Quad Cities. The methods are based on NRC-approved methods, demonstrate compliance with NRC-approved acceptance criteria, and include an acceptable augmentation for the issues described in IN 2009-23.

Therefore, there is reasonable assurance that, when operating in accordance with the limits specified in ANP-3918P, Quad Cities will remain consistent with Criterion 6 of its licensing basis.

3.2.2 Control Rod Drop Accident The amendment request proposes to transition from cold-worked, stress-relieved (CWSR)

Zircaloy-2 cladding associated with the ATRIUM 10XM fuel design, to recrystallized, annealed (RXA) Zircaloy-2 cladding associated with the GNF3 fuel design. The RXA cladding is more susceptible to fuel failure due to pellet-cladding mechanical interaction (PCMI) than the resident CWSR fuel cladding. Therefore, the NRC review focused on the methods to perform the CRDA and a technical justification for the acceptance criterion used for fuel cladding failure, the amount of deposited fuel enthalpy. The remaining fuel system damage criteria for GNF3 fuel following a postulated CRDA are addressed generically within NEDE-24011-P-A.

To implement the GNF3 fuel design, the licensee revised its radiological consequence analysis for the CRDA, among other licensing basis events. The estimated radiological consequences are based, in part, on a deterministic safety analysis described in NEDE-24011-P-A, Sections S.2.2.3.1.1 to S.2.2.3.1.4. The deterministic analysis uses 170 cal/gm of deposited fuel enthalpy as an acceptance criterion for fuel cladding failure, which is also consistent with the Quad Cities current licensing basis. Recent experimental evidence has shown that these damage thresholds are inadequate, and their use may lead to an underestimation of the true extent of cladding damage and attendant release of radioactivity. This information is presented in appendix B, Interim Acceptance Criteria and Guidance for the Reactivity Initiated Accidents, to section 4.2, Fuel System Design, of NUREG-0800 (ML070740002), and in NRC RG 1.236, Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents (ML20055F490).

The supplement dated April 11, 2022, provided a comparative evaluation of the extent of fuel failure predicted for GNF3 fuel to that predicted for the legacy Framatome and Westinghouse fuels, which are also 10x10 fuel bundle designs but are analyzed using different methods. This comparison indicated that the estimated extent of fuel cladding damage following the CRDA for GNF3 fuel was significantly higher than that predicted for either of the other fuel bundle designs.

Given that the accident characteristics are constrained by the banked position withdrawal sequence, which limits the peak reactivity of the dropped rod, this difference in predicted results is an indication that the GNF methods include conservative assumptions or modeling approaches.

In addition to the considerations identified by the licensee, the NRC staff independently reviewed documentation from the fuel vendor concerning the analytic methods and key assumptions in the CRDA in the following reports:

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NEDO-10527, Rod Drop Accident for Large Boiling Water Reactors (ML010870249),

MFN-034-087, Revised Generic Banked Position Withdrawal Sequence Control Rod Drop Analysis (ML102380315, non-public, proprietary), and

NEDC-33879P, Revision 4, GNF3 Generic Compliance with NEDE-24011-P-A (GESTAR II) (ML20244A104).

The GNF3 compliance document, NEDC-33879P, provides the estimated extent of fuel cladding damage for the GNF3 bundle design. It equates the cladding damage to that estimated for legacy fuel bundle designs including the legacy designs that were in use at the time GNF (then General Electric (GE)) submitted MFN-034-087 to the NRC. In MFN-034-087, the actual, deterministic analysis used to estimate the fuel cladding damage following a CRDA is described. Chapter 4 of MFN-034-087 identifies specific, ((

)).

Based on its review of the information provided by the licensee as well as other information available from the fuel vendor, the NRC staff determined that, while the licensee may apply an acceptance criterion for the CRDA analysis that could lead to an underestimation of the true extent of fuel cladding damage, there are additional conservatisms employed in the analysis that compensate.

In consideration of this information, the NRC staff determined that the CRDA analysis for the GNF3 fuel provides a technically adequate estimation of the extent of fuel cladding damage.

Because the licensee has acceptably addressed the consequences of the postulated CRDA, there is reasonable assurance that the applicable design criteria identified in Section 2.2.1, above, are met. Therefore, the NRC staff determined that the deterministic CRDA analysis for the GNF3 fuel is acceptable, insofar as it provides a basis for the inputs to the radiological consequences analysis, consistent with 10 CFR 50.67 requirements.

