ML23200A262

Issuance of Amendment Nos. 314 and 314, Regarding Technical Specification and Spent Fuel Pool Criticality Change

ML23200A262
Person / Time
Site:	Surry
Issue date:	11/02/2023
From:	John Klos Plant Licensing Branch II
To:	Carr E Virginia Electric & Power Co (VEPCO)
Klos, J
References
EPID L-2022-LLA-0118
Download: ML23200A262 (34)

Text

November 2, 2023 Mr. Eric S. Carr Senior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Blvd.

Glen Allen, VA 23060-6711

SUBJECT:

SURRY POWER STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENT NOS. 314 AND 314, REGARDING TECHNICAL SPECIFICATION AND SPENT FUEL POOL CRITICALITY CHANGE (EPID L-2022-LLA-0118)

Dear Mr. Carr:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 314 to Subsequent Renewed Facility Operating License No. DPR-32 and Amendment No. 314 to Subsequent Renewed Facility Operating License No. DPR-37 for the Surry Power Station, Unit Nos. 1 and 2, respectively. The amendments revise the technical specifications (TS) in response to your application dated August 15, 2022.

These amendments revise TS Section 5.3,1, Criticality, and related figures to accommodate changes to the new fuel storage area and spent fuel pool maximum initial fuel enrichment, storage configurations, and burnup credit.

A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's monthly Federal Register notice. If you have any questions, please contact me at john.klos@nrc.gov, or 301-415-5136.

Sincerely,

/RA/

John Klos, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-280 and 50-281

Enclosures:

1. Amendment No. 314 to DPR-32

2. Amendment No. 314 to DPR-37

3. Safety Evaluation cc: Listserv

VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 314 Subsequent Renewed License No. DPR-32

1. The Nuclear Regulatory Commission (NRC, the Commission) has found that:

A. The application for amendment by Virginia Electric and Power Company (the licensee) dated August 15, 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.

Enclosure 1

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 Subsequent Renewed Facility Operating License No. DPR-32 is hereby amended to read as follows:

(B) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 314, are hereby incorporated in the subsequent renewed facility operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Glenn Glenn E. E. Miller Date: 2023.11.02 Miller 10:11:40 -04'00' Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. DPR-32 and the Technical Specifications Date of Issuance: November 2, 2023

VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 314 Subsequent Renewed License No. DPR-37

1. The Nuclear Regulatory Commission (NRC, the Commission) has found that:

A. The application for amendment by Virginia Electric and Power Company (the licensee) dated August 15, 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.

Enclosure 2

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-37 is hereby amended to read as follows:

(B) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 314, are hereby incorporated in the subsequent renewed facility operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Glenn E. Glenn E. Miller Date: 2023.11.02 Miller 10:12:07 -04'00' Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes License No. DPR-37 and the Technical Specifications Date of Issuance November 2, 2023

ATTACHMENT SURRY POWER STATION, UNIT NOS. 1 AND 2 LICENSE AMENDMENT NO. 314 TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-32 DOCKET NO. 50-280 AND LICENSE AMENDMENT NO. 314 TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NO. 50-281 Replace the following pages of the Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages License License License No. DPR-32, page 3 License No. DPR-32, page 3 License No. DPR-37, page 3 License No. DPR-37, page 3 TSs TSs 5.0-2 5.0-2 Figure 5.3-1 Figure 5.3-1 Figure 5.3-2

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

A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2587 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 314 are hereby incorporated in the subsequent renewed license.

The licensee shall operate the facility in accordance with the Technical Specifications.

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

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

E. Deleted by Amendment 65 F. Deleted by Amendment 71 G. Deleted by Amendment 227 H. Deleted by Amendment 227 Surry - Unit 1 Subsequent Renewed License No. DPR-32 Amendment No. 314

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

A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power Levels not in excess of 2587 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 314 are hereby incorporated in this subsequent renewed license.

The licensee shall operate the facility in accordance with the Technical Specifications.

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

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

E. Deleted by Amendment 54 F. Deleted by Amendment 59 and Amendment 65 G. Deleted by Amendment 227 H. Deleted by Amendment 227 Surry - Unit 2 Subsequent Renewed License No. DPR-37 Amendment No. 314

TS 5.0-2 5.3 FUEL STORAGE 5.3.1 Criticality 5.3.1.1 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 < 0.95 if fully flooded with borated water to 350 ppm, which includes an allowance for uncertainties and biases as described in Appendix 9A of the UFSAR;

c. keff < 1.0 if fully flooded with unborated water, which includes an allowance for uncertainties and biases as described in Appendix 9A of the UFSAR, and;

d. A nominal 14 inch center to center distance between fuel assemblies placed in the storage racks.

5.3.1.2 The new 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. Required empty cells in accordance with Figure 5.3-1, when any stored fuel has an enrichment greater than 4.35 weight percent;

c. keff < 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties and biases calculated in accordance with the methodology described in Appendix 9A of the UFSAR;

d. keff < 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties and biases calculated in accordance with the methodology described in Appendix 9A of the UFSAR; and

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

5.3.1.3 The spent fuel pool is divided into a two-region storage pool. Region 1 comprises the first three rows of fuel racks (324 storage locations) adjacent to the Fuel Building Trolley Load Block. Region 2 comprises the remainder of the fuel racks in the fuel pool. During spent fuel cask handling, Region 1 is limited to storage of spent fuel assemblies which have decayed at least 150 days after discharge and shall be restricted to those assemblies in the acceptable domain of Figure 5.3-2. Administrative controls with written procedures will be employed in the selection and placement of these assemblies.

Amendment Nos. 314 and 314

TS Figure 5.3-1 Figure 5.3-1 NEW FUEL STORAGE RACKS REQUIRED EMPTY CELLS WHEN ANY STORED FUEL IS > 4.35 WT% U-235 Amendment Nos. 314 and 314

TS Figure 5.3-2 Figure 5.3-2 MINIMUM FUEL EXPOSURE VERSUS INITIAL ENRICHMENT TO PREVENT CRITICALITY IN DAMAGED RACKS

() = 849 2 + 21470 33110 where x is the U-235 enrichment in wt%.

Amendment Nos. 314 and 314

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 314 TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-32 AND AMENDMENT NO. 314 TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-280 AND 50-281

1.0 INTRODUCTION

By letter dated August 15, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22227A177), Virginia Electric and Power Company (the licensee) submitted a request for changes to the Surry Power Station, Unit Nos. 1 and 2 (Surry 1&2),

technical specifications (TSs). The requested changes would revise TS Section 5.3.1, Criticality, and related figures to accommodate changes to the new fuel storage area and spent fuel pool (SFP) maximum initial fuel enrichment, storage configurations, and burnup credit.

