ML22308A160

– Issuance of Amendment Nos. 291 and 287 New Fuel Vault and Spent Fuel Storage Pool Criticality Methodologies

ML22308A160
Person / Time
Site:	Quad Cities
Issue date:	12/02/2022
From:	Robert Kuntz NRC/NRR/DORL/LPL3
To:	Rhoades D Constellation Energy Generation
References
EPID L-2021-LLA-0196
Download: ML22308A160 (19)

Text

December 2, 2022 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. 291 AND 287 RE: NEW FUEL VAULT AND SPENT FUEL STORAGE POOL CRITICALITY METHODOLOGIES (EPID L-2021-LLA-0196)

Dear Mr. Rhoades:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 291 to Renewed Facility Operating License No. DPR-29 and Amendment No. 287 to Renewed Facility Operating License No. DPR-30 for Quad Cities Nuclear Power Station, Units 1 and 2 (Quad Cities), respectively. The amendments consist of changes to the technical specifications and Updated Final Safety Analysis Report in response to your application dated October 25, 2021 (Agencywide Documents Access and Management System Accession No. ML21298A168), as supplemented by letters dated July 13, 2022 (ML22194A085), and October 5, 2022 (ML22278A149).

The amendments revise the criticality safety analysis (CSA) methodology for performing the criticality safety evaluation for legacy fuel types in addition to the Global Nuclear Fuel -

Americas, LLC (GNF) GNF3 fuel in the Quad Cities spent fuel pool. The amendments also change the new fuel vault (NFV) CSA for storing GNF3 fuel in the NFV racks.

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/

Robert F. 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. 291 to DPR-29

2. Amendment No. 287 to DPR-30

3. 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. 291 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 October 25, 2021, as supplemented by letters dated July 13, 2022, and October 5, 2022, 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-29 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 291, 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 application, as supplemented.

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 2, 2022 Nancy L.

Salgado Digitally signed by Nancy L. Salgado Date: 2022.12.02 10:41:16 -05'00'

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. 287 Renewed License No. DPR-30

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 October 25, 2021, as supplemented by letters dated July 13, 2022, and October 5, 2022, 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-30 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 287, 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 application, as supplemented.

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 2, 2022 Nancy L.

Salgado Digitally signed by Nancy L. Salgado Date: 2022.12.02 10:41:50 -05'00'

ATTACHMENT TO LICENSE AMENDMENT NOS. 291 AND 287 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 revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas 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 4.0-2 4.0-2 Renewed License No. DPR-29 Amendment No. 291 B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 291, 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. 287 B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 287, 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.

Design Features 4.0 Quad Cities 1 and 2 4.0-2 Amendment No. 291/2874/248 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a.

keff 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1.2 of the UFSAR; b.

A nominal 6.22 inch center to center distance between fuel assemblies placed in the storage racks; c.

Fuel assemblies having a maximum kinf of 1.29 in the normal reactor core configuration at cold conditions; and d.

The installed neutron absorbing rack inserts having a Boron-10 areal density 0.0116 g/cm2.

4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 666 ft 8.5 inches.

4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 3657 fuel assemblies for Unit 1 and 3897 fuel assemblies for Unit 2.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 291 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-29 AND AMENDMENT NO. 287 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 October 25, 2021 (Agencywide Documents Access and Management System Accession No. ML21298A168), Exelon Generation Company, LLC submitted a request for amendments to Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station, Units 1 and 2 (Quad Cities), respectively. The amendments proposed changes to the Technical Specifications (TSs) and Updated Final Safety Analysis Report (UFSAR). On February 1, 2022 (ML22032A333), Exelon Generation Company, LLC was renamed Constellation Energy Generation, LLC (Constellation, the licensee). Subsequent to this name change, Constellation supplemented the original license amendment request (LAR) by letters dated July 13, 2022 (ML22194A085), and October 5, 2022 (ML22278A149). Specifically, the LAR, as supplemented, proposed a new criticality safety analysis (CSA) methodology for performing the criticality safety evaluation for legacy fuel types in addition to the Global Nuclear Fuel - Americas, LLC (GNF) GNF3 reload fuel in the Quad Cities spent fuel pool (SFP). The LAR initially proposed a change to the new fuel vault (NFV) CSA to utilize the GESTAR II methodology. An NRC staff audit of the GESTAR II CSA methodology revealed that there is not an NRC-approved GESTAR II CSA methodology. Subsequently, the licensee submitted a CSA for storing GNF3 fuel in the Quad Cities NFV.

The supplemental letters dated July 13, 2022, and October 5, 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 April 5, 2022 (87 FR 19715).