3.2.3 Mixed Core Thermal Mechanical Support Evaluation Conclusion Based on the review described above, the NRC staff determined that the proposed transition to GNF3 fuel at Quad Cities is acceptable with respect to the legacy fuel thermal-mechanical limits and the basis for estimated radiological consequences following a postulated CRDA.

3.3 Core Inventory Update and Resulting Dose Consequences The NRC staff established the requirements and methodologies for evaluating the radiological consequences of the postulated design-basis accidents (DBAs) using the dose criteria specified in 10 CFR 50.67 and the guidance described in RG 1.183. The requirements of 10 CFR 50.67 state that the applicable dose acceptance criteria are 5 rem (roentgen equivalent man) TEDE in the control room (CR), 25 rem TEDE at the exclusion area boundary (EAB), and 25 rem TEDE at the outer boundary of the low population zone (LPZ). RG 1.183 provides guidance to licensees on acceptable application of AST submittals, including acceptable radiological analysis assumptions and computer codes for use in developing DBA source terms and

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION accident dose consequence analysis. The NRC staff also considered relevant information in the Quad Cities UFSAR, TSs, and applicable previous licensing actions.

The original AST analyses for Quad Cities were submitted for NRC staff approval in a license amendment request dated October 10, 2002 (ML022940292). The submittal requested full implementation of AST and contained the radiological consequence analyses based upon the AST methodology for the following four DBAs that result in control room and offsite accident dose:

LOCA;

Main Steam Line Break (MSLB);

CRDA; and,

Fuel Handling Accident (FHA).

The submittal included changes to the Quad Cities TSs and associated licensing basis to reflect implementation of AST assumptions in accordance with 10 CFR 50.67. By letter dated September 11, 2006, the NRC approved the AST methodology and the associated TSs based on regulatory guidance provided in RG 1.183 (ML062070292).

Portions of the DBA analysis were subsequently revised in the approval of the LAR Application to Increase Technical Specifications Allowable MSIV Leakage Rates and Revise Secondary Containment Surveillance Requirement 3.6.4.1.1 (ML20150A328). This LAR revised the DBA LOCA in accordance with the previously approved AST methodology.

A modification to the licensing basis fuel type can have the potential to change the core isotopic distribution assumed in post-accident conditions, thereby affecting accident dose consequence analysis. To develop the core inventory used for the source term evaluation with the proposed GNF3 fuel, the licensee used the ORIGEN-ARP isotope generation and depletion computer code to develop a bounding equilibrium fission product core inventory. For the revised LOCA DBA accident dose consequence analysis, the licensee used a core inventory with no isotopic decay post shutdown. Consistent with RG 1.183, the licensee developed a bounding core inventory maximizing the fission product core inventory from the beginning of cycle to the end of cycle. Use of ORIGEN-ARP to develop the fission product core inventory and the assumption of an isotopic core inventory immediately after shutdown without decay thereby maximizing calculated accident dose consequence analysis is consistent with NRC guidance contained in RG 1.183, section 3.1, Fission Product Inventory.

The LAR stated that only the LOCAs radiological dose consequence analysis shows a more than minimal increase among the four DBAs. The LOCA DBA current licensing basis was updated in the June 26, 2020 Allowable Main Steam Isolation Leakage LAR approval (ML20150A328). All other previous DBA radiological dose consequence analysis have their current licensing basis established in the September 11, 2006 AST LAR. A minimal increase is defined as greater than 10 percent of the difference between the currently licensed dose consequence and the applicable regulatory limits (e.g., the acceptance criteria of 10 CFR 50.67 for AST and 10 CFR 50, appendix A, GDC 19 for CR dose consequences), with the updated fission product core inventory. Therefore, only the LOCA analysis was submitted for review.

As stated in the LAR, the FHA and CRDA DBAs radiological dose consequence analysis were analyzed utilizing the updated fission product core inventory. With the updated fission product core inventory, the FHA DBA is bounded by the current licensing basis dose consequence analysis, and the CRDA DBA radiological dose consequence analysis does not exceed the minimal increase criterion presented in NRC-endorsed guidelines of NEI 96-01, Guidelines for

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 10 CFR 50.59 Evaluations. The FHA and CRDA will be updated by the licensee under their approved 10 CFR 50.59 process.