2.0 REGULATORY EVALUATION

2.1 Proposed Surry 1&2 TS Changes The licensees proposed changes affect the following TS Sections:

• Revise TS 5.3.1.1 on the Spent Fuel Pool Storage Racks

• Revise TS 5.3.1.2 on the New Fuel Storage Racks

• Revise TS 5.3.1.3 on the Two Region Spent Fuel Pool Layout

• Adds new Figure 5.3-1 on the New Fuel Storage Racks Empty Cells

• Revises existing Figure 5.3-1 on the Spent Fuel Pool Region 1 Burnup Curve and renames it Figure 5.3-2 Enclosure 3

Specific TS changes (shown in bold text) are as follows:

o The proposed revision to TS 5.3.1.1 states:

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

a. Fuel assemblies having a maximum U-235 enrichment of 4.3 5.0 weight percent.

b. keff 0.95 if fully flooded with unborated water to 350 ppm, which includes an allowance for uncertainties and biases as described in Appendix 9A of the UFSAR; and

c. keff <1.0 if fully flooded with unborated water, which includes an allowance for uncertainties and biases as described in Appendix 9A of the UFSAR, and; c d. A nominal 14 inch center to center distance between fuel assemblies placed in the storage racks o The proposed revision to TS 5.3.1.2 states:

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

a. Fuel assemblies having a maximum U-235 enrichment of 4.3 5.0 weight percent.

b. Required empty cells in accordance with Figure 5.3-1, when any stored fuel has an enrichment greater than 4.35 weight percent; b c. keff 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties and biases calculated in accordance with the methodology described in Appendix 9A of the UFSAR Virginia Electric and Power Company letter dated November 5, 1997 (Serial No.97-614);

c d. Keff 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties and biases calculated in accordance with the methodology described in Appendix 9A of the UFSAR Virginia Electric and Power Company letter dated November 5, 1997 (Serial No.97-614); and d e. A nominal 21 inch center to center distance between fuel assemblies placed in the storage racks.

o The proposed revision to TS 5.3.1.3 replaces the reference to TS Figure 5.3-1 with a reference to TS Figure 5.3-2 as follows:

The spent fuel pool is divided into a two-region storage pool. Region 1 comprises the first three rows of fuel racks (324 storage locations)

adjacent to the Fuel Building Trolley Load Block. Region 2 comprises the remainder of the fuel racks in the fuel pool. During spent fuel cask handling, Region 1 is limited to storage of spent fuel assemblies which have decayed at least 150 days after discharge and shall be restricted to those assemblies in the acceptable domain of Figure 5.3-1 5.3-2.

Administrative controls with written procedures will be employed in the selection and placement of these assemblies.

o The proposed new TS Figure 5.3.1 is titled NEW FUEL STORAGE RACKS REQUIRED EMPTY CELLS WHEN ANY STORED FUEL IS > 4.35 WT% U-235 and includes an associated diagram identifying which cells must be empty. The Figure also explains which cells can store fuel enrichment to less than or equal to 5 weight-percent.

The proposed revision to current TS Figure 5.3-1 (1) renames it as Figure 5.3-2, (2) does not change the existing title, MINIMUM FUEL EXPOSURE VERSUS INITIAL ENRICHMENT TO PREVENT CRITICALITY IN DAMAGED RACKS, and (3) includes the following additional text with the equation of the burnup curve: Required Burnup (MWD/MTU) = 8492 + 21470 - 33110 where x is the U-235 enrichment in wt%. The licensee also states, The burnup curve is being updated to (i) extend its applicability range out to a U-235 enrichment of 5.0 weight percent and (ii) update it for the methodology described in the attached criticality safety analysis.

2.2 Regulatory Requirements and Guidance The regulatory requirements and guidance documents that the U.S. Nuclear Regulatory Commission (NRC) staff used in the review of the license amendment request (LAR) are listed below.

Title 10 to the Code of Federal Regulations (10 CFR), Part 50, Appendix A, General Design Criteria for Nuclear Power Plants (GDC) Criterion 62, Prevention of criticality in fuel storage and handling, requires that, Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.

The regulations in 10 CFR 50.68, Criticality accident requirements, requires, in part, under 10 CFR 50.68(a), each holder of a power reactor operating license shall comply with either 10 CFR 70.24 or the requirements in 10 CFR 50.68(b). The licensee has elected to comply with 10 CFR 50.68(b), which describes the keff limits for the nuclear criticality safety (NCS) analysis as well as limits on the maximum nominal U-235 enrichment of fresh fuel assemblies to five percent by weight. The applicable requirements of 10 CFR 50.68 are:

10 CFR 50.68(b)(1) states: 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.

10 CFR 50.68(b)(2) states, in part: 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.

10 CFR 50.68(b)(3) states, in part: 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.

10 CFR 50.68(b)(4) states, in part: 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.

10 CFR 50.68(b)(7) states: The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to five (5.0) percent by weight.

The regulations in 10 CFR 50.36, Technical specifications, contain the requirements for the content of TSs. The regulations in 10 CFR 50.36(b) require TSs to be derived from the analyses and evaluations included in the safety analysis report and amendments thereto. The regulation in 10 CFR 50.36(c)(4), Design features, requires that TSs include design 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 10 CFR 50.36.

The NRC staff review was performed consistent with Section 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling (ML070570006), and Section 9.1.2, New and Spent Fuel Storage (ML070550057), of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: [Light-water Reactor] Edition.

The guidance in NEI 12-16, Revision 4, Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants (ML19269E069) provides guidance for acceptable approaches that may be used by industry to perform criticality analyses for the storage of new and spent fuel at light-water reactor power plants in compliance with 10 CFR Part 50. NEI 12-16, Revision 4, was endorsed by the NRC in Regulatory Guide (RG) 1.240, Fresh and Spent Fuel Pool Criticality Analyses (ML20356A127) with some clarifications and exceptions.

3.0 TECHNICAL EVALUATION

3.1 Background

The licensees criticality safety analysis is documented in Attachments 5 and 6 of the application. The intent of the application is to modify the maximum allowable enrichment of fuel stored in the new and spent fuel pool from 4.35 to 5 wt% U-235. In order to accommodate this change, the licensee must also make modifications to the new fuel storage area loading patterns and update the criticality analysis burnup credit for the spent fuel pool.

