ML22192A104

Acceptance of Requested Licensing Action Amendment Request Application to Allow Use of Lead Test Assemblies for Accident Tolerant Fuel with Request for Additional Information

ML22192A104
Person / Time
Site:	Vogtle
Issue date:	08/01/2022
From:	John Lamb Plant Licensing Branch II
To:	Gayheart C Southern Nuclear Operating Co
Lamb J
References
EPID L-2022-LLA-0097
Download: ML22192A104 (12)

Text

August 1, 2022 Ms. Cheryl A. Gayheart Regulatory Affairs Director Southern Nuclear Operating Co., Inc.

3535 Colonnade Parkway Birmingham, AL 35243

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 - ACCEPTANCE OF REQUESTED LICENSING ACTION RE: AMENDMENT REQUEST; APPLICATION TO ALLOW USE OF LEAD TEST ASSEMBLIES FOR ACCIDENT TOLERANT FUEL WITH REQUEST FOR ADDITIONAL INFORMATION (EPID L-2022-LLA-0097)

Dear Ms. Gayheart:

By letter dated June 30, 2022 (Agencywide Documents and Access Management System (ADAMS) Accession No. ML22181B156), Southern Nuclear Operating Company (SNC, the licensee) submitted a license amendment request (LAR) for Vogtle Electric Generating Plant (Vogtle), Units 1 and 2. The proposed LAR would allow the use of lead test assemblies (LTAs) for accident tolerant fuel (ATF).

The purpose of this letter is to provide the results of the U.S. Nuclear Regulatory Commission (NRC) staffs acceptance review of this LAR. The acceptance review was performed to determine if there is sufficient technical information in scope and depth to allow the NRC staff to complete its detailed technical review. The acceptance review is also intended to identify whether the application has any readily apparent information insufficiencies in its characterization of the regulatory requirements or the licensing basis of the plant.

Consistent with Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), an application for an amendment to a license (including the technical specifications) must fully describe the changes requested, and following as far as applicable, the form prescribed for original applications. Section 50.34 of 10 CFR addresses the content of technical information required. This section stipulates that the submittal address the design and operating characteristics, unusual or novel design features, and principal safety considerations.

The NRC staff has reviewed your application and concluded that it does provide technical information in sufficient detail to enable the NRC staff to begin its detailed technical review and make an independent assessment regarding the acceptability of the proposed LAR in terms of regulatory requirements and the protection of public health and safety and the environment.

However, enclosed are requests for additional information that the NRC staff has identified as necessary early in its review. The NRC staff requests that SNC supplement the application to address the information requested in the enclosure within 30 working days from the date of this letter to keep the technical review on schedule.

C. Gayheart Given the lesser scope and depth of the acceptance review as compared to the detailed technical review, there may be instances in which issues that impact the NRC staffs ability to complete the detailed technical review are identified despite the NRC staffs determination that the application is acceptable for the Staffs technical review to commence. You will be advised of any further information needed to support the NRC staffs detailed technical review by separate correspondence.

Based on the information provided in your submittal and discussions during the pre-submittal meetings on January 27, 2022 (ML22028A046) and May 11, 2022 (ML22132A010), the NRC staff has estimated that this licensing review will take approximately 1200 hours0.0139 days <br />0.333 hours <br />0.00198 weeks <br />4.566e-4 months <br /> to complete.

The NRC staff expects to complete this review approximately 1 year from the date of this letter.

If there are emergent complexities or challenges in our review that would cause changes to the initial forecasted completion date or significant changes in the forecasted hours, the reasons for the changes, along with the new estimates, will be communicated during the routine interactions with the assigned project manager.

These estimates are based on the NRC staffs initial review of the application, and they could change, due to several factors including the need for requests for additional information, unanticipated addition of scope to the review, review by NRC advisory committees, or hearing-related activities. Additional delay may occur if the LAR submittal is provided to the NRC in advance of or in parallel with industry program initiatives or pilot applications.

If you have any questions, please contact me at (301) 415-3100 or by email at John.Lamb@nrc.gov.