LOCA The LAR stated and the NRC staff verified that all parameters, initial conditions and assumptions associated with the DBA LOCA radiological dose consequence analysis remain identical, with exception of the bounding core inventory, to the current licensing basis described in analysis (QDC-0000-N-1481, Revision 4) submitted with "Response to Request for Additional Information for the License Amendment Request to Change Technical Specifications to Increase Allowable MSIV Leakage Rates and Revise Secondary Containment Surveillance,"

dated March 31, 2020 (ML20091H576).

To support the LAR, the licensee revised the LOCA radiological consequence analysis using the previously approved LOCA initial conditions and assumptions with the revised bounding fission product core inventory. The revised calculation uses the updated bounding source term for the proposed GNF3 fuel design. All other methods, inputs and assumptions were approved in the amendment dated June 26, 2020 (ML20150A328) and remain unchanged. The revised analysis resulted in an increase in the calculated dose as shown in table 3-1: LOCA Dose Consequence Summary in attachment 1 of the LAR. The NRC staff compared the doses estimated and presented in the LAR and concluded that the radiological consequences at the EAB, LPZ, and in the CR are within the dose criteria specified in 10 CFR 50.67. The NRC staff performed independent confirmatory dose evaluations, as necessary, to ensure a thorough understanding of the licensees methods and assumptions using the RADTRAD code. The NRC staffs confirmatory calculations were consistent with licensee results and show that the CR, EAB, and LPZ, post-LOCA accident dose consequence analysis meet the applicable accident dose criteria, and therefore are acceptable.

Remaining DBAs which are affected by the proposed fuel transition, but do not result in more than a minimal increase in radiological dose consequence analysis will be updated under the Quad Cities 10 CFR 50.59 process.

FHA The LAR stated that the FHA DBA radiological dose consequence analysis with the updated fission product core inventory is bounded by the current licensing basis dose consequence values, and any calculation revisions will be implemented under the Quad Cities 10 CFR 50.59 process.

CRDA The LAR stated that the CRDA DBA radiological dose consequence analysis does not exceed the minimal increase criterion with the updated fission product core inventory, and any calculation revisions will be implemented under the Quad Cities 10 CFR 50.59 process.

MSLB The MSLB accident is described in Quad Cities UFSAR section 15.6.4. As stated in the UFSAR, no fuel damage is expected to result from a MSLB. The radionuclide inventory released from the primary coolant system is present in the coolant prior to the event. Therefore, MSLB accident analysis is not affected by a change in fuel design. Based upon this information, the NRC staff

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION finds that the proposed fuel design change does not alter the radiological consequences of a MSLB accident. The Quad Cities MSLB regulatory dose limits are unaffected and continues to meet the regulatory requirement in 10 CFR 50.67 and accident specific dose criteria described in SRP 15.0.1. and RG 1.183.

3.3.1 Meteorology Considerations The LAR uses an updated set of atmospheric dispersion factors (also known as /Q values and relative concentrations) to evaluate post-accident radiological consequences for the control room as well as offsite receptors at the EAB and LPZ. The NRC staff reviewed the meteorological data input and atmospheric dispersion estimates as described below.

Meteorological Data In support of the atmospheric dispersion analysis presented in the LAR, an hourly onsite meteorological dataset from January 1, 1995, through December 31, 1999, was used. The meteorological data was formatted for the ARCON96 atmospheric dispersion code (NUREG/CR-6331, Revision 1, Atmospheric Relative Concentrations in Building Wakes (ML17213A187)) to calculate updated /Q values for the main control room and offsite receptors. This format contained hourly data on wind speed, wind direction, and atmospheric stability class taken from the 10-m and 90-m levels of the onsite meteorological tower.

The NRC staff completed a detailed review related to the acceptability of the hourly meteorological data using the methodology described in NUREG-0917, Nuclear Regulatory Commission Staff Computer Programs for Use with Meteorological Data (ML12061A136).

Based on this review, the NRC staff considers the onsite meteorological database suitable for use in the atmospheric dispersion analyses to support this LAR.

Onsite Control Room Atmospheric Dispersion Estimates In support of the LAR, Quad Cities used the ARCON96 computer code to estimate /Q values for the main control room (MCR) for potential accidental releases of radioactive material.

RG 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants (ML031530505), endorses the ARCON96 model for determining /Q values to be used in the design basis evaluations of control room radiological habitability.