2.0 REGULATORY EVALUATION

2.1 Description of Proposed Changes The Quad Cities renewed licenses do not have TSs for the NFV. The October 25, 2021, letter proposed changes to the UFSAR to reflect storage requirements for GNF3 fuel in the NFV and submitted a change to the methodology for calculating k-effective for the NFV in the fully flooded condition. The October 5, 2022, letter submitted a new CSA to support storing GNF3 fuel in the NFV. The October 25, 2021, letter stated that optimum moderation is prevented in the NFV. The July 13, 2022, letter provided supplemental information regarding that statement.

The LAR, as supplemented, also proposed a change to Quad Cities TS 4.3, Fuel Storage, 4.3.1, Criticality, design feature 4.3.1.1.c to change the SFP storage criteria from an SFP k-infinity of 0.8991 to a maximum k-infinity (abbreviated kinf in the proposed TS) of 1.29 in the normal reactor core configuration at cold conditions (also known as standard cold core geometry (SCCG) k-infinity). The October 25, 2021, letter contained a CSA supporting the requested TS change. The July 13, 2022, letter submitted a revised CSA and supplemental information.

2.2 Applicable Regulatory Requirements and Guidance 2.2.1 Regulatory Requirements Paragraph 50.68(a) of Title 10 of the Code of Federal Regulations (10 CFR) states, in part, that Each holder of a construction permit or operating license for a nuclear power reactor issued under this part shall comply with either 10 CFR 70.24 of this chapter or the requirements in paragraph (b) of this section. With respect to Quad Cities, the licensee has chosen to comply with 10 CFR 50.68(b).

Paragraph 50.68(b)(1) of 10 CFR states that Plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.

Paragraph 50.68(b)(2) of 10 CFR states that The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.

Paragraph 50.68(b)(3) of 10 CFR states that If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.

Paragraph 50.68(b)(4) of 10 CFR states, in part, that If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. Paragraph 50.68(b)(4) of 10 CFR additionally provides requirements for when credit is taken for soluble boron; however, the Quad Cities SFP nuclear criticality safety (NCS) analysis does not take credit for soluble boron and, therefore, those requirements do not apply.

In addition, 10 CFR 50.36(c)(4), Design features, states that design features to be included in TSs are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c) (1), (2), and (3) of this section.

The Quad Cities construction permits predated the formal issuance of the current 10 CFR part 50, Domestic Licensing of Production and Utilization Facilities, appendix A, General Design Criteria for Nuclear Power Plants. Therefore, the Quad Cities construction permits were evaluated against the set of 70 draft general design criteria (GDCs) published for public comment in the Federal Register (32 FR 10213) by the Atomic Energy Commission on July 11, 1967. The design bases of each Quad Cities unit were reevaluated at the time of initial Final Safety Analysis Report preparation against the 70 draft GDCs at the time of operating license application. As stated in section 3.1 of the Quad Cities updated final safety analysis report (UFSAR) (ML22049A114), based on the licensees understanding of the intent of the 70 draft GDCs, Quad Cities fully satisfies the intent of those GDCs.

Section 3.1.8.1 of the Quad Cities UFSAR provides Criterion 66, Prevention of Fuel Storage Criticality, which states that Criticality in the new and spent fuel storage shall be prevented by physical systems or processes. Such means as geometrically safe configurations shall be emphasized over procedural controls. Quad Cities Criterion 66 is essentially the same as 10 CFR part 50, appendix A, GDC 62, Prevention of criticality in fuel storage and handling, which states that Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.

2.2.2 Guidance Regulatory Guide (RG) 1.240, Fresh and Spent Fuel Pool Criticality Analyses (ML20356A127),

describes an approach that the NRC staff considers acceptable to demonstrate that regulatory requirements are met for subcriticality of fuel assemblies stored in fresh fuel vaults and spent fuel pools at light-water reactor power plants. It endorses, with clarifications and exceptions, the Nuclear Energy Institute (NEI) guidance document NEI 12-16, Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants, Revision 4 (ML19269E069). The LAR stated that NEI 12-16, Revision 4, was used in its preparation.

NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology (ML010170125), provides guidance for performing criticality safety analyses.