3.1.1 Applicably Fuel Assembly Designs The NCS analysis must consider all potential fuel assembly designs which may be loaded into, or are currently within, the SFP. The fuel assemblies considered in the licensees NCS analysis include:

15x15 Standard design Westinghouse Upgrade Surry Improved Fuel Framatome Agora LTAs Four 17x17 demonstration assemblies 3.2 Method of Analysis There is no generic or standard NRC-approved methodology for performing NCS analyses for fuel storage and handling. The plant-specific methods used for NCS analyses for the fuel in the Surry new fuel storage area (NFSA) and SFP are described in Section 5, Overview of the Method of Analysis, of the license amendment request (LAR). The licensee used the guidance described in NEI 12-16, which was endorsed by the NRC in RG 1.240, to demonstrate compliance with the requirements of 10 CFR 50.68(b).

3.2.1 Computer Code Validation The licensees NCS analyses consider the decrease in fuel reactivity typically seen in pressurized water reactors (PWR) as the fuel is depleted during reactor operation. This approach is frequently used in PWR NCS analyses and is sometimes referred to as burnup credit. Burnup credit NCS analyses require a two-step process. The first step relates to depletion where a computer code simulates reactor operation to calculate the changes in fuel composition of the fuel assembly. The second step is a modeling of the depleted fuel assembly in the SFP storage racks and the determination of the system keff. The validation of the computer codes in each step is a significant portion of the analysis. Since the licensees NCS analyses credit fuel burnup, the NRC staff must consider validation of the computer codes and data used to calculate burned fuel compositions, and the computer code and data that utilize the burned fuel compositions to calculate keff for the systems with burned fuel.

The licensees analyses employ SCALE 6.2.3 with the ENDF/B-VII 238 group-cross section library along with the SCALE modules CSAS5, TRITON, and ORIGAMI.

3.2.1.1 Criticality Code Validation The licensees NCS analysis utilized CSAS5, a module of SCALE, for criticality calculations.

The code validation for CSAS5 is included in Appendix A, Code Validation for Criticality Analysis Using Laboratory Critical Experiments, of the LAR. The validation set includes critical configurations from the International Handbook of Evaluated Criticality Safety Benchmark Experiments and French Haut Taux de Combustion Critical Experiment Data (ML082880452).

The validation was performed in a manner consistent with NUREG-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology (ML050250061) and included an evaluation for temperature bias. Therefore, the NRC staff concludes that this validation of SCALE 6.2.3 CSAS5 is acceptable for use in this NCS analysis.

3.2.1.2 Depletion Code Validation The licensees depletion analysis utilized two modules of SCALE: TRITON and ORIGAMI. The code validation is included in Appendix B, Fuel Depletion Code Validation for Criticality Analysis Using the EPRI Benchmarks, of the LAR. The validation is performed in a manner consistent with EPRI Reports 3002010613, Benchmarks for Quantifying Fuel Reactivity Depletion Uncertainty - revision 1, and 3002016888, Utilization of the EPRI Depletion Benchmarks for Burnup Credit Validation - Revision 2.

The calculation depletion uncertainty is determined as a function of burnup and compared against the 5% depletion uncertainty described in the NEI 12-16. The depletion uncertainty described in the NEI 12-16 has been previously approved by the NRC numerous times. The licensees calculated depletion uncertainty is similar to the 5% NEI 12-16 depletion uncertainty with slight deviations due to some burnup dependence and application-specific considerations.

Additionally, the licensees method determines depletion uncertainty as an absolute value, instead of as a percentage of worth, as described in NEI 12-16.

At burnups greater than 30 GWD/MTU, the calculated depletion uncertainty slightly falls below 5 percent. The difference between this calculated depletion uncertainty and 5% is negligibly small for all burnups greater than 30 GWD/MTU considered in this analysis. At burnups less than 30 GWD/MTU, the calculated depletion uncertainty is slightly greater than 5 percent. This increase in uncertainty will result in a higher calculated keff which must still be demonstrated to be below the limits described in 10 CFR 50.68.

There is no indication that determining the depletion uncertainty in this manner will result in a non-conservative determination of the maximum keff. The licensees results in Table 6-6, Maximum k-eff with Kopp Uncertainty vs EPRI Benchmarks, show that the keff calculated using the EPRI benchmark methodology is higher than the keff calculated using the Kopp uncertainty methodology for burnups less than 30 GWD/MTU. Because the EPRI benchmark methodology is used in the nominal depletion calculations, no bias associated with the difference of the Kopp uncertainty or EPRI benchmark methodology is necessary Therefore, the NRC staff concludes that this validation of SCALE 6.2.3 with respect to depletion uncertainty is acceptable for use in this NCS analysis.

3.3 New Fuel Storage Criticality Safety Analysis 3.3.1 New Fuel Storage Area Moderator Conditions The NFSA is normally a dry environment. However, under abnormal conditions low-density hydrogenous fluid can be introduced into the storage area, increasing the reactivity of the system. This is referred to as the optimum moderation case and is regulated under 10 CFR 50.68(b)(3). The licensee performed a sensitivity study varying temperature and water density to determine the limiting conditions for optimum moderation for both storage configurations in the NFSA. The licensees method for determining the most limiting conditions is consistent with the guidance in NEI 12-16. The most limiting low-density water at 314 degrees Kelvin will be used throughout the NCS analysis as the nominal case.

The licensees NCS analysis must also consider a case in which the NFSA is fully flooded with unborated water, as required by 10 CFR 50.68(b)(2). In this case, the water has a density of 1 g/cc and a temperature of 314 K. This will result in the determination of the maximum keff for this scenario, as demonstrated by the licensee in Figure 7-5, k-eff of the NFSA vs. Water Density.

Therefore, the NRC staff concludes that the licensees determination of the optimum moderation and fully flooded water temperature and density conditions for the nominal case are acceptable because they will result in the determination of the maximum keff.

3.3.2 Bounding New Fuel Storage Rack The Surry NFSA contains four storage racks of three different sizes. Each of the differently sized racks has the same cell dimensions and materials and contain no neutron absorbers.

Because the storage racks are of the same design, their treatment is the same throughout the entire NCS analysis for the NFSA. A single storage rack design is modeled and its dimensional parameters are varied to maximize the maximum keff, as described in Section 3.3.4 of this SE.

The result is a storage rack model that bounds the storage racks currently in the Surry NFSA.

Therefore, the NRC staff finds the licensees modeling considerations with respect to the new fuel storage racks are acceptable because they will result in a conservative determination of the maximum keff.