Sincerely, John G. Lamb, Senior Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-424 and 50-425

Enclosure:

As stated cc: Listserv John G. Lamb Digitally signed by John G. Lamb Date: 2022.08.01 14:03:20 -04'00'

VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION REGARDING AMENDMENT REQUEST APPLICATION TO ALLOW USE OF LEAD TEST ASSEMBLIES FOR ACCIDENT TOLERANT FUEL DOCKET NOS. 50-424 AND 50425 BACKGROUND On January 27, 2022, (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML22028A046), a pre-submittal public meeting was held between the staff of the U.S. Nuclear Regulatory Commission (NRC) and representatives of Southern Nuclear Operating Company, Inc. (SNC). SNC stated that the proposed license amendment request (LAR) will revise (1) Technical Specification (TS) 4.2.1, Fuel Assemblies, and (2) TS 4.3.1, Criticality. SNC stated the proposed LAR could potentially revise TS 3.7.18, Fuel Assembly Storage in the Fuel Storage Pool.

On May 11, 2022 (ML22132A010), a partial open and partial closed pre-submittal meeting was held between the NRC staff and representatives of SNC. SNC presented slides 8 and 9 (ML22126A001) regarding a proposed license condition. During that meeting, the NRC staff expressed concerns with the proposed license condition in the absence of conforming changes to the applicable TS, and the NRC staff informed the licensee that the license condition and TS could not be inconsistent and that conforming changes were likely needed, as described in the January pre-submittal meeting discussions.

REQUEST FOR ADDITIONAL INFORMATION Introduction By letter dated June 30, 2022 (ML22181B156), Southern Nuclear Operating Company (SNC, the licensee) submitted a LAR for Vogtle Electric Generating Plant (Vogtle), Units 1 and 2. The proposed LAR would allow the use of lead test assemblies (LTAs) for accident tolerant fuel (ATF).

During its acceptance review of the LAR, the NRC staff requests that SNC supplement the application to address the information requested below, within 30 working days from the date of this letter. Timely submittal of the requested information would help keep the technical review on schedule.

Question 1 Regulatory Basis The regulation in Section 50.36(c)(4) of the Title 10 of the Code of Federal Regulations (10 CFR) describes what Design Features information shall be included in TSs. The regulation states: Design features to be included 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 information included in the proposed license condition appears to meet the definition of design features.

Question 1 The Vogtle Renewed Facility Operating License (RFOL) states, in part, that:

C.

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

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 214 [for Unit 1, and Amendment No. 197 for Unit 2], and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

TS 4.2.1, Fuel Assemblies, states:

The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy, ZIRLO, or Optimized ZIRLO' clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02) as fuel material.

Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

TS 4.3.1, Criticality, states:

(Unit 1) 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent; b.

Keff < 1.0 when fully flooded with unborated water which includes an allowance for uncertainties as described in Section 4.3 of the FSAR.

c.

Keff < 0.95 when fully flooded with water borated to 511 ppm, which includes an allowance for uncertainties as described in Section 4.3 of the FSAR;

d. New or partially spent fuel assemblies with a combination of burnup and initial nominal enrichment in the acceptable burnup domain of Figures 3.7.18-1 or satisfying a minimum Integral Fuel Burnable Absorber (IFBA) requirement as shown In Figure 4.3.1-7 may be allowed unrestricted storage in the Unit 1 fuel storage pool.

e. New or partially spent fuel assemblies with a maximum initial enrichment of 5.0 weight percent U-235 may be stored in the Unit 1 fuel storage pool in a 3-out-of-4 checkerboard storage configuration as shown in Figure 4.3.1-1.

Interfaces between storage configurations in the Unit 1 fuel storage pool shall be in compliance with Figure 4.3.1-3. "A" assemblies are new or partially spent fuel assemblies with a combination of burnup and initial nominal enrichment in the acceptable burnup domain of Figure 3.7.18-1, or which satisfy a minimum IFBA requirement as shown in Figure 4.3.1-7. "B assemblies are assemblies with initial enrichments up to a maximum of 5.0 weight percent U-235.

f. A nominal 10.25 inch center to center pitch in the Unit 1 high density fuel storage racks.

(Unit 2) 4.3.1.2 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;

b. Keff < 1.0 when fully flooded with unborated water which includes an allowance for uncertainties as described in Section 4.3 of the FSAR.

c. Keff < 0.95 when fully flooded with water borated to 394 ppm, which includes an allowance for uncertainties as described in Section 4.3 of the FSAR;

d. New or partially spent fuel assemblies with a combination of burnup and initial nominal enrichment in the acceptable burnup domain of Figure 3.7.18-2 may be allowed unrestricted storage in the Unit 2 fuel storage pool.

e. New or partially spent fuel assemblies with a combination of burnup and initial nominal enrichment in the "acceptable burnup domain" of Figure 4.3.1-8 may be stored in the Unit 2 fuel storage pool in a 3-out-of-4 checkerboard storage configuration as shown In Figure 4.3.1-1.