3.0 TECHNICAL EVALUATION

The NRC staff reviewed the LAR, as supplemented, to determine whether the proposed revised CSAs for the Quad Cities NFV and SFP meet the requirements in 10 CFR 50.68. The CSA for storing GNF3 in the Quad Cities SFP is described in GNF NEDC-33932P, Quad Cities Units 1 and 2 Fuel Storage Criticality Safety Analysis, Revision 2, July 2022, attachment 7 to the July 13, 2022, letter, as supplemented by the letter dated July 13, 2022 (the non-proprietary version of this analysis, NEDO-33932, is provided in attachment 2 to this letter). The CSA for storing GNF3 in the Quad Cities NFV is described in GE Hitachi Nuclear Energy (GEH) 003N7421-P, Generic Criticality Safety Analysis of GE New Fuel Storage Racks for GNF3 Fuel, Revision 1, September 2022, attachment 4 to the October 5, 2022, letter (the non-proprietary version of this analysis, GEH 003N7421-NP, is provided in attachment 2 to this letter).

3.1 Computer Code Versions and Applications The LAR analyses use TGBLA06 for in-core calculations and MCNP-05P for in-rack calculations.

TGBLA06 is a lattice physics computer code that calculates the exposure dependent pin-by-pin isotopic specifications used in developing the design basis lattice for the SFP CSA, but also has application to many other GEH/GNF analysis methods. Letter from S. Richards, NRC, to G.

Watford, GE Nuclear Energy, Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, GESTAR II - Implementing Improved GE Steady-State Methods (TAC No. MA6481), dated November 10, 1999 (ML993230184), documents the NRCs original acceptance of TGBLA06 and its associated application methodology. This initial approval was updated to address the applicability to the GNF3 product line in a letter from B. Moore, GNF - Americas, LLC, to the NRC, Administrative Amendment 49 to NEDE-24011-P-A-27, General Electric Standard Application for Reactor Fuel (GESTAR II), dated October 1, 2018 (ML18274A195). Based on the prior NRC review and approval of TGBLA06 for depletion and reactivity calculations, the NRC staff found the use of this code for the purpose described in the LAR analyses to be acceptable.

The Monte Carlo N-Particle (MCNP) computer code can be used for the general-purpose transport of many particles including neutrons, photons, electrons, ions, and many other elementary particles. MCNP is developed, maintained, and distributed by the Los Alamos National Laboratory. MCNP-05P is the GEH/GNF proprietary version of MCNP5. MCNP-05P is a Monte Carlo program for solving the linear neutron transport equation for a fixed source or an eigenvalue problem. The code implements the Monte Carlo process for neutron, photon, electron, or coupled transport involving all these particles, and computes the eigenvalue for neutron-multiplying systems. For the Quad Cities NFV and SFP CSAs only neutron transport is considered. Based on the generally accepted use of MCNP for neutron transport calculations and the CSA-specific validation described below, the NRC staff found the use of this code for the purpose described in the LAR analyses to be acceptable.

3.2 Computer Code Validation The purpose of the computer code validation is to ensure that appropriate code bias and bias uncertainty are determined for use in the criticality calculation. The licensee followed NEI 12-16, Revision 4, guidance for determining the depletion uncertainty for TGBLA06. The licensee followed NUREG/CR-6698 to validate and determine a code bias and bias uncertainty for MCNP-05P. Using NUREG/CR-6698 to validate the Monte Carlo code is consistent with NEI 12-16, Revision 4. Both are consistent with the endorsement in RG 1.240 of NEI 12-16, Revision 4, and, therefore, the NRC staff found the validation to be acceptable.

3.3 New Fuel Vault Paragraph 50.68(b) of 10 CFR has two subparagraphs that address the storage of fuel in the NFV. Both are accident driven as the NFV is dry/unmoderated. Paragraph 50.68(b)(2) of 10 CFR addresses the scenario should the NFV become fully flooded. Paragraph 50.68(b)(3) of 10 CFR addresses the scenario should the NFV be the subject of an optimum moderation condition. Typically, the source of the optimum moderation is assumed to be firefighting water or aqueous foam from firefighting efforts in the building housing the NFV. If the building housing the NFV is susceptible to environmental damage, that could be another source of moderating medium.

With respect to the fully flooded scenario, the October 25, 2021, letter stated that the Quad Cities NFV CSA methodology was being changed to that in GESTAR II and cited references.

This letter also stated that there was a GNF3 GESTAR II NFV CSA validation report, again citing references. However, the cited references did not contain an NFV GESTAR II methodology or a GNF3 GESTAR II NFV CSA validation report. The July 13, 2022, letter did not provide this methodology or validation report. To expedite its review, the NRC staff conducted a virtual audit of GESTAR IIs NFV criticality coverage. That audit was held from August 4, 2022, until September 2, 2022 (a report of the audit activities is available at ML22300A253). The audit revealed that there is no NRC-approved GESTAR II NFV or SFP CSA methodology. The audit also revealed that there is a generic GNF3 NFV CSA covering storage of GNF3 in GE-designed NFV racks, but that that CSA hadnt previously been made available to the NRC. The licensee submitted the generic GNF3 NFV CSA covering storage of GNF3 in GE-designed NFV racks with the October 5, 2022, letter.