3.3.3 Bounding Fresh Fuel Assembly The fuel assemblies considered in this analysis are described in Section 3.1, Fuel Description, of the LAR. A single composite assembly design is used to bound the NFSA reactivity for all historical and anticipated fuel designs. Variations in fuel designs are accounted for in the composite fuel design by selecting the most limiting extremes of each feature, such as material thicknesses and whether or not to include alloying elements. The most limiting fuel assembly features were determined through reactivity comparisons and sensitivity studies. Reduced enrichment and annular pellet axial blankets are not modeled nor credited in this composite fuel design. This approach of using a hybrid set of parameters from multiple assemblies, resulting in a more limiting design basis assembly, is consistent with the guidance in NEI 12-16.

Therefore, the NRC staff concludes that the composite fuel assembly design is appropriately bounded and acceptable for use throughout the NCS analysis for the NFSA because the most limiting parameters are selected that will result in a conservatively high determination of the maximum keff.

3.3.4 Manufacturing Tolerances and Uncertainties The guidance in NEI 12-16 allows for the determination of the keff uncertainty due to manufacturing tolerances and uncertainties to be determined by: (1) a root sum square of the individual keff uncertainty values, (2) all tolerance values selected to maximize keff, or (3) a combination of (1) and (2). The licensee used method (1).

Both the fuel assembly and storage rack have manufacturing tolerances and uncertainties that can affect the reactivity of the system. The parameters related to fuel assembly reactivity are described in Section 7.5, Biases and Uncertainties for the New Fuel Storage Area Analysis, of the LAR. The licensee only calculates the effects of manufacturing tolerances and uncertainties on the nominal case of 4.35 wt% U-235 enriched fuel with no empty cells. The bias calculated from manufacturing tolerances and uncertainties on the nominal 5 wt% U-235 with a 3-out-of-4 arrangement would be less than the 4.35 wt% U-235 4-out-of-4 arrangement because the 3-out-of-4 arrangement has a much lower nominal keff. The impact from manufacturing tolerances and uncertainties is proportional to the nominal keff. The licensee conservatively applies the larger bias from the 4-out-of-4 arrangement to both arrangements.

Therefore, the NRC staff finds that licensees calculation of reactivity effects related to manufacturing tolerances and uncertainties is acceptable because all significant contributors to reactivity are analyzed and their associated biases to the maximum keff are treated appropriately.

3.3.5 Eccentricity of Fuel Within the Storage Cell The nominal keff calculation models all fuel assemblies in the center of their respective storage cells. However, the fuel assemblies can be anywhere within their respective storage cells. The eccentricity portion of the analysis is intended to determine the reactivity effect of the fuel assemblies being in positions other than the center of their storage cell. While the number of locations a fuel assembly could be within its storage cell is numerous, there is no practical difference between most. Per NEI 12-16, the analysis should include the reactivity effect of the most limiting eccentric position, if any, as either a bias from the nominal centrally positioned assembly or as part of the design basis calculation.

Section 7.3.3, Asymmetric Fuel Placement, of the LAR describes the analysis the licensee performed to consider the eccentric positioning of fuel assemblies within a storage cell. The licensees analysis indicates an increase in reactivity resulting from uniformly applied directional shift. The exact direction of the shift is not significant, but a single direction will be used as the nominal shift for simplicity. The NRC staff determined that this approach is acceptable because a directionally uniform shift toward the middle of the NFSA, resulting in an increase to keff, used in the nominal analysis will ensure the determination of the maximum keff.

3.3.6 New Fuel Storage NCS Analysis Biases and Uncertainties The Surry NCS analysis for the NFS appropriately captures all the necessary biases and uncertainties described by the guidance in NEI 12-16. All of the biases and uncertainties will be added to the nominal keff to determine the maximum keff for the NFS. The licensee also allots an additional bias of 0.01 k as review margin for the NRC to use in its technical review to disposition any non-conservatisms or modeling considerations. This value will be taken into consideration for any potential non-conservatisms in the analysis.

Section 7.6, NFSA Maximum k-eff, describes the maximum keff and associated biases and uncertainties. The licensees analysis, consistent with the requirements of 10 CFR 50.68, evaluates the maximum keff for both the optimum moderation case and the fully flooded case.

The licensee considers both of these cases for both loading patterns: 4.35 wt% fuel with no empty cells and 5 wt% fuel in a 3-out-of-4 configuration. The regulatory limit on keff for these cases is 0.98 for optimum moderation and 0.95 for fully flooded. Table 1, below, contains a summary of the results of the licensees analysis.

Table 1: Summary of NFSA NCS Analysis Results Case Optimum Moderation (limit 0.98) Fully Flooded (limit 0.95)

Configuration 4.35% Enriched 5% Enriched 4.35% Enriched 5% Enriched No Empty Cells 3-out-of-4 No Empty Cells 3-out-of-4 Nominal keff 0.9435 0.8568 .8915 .9119 Total Bias and 0.0300 0.0312 .0198 .0198 Uncertainty Maximum keff 0.9735 0.8879 .9112 .9316 Margin 0.0065 0.0921 .0388 .0184 The results summarized in Table 1 indicate that the licensees analysis has demonstrated compliance with the regulations of 10 CFR 50.68(b)(2) and 10 CFR 50.68(b)(3). The licensee evaluated low-density hydrogenous fluid cases and determined that the fully flooded 5 wt%

U-235 3-out-of-4 storage case is bounding. Therefore, 10 CFR 50.68(b)(3) does not apply to the 5 wt% U-235 3-out-of-4 storage configuration as optimum moderation occurs when fully flooded with unborated water. The maximum keff is lower than the regulatory limits and includes some margin for all cases. The total bias and uncertainty value includes the licensees proposed 0.01 k review margin. Therefore, the NRC staff concludes that the licensees NCS for the NFSA is acceptable and meets the requirements of 10 CFR 50.68(b) because the calculated maximum keff is less than the regulatory limit of 0.98 and all necessary biases and uncertainties have been accounted for.

3.4 Depletion Analysis The licensee uses TRITON to model fuel depletion. TRITON is a best estimate code; however, the licensee used bounding parameters to conservatively model the fuel. Uncertainties that are accounted for elsewhere in the NCS, such as manufacturing tolerances and uncertainties, are not included as part of this analysis. The purpose of the depletion analysis is to determine the reactivity effect on fuel in SFP given various depletion characteristics while irradiated in the core. The depletion characteristics that result in the maximum keff of spent fuel will be accounted for in the final calculation of the maximum keff, henceforth referred to as the nominal depletion case. This ensures that the most reactive condition is analyzed.

3.4.1 Core Moderator and Fuel Temperature Higher core moderator and fuel temperatures during in-core depletion typically result in an increase in the nominal SFP keff due to spectral hardening. The licensees analysis indicates that the maximum moderator temperature results in a significant increase in SFP keff, which appears to become more prominent at higher fuel burnups. Increased fuel temperatures appear to have a small positive increase in SFP keff. Therefore, the maximum core moderator and fuel temperatures will be used as the nominal depletion case. The NRC staff determined that the use of the maximum core moderator and fuel temperatures in the nominal depletion case is acceptable because these characteristics ensure that the most reactive condition is analyzed.