New or partially spent fuel assemblies with a maximum initial enrichment of 5.0 weight percent U-235 may be stored in the Unit 2 fuel storage pool in a 2-out-of-4 checkerboard storage configuration as shown in Figure 4.3.1-1.

New or partially spent fuel assemblies with a combination of burnup, decay time, and initial nominal enrichment in the acceptable burnup domain of Figure 4.3.1-10 may be stored in the Unit 2 fuel storage pool as "low enrichment' fuel assemblies in the 3x3 checkerboard storage configuration as shown in Figure 4.3.1-2. New or partially spent fuel assemblies with initial nominal enrichments less than or equal to 3.20 weight percent U-235 or which satisfy a minimum IFBA requirement as shown in Figure 4.3.1-9 for higher initial enrichments may be stored in the Unit 2 fuel storage pool as high enrichment fuel assemblies in the 3x3 checkerboard storage configuration as shown in Figure 4.3.1-2.

Interfaces between storage configurations in the Unit 2 fuel storage pool shall be in compliance with Figures 4.3.1-3, 4.3.1-4, 4.3.1-5, and 4.3.1-6. A assemblies are new or partially spent fuel assemblies with a combination of burnup and initial nominal enrichment in the acceptable burnup domain of Figure 3.7.18-2.

B assemblies are new or partially spent fuel assemblies with a combination of burnup and initial nominal enrichment in the acceptable burnup domain of Figure 4.3.1-8. C assemblies are assemblies with initial enrichments up to a maximum of 5.0 weight percent U-235. L assemblies are new or partially spent fuel assemblies with a combination of burnup, decay time, and initial nominal enrichment in the "acceptable burnup domain" of Figure 4.3.1-10. H assemblies are new or partially spent fuel assemblies with initial nominal enrichments less than or equal to 3.20 weight percent U-235 or which satisfy a minimum IFBA requirement as shown in Figure 4.3.1-9 for higher initial enrichments.

f. A nominal 10.58-inch center to center pitch in the north-south direction and a nominal 10.4-inch center to center pitch in the east-west direction in the Unit 2 high density fuel storage racks.

4.3.1.3 The new fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.05 weight percent;

b. Keff < 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 4.3 of the FSAR;

c. Keff < 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in Section 4.3 of the FSAR; and

d. A nominal 21-inch center to center distance between fuel assemblies placed in the storage racks.

TS 3.7.18, Fuel Assembly Storage in the Fuel Storage Pool, states, in part:

LCO 3.7.18 The combination of initial enrichment burnup and configuration of fuel assemblies stored in the fuel storage pool shall be within the Acceptable Burnup Domain of Figures 3.7.18-1 (Unit 1), 3.7.18-2 (Unit 2), or in accordance with Specification 4.3.1.1 (Unit 1) or 4.3.1.2 (Unit 2).

TS 5.6.5, Core Operating Limits Report (COLR), states, in part, that:

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

LCO 3.1.1 SHUTDOWN MARGIN LCO 3.1.3 Moderator Temperature Coefficient LCO 3.1.5 Shutdown Bank Insertion Limits LCO 3.1.6 Control Bank Insertion Limits LCO 3.2.1 Heat Flux Hot Channel Factor LCO 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor LCO 3.2.3 Axial Flux Difference LCO 3.9.1 Boron Concentration

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:

The following provisions are, in part, the proposed license conditions in the LAR.

Lead test assemblies (LTAs) 7ST1, 7ST2, 7ST3, and 7ST4 contain fuel rods that include advanced coated cladding features and doped or standard fuel material.

Each of the four LTAs may contain up to four fuel rods with a maximum nominal U-235 enrichment of 6.0 weight percent; the maximum nominal U-235 enrichment of the remaining 260 fuel rods must be 5.0 weight percent.

In lieu of the requirements in Technical Specification (TS) Section 4.2, the LTAs are permitted to be placed in limiting core regions for up to two cycles of operation without completion of representative testing.