The CSA for storing GNF3 in the Quad Cities NFV is described in GEH 003N7421-P, included with the October 5, 2022, letter. The analysis is generic and is based on demonstrating that GNF3 fuel with a maximum cold, uncontrolled peak in-core (otherwise known as SCCG) k-infinity of 1.31 will satisfy 10 CFR 50.68(b)(2) when stored in either of two GE rack types. As long as a particular licensee has one of the two GE rack types and meets the cell pitch listed in GEH 003N7421-P, table 1-1, New Fuel Vault Rack Dimensions, then the analysis would be representative of the applicable NFV.

The NFV CSA is a fresh fuel analysis with the SCCG calculated using TGBLA06. As a fresh fuel analysis, core depletion impact is not a factor. With the fuel design set as GNF3, the SCCG is essentially set by a combination of Uranium-235 enrichment and gadolinia loading. The fuel design identified by TGBLA06 is then modeled in the NFV storage racks with MCNP-05P to calculate the in-rack k-infinity. It is this in-rack k-infinity that is used to demonstrate compliance with 10 CFR 50.68(b)(2). The licensee checks each fuel assemblys peak lifetime reactivity against the SCCG k-infinity of 1.31. Therefore, establishing the relationship between the fuel assemblys maximum cold, uncontrolled peak in-core k-infinity and its in-rack k-infinity is required. GEH 003N7421-P establishes that relationship by calculating a rack efficiency for each fuel design.

GEH 003N7421-P calculated an in-rack k-infinity of 0.93919 at a 95 percent probability, 95 percent confidence level for a fresh GNF3 fuel assembly with a SCCG k-infinity of 1.31. This represents 0.01 k-effective of margin to the regulatory limit of 0.95 in 10 CFR 50.68(b)(2). The NRC staff considered this margin in its review to apply engineering judgement to the depth of the review on items unlikely to challenge the margin.

The NRC staff used RG 1.240 to review the analysis in GEH 003N7421-P. RG 1.240 provides guidance for salient aspects of an NFV CSA such as: determining the reactivity effects of fuel assembly manufacturing tolerances and rack tolerances; evaluating potential biases such as fuel assembly eccentricity within the rack storage cells; and selecting representative fuel assembly designs for the licensees total inventory. The information provided in GEH 003N7421-P indicates that these topics were treated in a manner consistent with the guidance.

The NRC staff determined that the information provided in GEH 003N7421-P, along with the margin in the analysis of 0.01 k-effective, provide reasonable assurance that the requirements in 10 CFR 50.68(b)(2) will be met.

With respect to the 10 CFR 50.68(b)(3) optimum moderation scenario, the October 25, 2021, letter stated that the The optimum moderation case is not applicable to the [Quad Cities] NFV as it is a moderation controlled area. The NRC staff notes that 10 CFR 50.68(b)(3) still applies even if a licensee has administrative controls and/or design features to prevent such moderation; however, the method of compliance changes from an analysis and operation in accordance with that analysis that demonstrates sufficient sub-critical margin to the maintenance of the administrative controls and/or design features and operation within those.

Regardless, the NRC staff determined that the optimum moderation scenario is outside the scope of the LAR.

GEH 003N7421-P contains analyses for a fully flooded and an optimum moderated NFV. As noted, the optimum moderated scenario was considered out of scope for this LAR. Therefore, the NRC staff did not review that portion of GEH 003N7421-P and makes no determination regarding the optimum moderation scenario. As discussed above, the NRC staff finds the CSA for the fully flooded NFV to be acceptable to demonstrate compliance with the applicable 10 CFR 50.68 regulations based on consistency with NRC regulatory guidance and a technically justified approach.

3.4 Spent Fuel Pool Paragraph 50.68(b) of 10 CFR has one subparagraph that addresses storage of fuel in the SFP.

Paragraph 50.68(b)(4) of 10 CFR states, in part, that If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. Paragraph 50.68(b)(4) of 10 CFR additionally provides requirements for when credit is taken for soluble boron; however, the Quad Cities SFP NCS analysis does not take credit for soluble boron and, therefore, those requirements do not apply.

There is no optimum moderation paragraph for SFPs since TS 4.3.2, Drainage, ensures a minimum water level in the SFP providing reasonable assurance that an optimum moderation will not occur.