3.4.2 Core Soluble Boron Concentration Soluble boron, which is used to reduce reactivity of fuel in the core can have a significant effect on the nominal reactivity of fuel once its moved to the SFP. It is expected that the maximum core soluble boron concentration will result in the maximum SFP keff. The licensees treatment of using a burnup-weighted cycle-averaged soluble boron concentration is consistent with the guidance in NEI 12-16. The licensees analysis indicates that an increase in soluble boron concentration results in a minor increase in SFP keff. Therefore, the maximum core soluble boron concentration is used as the nominal depletion case. The NRC staff determined that the use of the maximum core soluble boron concentration in the nominal depletion case is acceptable because this characteristic ensures that the most reactive condition is analyzed.

3.4.3 Fixed and Integral Burnable Absorbers The use of burnable absorbers can have a significant effect on the reactivity of a spent fuel assembly due to changes in the local neutron flux spectrum. The burnable absorbers that should be considered in the NCS analyses include poison rods, integral absorbers, such as gadolinium, and control rods. The absorbers considered in the licensees NCS analysis include Borosilicate Glass BP and Alumina B4C as poison rod absorbers, and Integral Fuel Burnable Absorbers (IFBA) and Gadolinia as integral absorbers. The poison rod absorbers are discrete rods, referred to as fingers, that are inserted into fuel assembly guide tubes. IFBA is a ZrB2 absorber coating on the surface of fuel pellets. Gadolinia is a neutron absorber that is integrated into the UO2 fuel matrix. The licensee stated that spent fuel at Surry have never included both poison rods and integral absorbers.

The licensees mixed burnable poison analysis for SFP keff indicates that some combination of Alumina B4C and IFBA will be the most limiting burnable poison loading. The burnup credit curves developed by the licensee assumes only 20 fingers of Alumina B4C. Therefore, the allowable burnable absorber combinations are 20 fingers of Alumina B4C and any combinations of absorbers that are less reactive. Table 8-10, Mixed Burnable Poison Loadings During Depletion, indicate that the allowed BP configurations include: (1) up to 12 fingers of Alumina B4C and up to 148 IFBA, or (2) up to 20 fingers of Alumina B4C and no IFBA. Table 8-10 also indicates that a BP loading of 148 IFBA and 20 fingers of Alumina B4C is the most limiting BP loading because it results in a reactivity increase of 0.00262 k over the nominal case.

However, this configuration is not credited by the licensee in the final determination of the maximum keff, and therefore assemblies with this burnable absorber combination are not allowed to be stored in the Surry SFP. The highest reactivity BP loading that is allowed is 20 fingers of Alumina B4C and 0 IFBA. This is used in the nominal depletion case.

The licensee also considered the reactivity effects associated with replacing the 12 Alumina B4C rods from configuration (1) with source rods and stainless steel rods, which results in a small reactivity decrease. The Surry SFP contains assemblies which were irradiated with source rods.

The licensees analysis indicates that these assemblies are bounded by configuration (2).

Therefore, these assemblies may also be stored in the Surry SFP.

The NRC staff has determined that the licensees burnable poison analysis is acceptable because a bounding case is selected and used as the nominal depletion case in future calculations. All assemblies currently stored in the Surry SFP fall within the BP loading combinations described above.

3.4.4 Control Rod Usage If rod cluster control assemblies are present in assemblies for significant amounts of time in the reactor during the operating cycle, the associated spectral hardening can increase plutonium generation, leading to higher fuel reactivity for the same burnup. The licensee performed sensitivity studies to determine the effect that rodded operation had on fuel assemblies in order to determine the bounding nominal case. The licensees analysis does indicate that control rod insertion does result in higher SFP keff. Historically, the Surry reactors operate with the cycle average control rod insertion at 2 steps (3.175 cm). The licensees SFP keff analysis indicates that control rod insertion can be conservatively neglected as long as the cycle average control rod insertion is less than 12 steps (19.05 cm). There is significant margin to the actual Surry cycle average control rod insertion; therefore, it is acceptable to neglect control rod insertion as part of the nominal depletion case.

3.4.5 Axial and Radial Burnup Profile Fuel depletion follows a near-cosine shape along the axial height of the fuel assembly with the highest burnups occurring in the middle of the assembly. The result is that the top and bottom ends of a fuel assembly have a lower burnup and assembly average burnup; therefore, the top and bottom ends are more reactive. As burnup increases the cosine shape flattens. Consistent with the guidance in NEI 12-16, the licensee uses axial burnup distributions from NUREG/CR-6801, Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses, (ML031110292). The licensee compared plant-specific burnup distributions to the selected distributions from NUREG/CR-6801 to demonstrate that the NUREG burnup distributions are conservative. Therefore, the NUREG/CR-6801 axial burnup profiles will be used for the nominal depletion case. The NRC staff determined that the licensees axial and radial burnup evaluation is acceptable because conservative burnup profiles from NUREG/CR-6801 are used.

Assemblies can also experience asymmetric burnup along the radial direction. Certain halves or quadrants of an assembly can be more reactive than their other symmetric half due to control rod insertion and placement in low power core positions. The licensee accounts for the effect of asymmetric planar burnup by introducing an assembly burnup dependent bounding planar tilt effect to half of the design basis assembly. Therefore, the NRC staff finds this approach is acceptable because it will capture any reactivity increasing effects of planar tilt.

3.4.6 Fuel Assembly Power Fuel assembly power is typically closely linked with core moderator and fuel temperature. It is expected that an increase in power will result in a net increase in SFP keff for the assembly. The licensee performed a sensitivity study comparing the SFP keff at burnups of 20 and 40 GWD/MTU with powers at the nominal maximum power and an increase of 5 percent power.

The results indicate that modeling fuel depletion at maximum power will result in a higher keff.

Therefore, the maximum fuel assembly power is used as the nominal case.

Assembly power tends to be lower near end-of-life (EOL) due to placement in a low power location during the last cycle or by power coast down. This results in reduced fuel and moderator temperatures which will reduce SFP keff. However, low EOL power will result in Sm-149 peaking, which will increase SFP keff. The licensee has determined a bounding EOL power coast down to determine a bounding nominal depletion case. The licensees depletion model will deplete at 50 percent power for the last twenty days, allowing Sm-149 to reach

equilibrium levels. The licensees SFP keff analysis indicates that this is a conservative assumption compared to not having an EOL power coast down. Therefore, the EOL power coast down will be included in the nominal depletion case.