The ARCON96 code estimates /Q values for various time-averaged periods ranging from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 30 days. The meteorological input to ARCON96 consists of hourly values of wind speed, wind direction, and atmospheric stability class. The /Q values calculated through ARCON96 are based on the theoretical assumption that material released to the atmosphere will be normally distributed (Gaussian) about the plume centerline. A straight-line trajectory is assumed between the release points and receptors. The diffusion coefficients account for enhanced dispersion under low wind speed conditions and in building wakes.

The hourly meteorological data are used to calculate hourly relative concentrations. The hourly relative concentrations are then combined to estimate concentrations ranging in duration from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 30 days. Cumulative frequency distributions are determined from the average relative concentrations and the relative concentrations that are exceeded no more than five percent of the time for each averaging period, are determined.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The dispersion coefficients used in ARCON96 have three components. The first component is the diffusion coefficient, as discussed in NUREG/CR-6331. The other two components are corrections to account for enhanced dispersion under low wind speed conditions and in building wakes. These components are based on analysis of diffusion data collected in various building wake diffusion experiments under a wide range of meteorological conditions. Because the dispersion occurs at short distances within the plants building complex, the ARCON96 dispersion parameters are not affected by nearby topographic features such as bodies of water.

Therefore, the staff finds the use of the ARCON96 dispersion parameter assumptions consistent with the guidance provided in NUREG-0800, Section 2.3.4, Short-term Dispersion Estimates for Accident Releases, (ML070730398) and finds it acceptable.

The LAR provided the following as necessary input to ARCON96:

Onsite Hourly Meteorological Data: 1995 through 1999Input and log files for each run of the ARCON96 model The NRC staff confirmed the atmospheric dispersion estimates by running the ARCON96 computer model and obtaining similar results. Both the staff and the licensee used a ground level release assumption for each of the release/receptor combinations as well as the source-receptor distances provided in LAR Reference 9.14, Exelon Calculation No. QDC-0000-M-1408, Rev 2, Atmospheric Dispersion Factors (/Qs) for Accident Release. The staff used the source-receptor directions found in Reference 9.14 and found them to be acceptable, as discussed in the resulting Audit Report (ML22229A030). The NRC staff has reviewed the control room dispersion factors and finds the values to be acceptable.

The NRC staff reviewed the methodology used to derive the /Q values associated with postulated releases from potential release points. The NRC staff performed a screening analysis of the onsite meteorological data and found the licensee used appropriate atmospheric dispersion models to derive the resulting /Q values in accordance with staff regulatory guidance. On the basis of this review and the staffs confirmatory atmospheric dispersion analyses, the staff concluded that the meteorological data and the resulting control room /Q values followed the guidance provided in RG 1.194 and are therefore acceptable for use in the radiological consequence assessments supporting this LAR.

Offsite EAB and LPZ Atmospheric Dispersion Estimates In support of the LAR, the licensee used the NRC computer code PAVAN to estimate /Q values at the EAB and outer boundary of the LPZ for potential accidental releases of radioactive material. PAVAN implements the methodology described in RG 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants (ML003740205), for determining /Q values at the EAB and LPZ outer boundary.

The licensee calculated EAB and LPZ outer boundary /Q values using meteorological data from January 1, 1995, through December 31, 1999. The licensee chose to implement the diffusion parameter assumptions outlined in RG 1.145 as a function of atmospheric stability for its PAVAN model runs. The NRC staff evaluated the applicability of the PAVAN diffusion parameters and concluded that no unique topographic features (such as rough terrain, restricted flow conditions, or coastal or desert areas) preclude the use of the PAVAN model for the Quad

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Cities site. The NRC staff finds that the licensees use of the PAVAN diffusion parameters, as outlined in RG 1.145, to be acceptable.

The licensee modeled both a ground-level release point and an elevated release point to account for the station chimney stack. The NRC staff confirmed that including the building wake effects for a ground level release has minor influence on the predicted /Q values. A ground-level release assumption that assumes the appropriate building dimensions is acceptable to the staff because the PAVAN model includes both plume meander and building wake effects, which are mutually exclusive. The NRC staff also confirmed the use of an elevated release and finds it acceptable based on a stack height of 85.8 meters at the site. The licensee appropriately used the onsite meteorological measurement level of 90.2 meters to analyze the elevated releases.