The CSA for storing GNF3 in the Quad Cities SFP is described in GNF NEDC-33932P. The analysis is specific to the Quad Cities SFP demonstrating that GNF3 fuel with a maximum cold, uncontrolled peak in-core (SCCG) k-infinity of 1.29 will satisfy 10 CFR 50.68(b)(4) with no credit taken for soluble boron. The analysis includes consideration for Quad Cities legacy fuel, stating that the legacy fuel is bounded by the analysis in NEDC-33932P.

The SFP CSA is an irradiated fuel analysis with the SCCG calculated using TGBLA06. As an irradiated fuel analysis, the core depletion impact is a significant factor. In addition to the fuel design, Uranium-235 enrichment, and gadolinia loading, the reactor operating conditions while the fuel is being used will impact its reactivity.

The fuel design with a specified Uranium-235 enrichment and gadolinia loading is modeled by TGBLA06 over a range of reactor operating parameters. The irradiated fuel assembly is then modeled with MCNP-05P to calculate the in-rack k-infinity. This can be an iterative process until the licensee finds a solution that meets the regulatory requirement and any other needs it might have. Only the final set of analyses are provided in the LAR.

As with the NFV analyses, the SFP in-rack k-infinity is used to demonstrate compliance with 10 CFR 50.68(b)(4). The licensee checks each fuel assemblys peak lifetime reactivity against the SCCG k-infinity of 1.29. Therefore, establishing the relationship between the fuel assemblys maximum cold, uncontrolled peak in-core k-infinity and its in-rack k-infinity is required. NEDC-33932P establishes that relationship by calculating a rack efficiency for each fuel design.

NEDC-33932P calculated an in-rack k-infinity of 0.94200 at a 95 percent probability, 95 percent confidence level for an irradiated GNF3 fuel assembly with a SCCG k-infinity of 1.29. This represents 0.008 k-effective of margin to the regulatory limit in 10 CFR 50.68(b)(4). The NRC staff considered this margin in its review to apply engineering judgement to the depth of the review on items unlikely to challenge the margin.

The NRC staff used RG 1.240 to review the analysis in NEDC-33932P. RG 1.240 provides guidance for salient aspects of an SPF CSA such as: modeling depletion parameters and uncertainty; determining the reactivity effects of fuel assembly manufacturing tolerances and rack tolerances; evaluating potential biases such as fuel assembly eccentricity within the rack storage cells; and selecting representative fuel assembly designs for the licensees total inventory. Many of the general steps are the same between an unirradiated and irradiated CSA, but the irradiated CSA is typically more detailed due to the additional complexity of the fuels depletion in the reactor during operation. The information provided in NEDC-33932P indicates that these topics were treated in a manner consistent with the guidance. The NRC staff determined that the information provided in NEDC-33932P, along with the margin in the analysis of 0.008 k-effective, provide reasonable assurance that the requirements in 10 CFR 50.68(b)(4) will be met. Therefore, the NRC staff found the licensees justification for the proposed TS 4.3.1.1.c change to be acceptable and that 10 CFR 50.36(c)(4) will continue to be met.

3.5 Technical Evaluation Conclusion

The NRC staff finds the CSA for the fully flooded NFV to be acceptable to demonstrate compliance with the applicable 10 CFR 50.68 regulations based on consistency with NRC regulatory guidance and a technically justified approach.

The licensee has provided reasonable assurance that a GNF3 fuel assembly with a peak SCCG of 1.31 k-infinity will comply with 10 CFR 50.68(b)(2). The licensee has also provided reasonable assurance that a GNF3 fuel assembly with a peak SCCG of 1.29 k-infinity will comply with 10 CFR 50.68(b)(4). Therefore, Quad Cities Criterion 66, which is essentially the same as 10 CFR part 50, appendix A, GDC 62, will continue to be met.

The NRC staff notes that the difference in the two peak SCCG k-infinity numbers means that fuel that is acceptable for storage in the NFV is not acceptable for storage in the SFP unless it also meets the peak SCCG of 1.29 k-infinity for the SFP.

The licensee has provided an acceptable justification for the proposed TS 4.3.1.1.c change, thereby satisfying 10 CFR 50.36(c)(4).

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Illinois State official was notified of the proposed issuance of the amendments on November 3, 2022. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to the installation or use of facility 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 published in the Federal Register on April 5, 2022 (87 FR 19715). 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.

6.0 CONCLUSION

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

Principal Contributors: Kent Wood, NRR Date of Issuance: December 2, 2022

ML22308A160 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/STSB/BC NRR/DSS/SFNB/BC NAME RKuntz SRohrer VCusumano SKrepel DATE 11/4/2022 11/7/2022 11/8/2022 11/2/2022 OFFIC OGC NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME JWatchuka NSalgado RKuntz DATE 11/30/2022 12/2/2022 12/2/2022