Therefore, the NRC staff determined that a maximum nominal fuel assembly power with an EOL power coast down is acceptable for modeling fuel assembly depletion because these conditions will result in the highest calculated SFP keff.

3.4.7 Other Components Any instruments that displace water during the operating cycle harden the neutron spectrum, potentially resulting in an increase in SFP keff. The licensee has chosen to model the in-core detector thimble as part of the nominal depletion case for every assembly even though the thimble can only be inserted into specific assembly locations.

NEI 12-16 provides a conservative method for simplifying the fuel assembly model by neglecting grid volume and including a 50-ppm penalty on soluble boron concentration. However, this is not a requirement, and the licensee seeks to credit grid volume by including some grid material in the nominal model for both depletion and storage. The NRC staff reviewed this deviation from the guidance in NEI 12-16 and determined the method to be an acceptable for demonstrating compliance with 10 CFR 50.68(b) because the reactivity effect of grid material will be accounted for in the keff calculation instead of conservatively neglecting grid material. Any biases and uncertainties associated with this analysis will be included in the calculation of the maximum keff.

The licensees analyses demonstrate that it is conservative to model maximum grid volume during depletion and minimum grid volume during storage. The licensees analyses that credit soluble boron will still account for the 50-ppm penalty described above. The licensee performed a SFP keff sensitivity analysis which included neglecting grid volume in both depletion and storage, which resulted in a positive reactivity increase over the nominal case of 0.00197 k.

Because the licensee does not neglect grid volume, this positive reactivity increase is not accounted for in the final calculation of maximum keff, nor is it required to be. However, the NRC staff has decided to incorporate this positive reactivity increase into the 0.01 k review margin should there be any non-conservatisms in the licensees analysis of grid volume versus SFP keff.

Similar to grid volume, cladding volume has counteractive effects during depletion and SFP storage. More cladding during depletion increases SFP keff due to spectral hardening. However, more cladding material also displaces water and decreases SFP keff. The licensee performed a sensitivity analysis to determine whether modeling with maximum cladding or minimum cladding volume is more conservative. The results in Table 8-22 of the LAR demonstrate that modeling with a minimum cladding volume results in a higher SFP keff. This will be used for the nominal depletion case.

The NRC staff determined that the licensees modeling decisions are acceptable because the credited components are modeled in a conservative manner with respect to SFP keff. Any deviations from NEI 12-16 guidance are analyzed in sensitivity studies to show that limiting parameters are used in the nominal model. The licensees analyses and use of review margin provide the NRC staff with assurance that this NCS analysis demonstrates compliance with 10 CFR 50.68 without following the exact guidance from NEI 12-16.

3.4.8 Depletion Analysis Conclusion The licensees depletion analysis accounts for the many effects that depletion of fuel can have on assembly SFP keff. The nominal depletion model is created to maximize SFP keff using bounding or conservative parameters. A summary of the licensees TRITON depletion model is found in Section 8.15, TRITON Depletion Model Summary, of the LAR which contains all the parameters and modeling decisions determined to be bounding or conservative by the licensee according to the depletion NCS analysis.

Therefore, the NRC staff has determined that the licensees nominal depletion model is acceptable because the model results in a maximum SFP keff that is conservative with respect to all applicable variable parameters in the Surry cores which may impact assembly reactivity.

3.5 Spent Fuel Pool Criticality Safety Analysis The Surry SFP is comprised of two regions. The method of analysis for each region is different.

Region 1 is susceptible to a cask drop accident, which is a bounding accident condition, as demonstrated in Section 12, Accident Analysis, of the LAR. Therefore, the nominal case for the Region 1 analysis will consider a cask drop accident. Additionally, consistent with the double contingency principle, the licensee does not need to consider two independent simultaneous accidents, such as a cask drop during a boron dilution accident, so the Region 1 nominal case will include a soluble boron concentration of 2250 ppm [parts per million by weight]. Because Region 2 is not susceptible to a cask drop accident, the nominal case will not consider any accident conditions and instead include unborated water, which bounds all applicable accident conditions. The NRC staff determined this modeling approach to be acceptable as it considers the most adverse effects on reactivity in the determination of a nominal keff.

3.5.1 Spent Fuel Pool Water Temperature The guidance in NEI 12-16 states that the NCS analysis should use a water temperature and density that results in the maximum reactivity. Typically, the maximum reactivity will occur at either the highest or lowest temperatures allowed. The licensees analysis considers SFP water temperatures at 32 °F, 100 °F, and 170 °F for Region 1 and 32 °F, 68 °F, and 170 °F for Region 2. It was determined that the maximum reactivity occurs at the highest SFP water temperature for both regions, which will be used in the nominal SFP NCS analysis. The NRC staff determined that the licensees evaluation of SFP water temperature is acceptable cause the sensitivity study clearly shows which temperature results in the most reactive conditions and those conditions will be used in the determination of the maximum keff.

3.5.2 Spent Fuel Pool Storage Racks The Surry SFP contains twenty-nine 6x6 storage racks. The racks are all of the same design and materials; therefore, their treatment is the same throughout the entire NCS analysis. The racks do not credit any neutron absorbers. Reactivity is controlled by cell pitch and storage restrictions according to initial fuel enrichment and discharge burnup.

3.5.3 Bounding Spent Fuel Assembly The bounding spent fuel assembly is exactly the same as the composite bounding assembly described in 3.3.3, Bounding Fresh Fuel Assembly, of this SE except with consideration of the depletion characteristics and uncertainties described in Section 3.4, Depletion Analysis of this

LAR. Section 3.5.6 of this SE includes a discussion of five assemblies which are unbounded by the characteristics of the composite bounding assembly. It is shown that despite the unbounded characteristics of those assemblies, the composite bounding assembly still has a higher reactivity, and therefore remains bounding with respect to the determination of the maximum SFP keff. There are no conditions within this LAR that could result in additional unbounded assembles being introduced to the SFP. The NRC staff has determined that the licensees evaluation of the bounding spent fuel assembly is acceptable because the evaluation adheres to the guidance of NEI 12-16 and includes a disposition of any assemblies that are not strictly bounded by the composite design.

3.5.4 Manufacturing Tolerances and Uncertainties The guidance in NEI 12-16 allows for the determination of the keff uncertainty due to manufacturing tolerances and uncertainties to be determined by: (1) a root sum square of the individual keff uncertainty values, (2) all tolerance values selected to maximize keff, or (3) a combination of (1) and (2). The licensee used method (1).