Using the information provided by the licensee, including the 10-meter level joint frequency distributions of wind speed, wind direction, and atmospheric stability for ground level releases, and the 90.2-meter level for elevated releases, the NRC staff confirmed the licensees /Q values by running the PAVAN computer code and obtaining consistent results. On the basis of this review and the NRC staffs confirmatory calculations using PAVAN, the NRC staff concluded that the EAB and LPZ /Q values are acceptable for use in the proposed radiological consequence assessments in support of this LAR.

Meteorology Considerations Conclusion The NRC staff reviewed the methodology used by the licensee to derive the /Q values associated with postulated releases from potential release points. The NRC staff performed a screening of meteorological data and found the licensee used appropriate atmospheric dispersion models to derive the resulting /Q values in accordance with NRC staff RGs 1.23, 1.145, and 1.194. On the basis of this review and the staffs confirmatory atmospheric dispersion analyses, the NRC staff concluded that the meteorological data and the resulting onsite and offsite /Q values are acceptable for use in the radiological consequence assessments supporting this LAR.

3.3.2 Core Inventory Update and Resulting Dose Consequences Summary and Conclusions The NRC staff reviewed the analyses presented in the LAR to assess the radiological impacts of the transition from ATRIUM 10XM fuel to GNF3 fuel design. The staff finds that the methods used are consistent with the regulatory requirements and guidance associated with accident dose consequence analysis. The staff also finds, with reasonable assurance that results of the EAB, LPZ, and control room doses associated with the revised LOCA DBA comply with the applicable regulations and regulatory guidance that is included in the Quad Cities current licensing basis. Therefore, the proposed change is acceptable with regard to the radiological consequences of the revised LOCA DBA.

3.4 EQ Impacts Section 3.4, Environmental Qualification Impacts, of attachment 1 of the LAR, stated that the change in core inventory has no impact on normal or post-accident temperature, pressure, or humidity. Further, the LAR stated that the GNF3 core inventory does not affect the normal operating doses for EQ and a detailed review of EQ equipment was conducted using the GNF3 scaled total integrated dose (i.e., accident plus the normal radiation dose).

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The letter dated November 3, 2021, provided supplemental information and stated that the core inventory changes associated with the transition to GNF3 fuel cause the numerical value of the accident dose used in the Quad Cities EQ program to be revised. In addition, the LAR 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. The licensee concluded that the fuel transition does not compromise the ability of any qualified equipment to continue to perform its required safety function and thus, continued compliance with 10 CFR 50.49.

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 reviewed the 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 May 4, 2022 (ML22098A083).

The letter dated November 3, 2021, stated that the EQ documentation revisions are being tracked and prioritized using the appropriate internal processes to ensure timely revision.

EQ Impacts Summary Based on its review of the information in the LAR and its supplement as well as the results of the regulatory audit, the NRC staff finds that the proposed change should have no adverse impact on the Quad Cities EQ program or its ability to continue to meet the requirements of 10 CFR 50.49 and IEB 79-01B.

3.5 TS Change The NRC staff determined that the proposed TS change meets the standards for TS in 10 CFR 50.36 because the modified TS 5.6.3.b addresses addition of the GNF Report NEDC-33930P by number, title, date, and identifies the report by the staffs approved SE with a reference to the NRCs dated letter and thus satisfies the 10 CFR 50.36(c)(5) requirement.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations on, the Illinois State official was notified of the proposed issuance of the amendment on November 3, 2022. The State official had no comments.

5.0 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 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding January 4, 2022 (87 FR 256). Accordingly, the amendments meet the eligibility criteria for

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION 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.

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 Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors:

Sean Meighan, NRR Sheila Ray, NRR Matthew McConnell, NRR Benjamin Parks, NRR Joshua Kaizer, NRR Ravi Grover, NRR Date of Issuance: December 15, 2022

ML22298A003 (Non-Public); ML22298A002 (Public)

OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/STSB/BC NRR/DRA/ARCB/BC NAME RKuntz SRohrer VCusumano KHsueh DATE 10/20/2022 10/31/22 10/28/2022 10/28/2022 OFFIC NRR/DEX/ELTB/BC NRR/DSS/SFNB/BC NRR/DEX/EXHB/BC OGC NAME JPaige SKrepel BHayes LShrum DATE 10/25/2022 10/20/2022 10/12/22 11/17/22 OFFICE NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME NSalgado RKuntz DATE 12/15/22 12/15/22