In lieu of the requirements in TS Section 4.3, the LTAs are subject to the following alternate requirements:

1. These LTAs may be stored in the spent fuel storage racks as specified below:

a. TS 4.3.1.2.b and 4.3.1.2.c must be met.

b. Storage in the Unit 1 and Unit 2 spent fuel storage racks is prohibited except:

i. Unrestricted storage is allowed in the Unit 2 2-out-of-4 checkerboard storage configuration as shown in TS Figure 4.3.1-1.

ii. Storage is allowed in the Unit 2 all-cell storage configuration (A assemblies as shown on TS Figures 4.3.1-3 and 4.3.1-5) when the LTAs reach 64,000 MWd/MTU of burnup.

2. These LTAs may be stored in the new fuel storage racks.

Limiting Condition for Operation (LCO) 3.7.18 shall be considered met for the LTAs provided the alternate Section 4.3 requirements are met.

As previously noted by the NRC staff during the pre-submittal meeting held on May 11, 2022, the proposed license condition for the LTAs, as written, could be read as contradictory to and incompatible with the RFOL license condition C.(2), and subsequently TSs 4.2.1, 4.3.1, 3.7.18, and 5.6.5. Since exceptions applicable to the LTAs are only described in the proposed license condition and not in the applicable TS sections, a reader could interpret the more limiting condition in the TSs to take precedence over the proposed license condition, or vice versa.

Please provide any supplemental information to the application, as appropriate, to explain or resolve this contradiction.

Since the information in the proposed license condition appears to meet the definition of Design Features, please explain how the regulatory requirements of 10 CFR 50.36(c)(4) will be met if the information in the proposed license condition is not incorporated in the TSs.

Question 2 In its letter dated June 30, 2022 (ML22181B156), SNC indicated it is using Westinghouse Topical Reports (TRs) (1) WCAP-18546-P/NP, Westinghouse AXIOM Cladding for Use in Pressurized Water Reactor Fuel (ML21090A110), and (2) WCAP-18482-P/WCAP-18482-NP, Revision 0, Westinghouse Advanced Doped Pellet Technology (ADOPTTM) Fuel (ML20132A014). Confirm whether these TRs are being used to generate operating limits, and if so, how they are addressed in administrative controls.

Question 3 Regulatory Basis General Design Criteria 35 (GDC 35), Emergency core cooling, states, in part, that:

A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

The regulation 10 CFR 50.46, Acceptance criteria for emergency core cooling systems (ECCSs) for light-water nuclear power reactors, requires nuclear power reactors fueled with uranium oxide pellets within cylindrical Zircaloy or ZIRLO cladding to be provided with an ECCS with certain performance requirements and ensures compliance with GDC 35.

The analysis of loss-of-coolant accidents (LOCAs) described in the LAR ensures compliance with 10 CFR 50.46 and GDC 35. Thermal hydraulic (T/H) design methods are used to analyze the outcome of a postulated LOCA.

The LAR makes the following statement:

The T/H design methods for the chromium coated cladding evaluation were reviewed in accordance with the NRC interim guidance. No modification or update to any NRC-approved topical reports on DNB correlations and thermal-hydraulic analysis methods is needed for applications to the LTA coated fuel rods.

The referred-to NRC chromium coated cladding interim staff guidance document, ATF-ISG-2020-01, Supplemental Guidance Regarding the Chromium-Coated Zirconium Alloy Fuel Cladding Accident Tolerant Fuel Concept, Interim Staff Guidance, January 2020 (ML19343A121), contains a discussion of cladding thermal emissivity, and states the following:

Some system codes and accident analysis codes account for cladding surface emissivity and radiation heat transfer from fuel rods to other reactor core components, as well as radiation heat transfer to steam. In general, shinier surfaces have lower emissivity and therefore lower radiative heat transfer. As chromium coatings resist oxidation and retain their surface appearance, it is likely that the coating will negatively impact cladding temperature for transients where radiation to steam is the dominant mode of heat transfer. Therefore, it is likely necessary to revise the outer surface emissivity for accident analyses. This would apply equally to metallic and ceramic coatings (Seshadri, Philips, &

Shirvan, 2018, (Reference 8)) [Seshadri, A., Philips, B., & Shirvan, K. (2018).

Towards Understanding the Effects of Irradiation on Quenching Heat Transfer.

International Journal of Heat Transfer.].

Question 3 In a previously approved LAR for LTAs with coated cladding, thermal emissivity of the cladding was considered during the LOCA analysis, as mentioned in that safety evaluation (ML20363A242).