Both the fuel assembly and storage rack have manufacturing tolerance and uncertainties that can affect the reactivity of the system. The parameters related to fuel assembly reactivity are described in Sections 9.1.7 and 9.2.7, both Manufacturing Uncertainties, of the LAR.

The licensees calculation of reactivity effects related to manufacturing tolerances and uncertainties is acceptable because all significant contributors to reactivity are analyzed and their associated biases to the maximum keff are treated appropriately.

3.5.5 Eccentricity of Fuel Within the Storage Cell Eccentric position of fuel assemblies is implicitly considered in Region 1 by considering the different configurations fuel pellets can arrange themselves during a cask drop accident. Thus, the most reactive configuration of fuel is accounted for in Region 1.

Region 2 is not susceptible to a cask drop accident, so eccentric positioning must be considered. Fuel assemblies can be anywhere within their respective storage cells. The eccentricity portion of the analysis is intended to determine the reactivity effect of the fuel assemblies being in positions other than the center of their storage cell. While the number of locations a fuel assembly could be within its storage cell is numerous, there is no practical difference between most. The analysis should include the reactivity effect of the most limiting eccentric position, if any, as either a bias from the nominal centrally positioned assembly or as part of the design basis calculation. The licensee includes eccentric positioning as part of the design basis calculation. The nominal case of eccentric positioning for Region 2 is acceptable because it considers all central assemblies in a 4x4 array eccentrically positioned in the corners of the cells oriented toward the center of a 6x6 model to maximize reactivity.

3.5.6 Non-Standard Fuel The Surry SFP contains four categories of non-standard fuel, or fuel that deviates from normal storage conditions. These categories are fuel rod storage container (FRSC), reconstituted assemblies, demonstration assemblies, and assemblies discharged mid-cycle. The FRSCs and reconstituted assemblies are discussed in Section 3.8.1, Normal Conditions, of this SE.

There are five assemblies that are not bounded by the bounding fuel assembly described in Section 3.5.3, Bounding Spent Fuel Assembly, of this LAR. They include one assembly, 4G7, depleted at a high soluble boron concentration that was not considered in the licensees NCS analysis because the assembly was discharged mid-cycle. The remaining four assemblies are 17x17 demonstration assemblies, which differ from the 15x15 composite bounding spent fuel assembly.

The licensee performed sensitivity studies on the effects of varying depletion boron concentration to determine the reactivity effect on assembly 4G7. While also considering the assemblys initial enrichment and discharge burnup, the licensee demonstrates that there is margin to the maximum SFP keff calculated for standard fuel. This is primarily due to the discharge burnup of assembly 4G7 being in excess of the burnup curve. Additionally, this assembly is currently stored in a dry cask with no intention of ever returning to the SFP.

Regardless, the NRC staff has determined that there is reasonable assurance that storage of this assembly in the Surry SFP will not result in any additional adverse reactivity increases.

The four demonstration assemblies also have discharge burnups in excess of the burnup curve for their respective initial enrichments. The licensee performed a SFP keff model, similar to what was done for the nominal 15x15 model, to show that the 17x17 demonstration assemblies are less reactive than the bounding spent fuel assembly. Additionally, the depletion characteristics of the demonstration assemblies typically result in less reactive spent fuel assemblies.

Therefore, there is reasonable assurance that the 17x17 demonstration assemblies will not result in any adverse increases in reactivity.

3.5.7 Spent Fuel Storage NCS Analysis Biases and Uncertainties The Surry NCS analysis for the SFP appropriately captures all the necessary biases and uncertainties described by the guidance in NEI 12-16. All of the biases and uncertainties will be added to the nominal keff to determine the maximum keff for the SFP. The licensee also allots an additional bias of 0.01 k as review margin for the NRC to use in its technical review to disposition any non-conservatisms or modeling considerations. This value will be taken into consideration for any potential non-conservatisms in the analysis.

Section 9.1.16, Region 1 Damaged Rack Analysis Results and Section 9.2.10, Region 2 Nominal Rack Analysis Results describes the maximum keff and associated biases and uncertainties. The licensees analysis considers the most adverse reactivity conditions for both borated (Region 1) and unborated (Region 2) conditions, consistent with the requirements of 10 CFR 50.68. Table 2, below, contain a summary of the results of the licensees analysis. Note that only the most limiting burnup and enrichment condition is included for the Region 1 analysis.

Table 2: Summary of SFP NCS Analysis Results Region 1 Region 2 Model Damaged Rack, Borated Nominal Rack, Unborated Nominal keff 0.9037 0.9686 Total Bias and Uncertainty 0.0366 0.0219 Maximum keff 0.9403 0.9905 10 CFR 50.68 Limit 0.9500 1.0000 Margin 0.0097 0.0095

The results summarized in Table 2 indicate that the licensees analysis has demonstrated compliance with the regulations of 10 CFR 50.68. The licensees analysis must show that the maximum keff remains below the regulatory limit under two separate conditions stipulated in 10 CFR 50.68(b)(4):

(1) for borated conditions the keff must be less than 0.95, and (2) for unborated conditions the keff must be less than 1.0.

Region 2 is shown to satisfy the requirements of (2) in the table above. The limiting accident scenario for Region 2 is a boron dilution event, which is further described in Section 3.8.3, Boron Dilution, of this SE. It is shown there that the licensee meets the requirements of (1).

The limiting accident scenario for Region 1 is a cask drop accident which is further described in Section 3.8.4, Cask Drop, of this SE. The results of that analysis are noted in the table above, which indicate that the licensee meets the requirements of (1). Region 1 contains fuel that is less reactive than Region 2. Because it is already shown that Region 2 meets the requirements of (2), therefore Region 1 is bounded and also meets the requirements of (2). The double contingency principle precludes the need to analyze two independent and simultaneous accidents, such as a cask drop during a boron dilution accident.

The maximum keff is lower than the regulatory limit and includes some margin for all cases. The total bias and uncertainty value includes the licensees proposed 0.0100 k review margin. The NRC staff concludes that the licensees NCS analysis for the SFP is acceptable and meets the requirements of 10 CFR 50.68(b) because the calculated maximum keff is less than the regulatory limits and all necessary biases and uncertainties have been accounted for.

3.6 Determination of Soluble Boron Requirements The licensees design basis calculations include a soluble boron concentration of 400 ppm and adequately demonstrates compliance with the requirements of 10 CFR 50.68(b)(4) at a boron concentration of 400 ppm under normal conditions. The soluble boron concentration required to mitigate a cask drop accident has been demonstrated to be 2250 ppm.