Please describe how the impact of thermal emissivity of the coated cladding has been dispositioned in the LOCA analysis performed for the licensees proposed LAR for LTAs.

Question 4 Regulatory Basis GDC 5 states, Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.

GDC 62 requires, Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.

The regulation 10 CFR 50.68(a) requires, Each holder of a construction permit or operating license for a nuclear power reactor issued under this part or a combined license for a nuclear power reactor issued under Part 52 of this chapter, shall comply with either 10 CFR 70.24 of this chapter or the requirements in paragraph (b) of this section. The licensee has chosen to comply with 10 CFR 50.68.

The regulation 10 CFR 50.68(b)(1) requires, Plant procedures shall prohibit the handling 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.

The regulation 10 CFR 68(b)(2) requires, 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.

The regulation 50.68(b)(3) requires, 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.

The regulation 10 CFR 50.68(b)(4) requires, 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. If credit is taken for soluble boron, 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 borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.

The Vogtle spent fuel pool (SFP) criticality safety analysis (CSA) does not take credit for soluble boron, so the 10 CFR 50.68(b)(4) requirements regarding soluble boron do not apply.

In addition, 10 CFR 50.36(c)(4), Design features, requires, Design features to be included 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.

Question 4 The licensee is justifying storage of the LTAs in its New Fuel Storage Racks (NFSR) by comparison to its Analysis of Record (AOR). However, SNC did not specify a reference for its NFSR AOR and contrary to Vogtle, Units 1 and 2, TS 4.3.1.3.b and TS 4.3.1.3.c, the NRC staff did not find a discussion of Vogtles NFSR biases and uncertainties in Section 4.3 of the Vogtle FSAR. To allow the NRC staff to evaluate the licensees application against the criteria of 10 CFR 50.68(b)(2) and 10CFR50.68(b)(2), please provide the appropriate reference for Vogtle, Units 1 and 2, AOR for its NFSR.

Question 5 SNCs evaluation of the acceptability to store the LTAs in its NFSR, in its Vogtle, Unit 2, SFP two-out-of-four (2oo4) storage configuration (a repeating 2x2 array of storage cells where fuel assemblies are checkerboarded with empty storage cells), and demonstrating an acceptable response to the identified limiting accident in the SFP are all primarily dependent on the integral fuel burnable absorber (IFBA) loading in the LTAs. However, the IFBA loading is not included in the proposed license condition description of the LTAs. The NRC staffs review will rely on the IFBA loading in the LTAs. Describe the controls SNC will use to ensure that the IFBA loading used in the submittal will be met or exceeded in all respects.

Question 6 The licensee is proposing to allow storage of the depleted LTAs in its Vogtle, Unit 2, SFP All Cell storage configuration, a repeating 2x2 array of storage cells where all fuel assemblies are assumed to have equivalent reactivity. SNC has not performed a CSA for the depleted LTAs in its Vogtle, Unit 2, SFP All Cell storage configuration. Rather, the licensee is extrapolating the AOR for its existing fuel to the depleted LTAs. However, SNC has not established the appropriateness of that extrapolation. In order for the NRC staff to evaluate SNCs application against the criteria of 10 CFR 50.68(b)(4), provide the basis for extrapolating the Vogtle, Unit 2, SFP All Cell AOR to the LTAs. The All Cell CSA in the AOR should be compared to a hypothetical CSA that would otherwise be performed for the depleted LTAs for the Vogtle, Unit 2, SFP All Cell storage configuration, and the effect of any differences should be evaluated in terms of their impact on reactivity. The licensee should describe how it considered the following:

Regulatory Guide 1.240, Fresh and Spent Fuel Pool Criticality, Revision 0, March 2021 (ML20356A127),

NEI 12-16, Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants, Revision 4 (ML19269E069), and

NEI 12-16, Revision 4, Attachment C, Appendix C: Criticality Analysis Checklist.

ML22192A104 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DSS/STSB/BC NAME JLamb KGoldstein VCusumano DATE 7/20/2022 7/21/2022 7/22/2022 OFFICE NRR/DSS/SFNB/BC OGC - NLO NRR/DORL/LPL2-1/BC NAME SKrepel BAyersman SDevlin-Gill for MMarkley DATE 7/25/2022 7/27/2022 7/27/2022 OFFICE NRR/DSS/D NRR/DORL/D NRR/DORL/LPL2-1/PM NAME JDonoghue BPham JLamb DATE 7/28/2022 7/27/2022 8/1/2022