The licensees minimum SFP boron concentration, as stipulated in TS 5.3.2, Boron Concentration, is 2300 ppm. This is higher than the boron concentration requirements above and no change is being proposed to this limit. Therefore, the licensees current soluble boron concentration requirements are bounding and will ensure the licensee meets the requirements of 10 CFR 50.68.

3.7 Interface Analysis 3.7.1 Interfaces Between Dissimilar Storage Racks Both the Surry NFSA and SFP each contain only one rack design. In the NFSA, the racks vary in size, but all other parameters remain the same. Therefore, there are no interfaces between dissimilar storage racks that may result in an increase in the maximum keff.

3.7.2 Storage Configurations 3.7.2.1 New Fuel Storage Area Storage Configurations The NFSA has two storage configurations depending on the enrichment of fresh fuel. Fuel below 4.35 wt% U-235 enrichment is allowed unrestricted storage in the NFSA. Fuel above 4.35 wt% U-235 and up to 5 wt% U-235 is allowed storage in a 3-out-of-4 arrangement, as depicted in Figure 7-3, Empty Cells Required New Fuel Storage Area, of the LAR. These configurations ensure compliance with the requirements of 10 CFR 50.68(b).

3.7.2.2 Spent Fuel Pool Storage Configurations The Surry SFP contains two storage regions. Region 1 must contain less reactive fuel due to the reactivity impact of a cask drop accident but does not require any empty cells. Region 2, which is unaffected by a cask drop accident, is allowed unrestricted storage of spent fuel assemblies. The storage regions are restricted to certain storage racks. Region 1 constitutes the first three rows of storage racks closest to the cask loading area. Region 2 constitutes the remainder of the storage racks. Assemblies stored in Region 1 must be in the acceptable domain of proposed TS Figure 5.3-2, which describes the burnup and initial enrichment conditions for acceptable store in Region 1 and have 150 days of cooling time.

3.7.3 Interfaces between Different Storage Configurations 3.7.3.1 New Fuel Storage Area Storage Configuration Interfaces The NFSA contains two storage configurations depending on the assembly enrichment. If any fuel in the NFSA is enriched to greater than 4.35 wt% U-235, then the NFSA must be configured consistent with Figure 5.3-1 of the Surry TS. Otherwise, for fuel enriched to less than 4.35 wt%

U-235, the assemblies are allowed unrestricted storage. Because of this TS requirement, there are no interface conditions in the NFSA.

3.7.3.2 Spent Fuel Pool Storage Configuration Interfaces The Surry SFP contains two storage regions. Region 1 is the most reactive storage region. The licensee demonstrates that there is no adverse reactivity effect associated with the interface between Region 1 and Region 2. The reactivity of the interface is dominated by the Region 1 reactivity and there is no reactivity increase associated with the interface. Therefore, the interface analysis is bounded by the Region 1 analysis.

3.7.4 Interface Analysis Conclusion Therefore, the NRC staff has determined that licensees interface analysis is acceptable because all interfaces are analyzed, and it is demonstrated that the interfaces do not result in any adverse reactivity trends.

3.8 Normal and Accident Conditions 3.8.1 Normal Conditions In addition to routine storage of fuel assemblies whose reactivity is bounded by the design basis assembly for this criticality analysis, the licensee also considers the following activities to be regular:

Use of fuel inserts, Fuel handling and movement, Fuel Inspection, Storage of non-standard fuel, and Storage of non-fuel items.

Fuel inserts are control rods, discrete burnable poisons, flux suppression assemblies, source rods, and other items that may be inserted into the guide tubes of spent fuel assemblies. These inserts strictly reduce the reactivity of the assembly. Therefore, there is no adverse effect on reactivity resulting from the storage of fuel assemblies with inserts.

The licensee has placed controls on fuel handling and movement. These controls, among others, include limiting the fuel transfer canal to one assembly at a time and only moving assemblies into procedure-approved locations. It is possible for the licensee to move two fuel assemblies at the same time; therefore, the licensee must consider a condition in which two assemblies are near each other without the geometric constraints of the spent fuel rack. It is generally accepted, and the licensees analysis confirms, that 12 inches of water between two assemblies is sufficient to neutronically decouple the assemblies. Therefore, Surrys update to the fuel handling procedures of maintaining 12 inches of separation between fuel assemblies when outside the storage racks is sufficient to preclude an unanalyzed increase in reactivity and possibly a violation of 10 CFR 50.68.

Fuel inspection is necessary to monitor the stored spent fuel assemblies. This inspection can take place when the fuel assembly is inside or outside of the storage rack. When outside of the storage rack, the licensee will maintain the minimum separation distance as described above while performing the inspection. Inspection of fuel assemblies stored in the storage rack will have no adverse effects on reactivity control.

The Surry SFP contains two forms of non-standard fuel: FRSC and reconstituted fuel. FRSCs store failed fuel rods that have been removed from standard fuel assemblies. The FRSCs when fully loaded with fuel rods contain fewer fuel rods than the bounding assembly design making it is less reactive. The licensee notes that the FRSC is only bounded when stored in a Region 2 cell. Therefore, the FRSC may only be stored in a Region 2 cell. The Surry SFP also includes reconstituted fuel, which are assemblies with fuel pins that have been removed and possibly replaced. The removed fuel pin can be replaced with an insert pin or nothing at all. Not replacing a removed fuel pin with an insert pin can result in a reactivity increase by increasing moderation.

The licensee compensates for this by limiting reconstituted fuel assemblies with open holes to storage in Region 2, as long as a dry storage cask is present in the pool. This storage limitation is sufficient to bound any reactivity increase associated with leaving open holes in reconstituted fuel.

Some storage cells in the Surry SFP function as trash cans to store non-fuel items. The non-fuel items only displace water, reducing moderation, and thus reducing reactivity. As there are no empty cells required in the Surry SFP, there is no restriction necessary with regard to the trash can cells.

Therefore, the NRC staff has determined that the licensees analysis of normal conditions is acceptable because any adverse reactivity effects are accounted for by placing administrative controls on fuel storage and handling to preclude any reactivity increases.

3.8.2 Bounding Accident The licensee considered several accident conditions that could potentially result in an increase in keff. The conditions analyzed are:

Boron dilution, Single fuel assembly misload, Multiple fuel assembly misload, Loss of SFP cooling, Dropped fuel assembly, Misplacement of a fuel assembly outside the storage rack, Seismic event, and Cask drop The licensees analyses demonstrate that the cask drop accident is the bounding accident condition and is used as the nominal case for Region 1 in the SFP NCS analysis described in Section 3.5 of this SE. The licensee uses a nominal soluble boron concentration of 350 ppm in the NCS analysis for Region 2 of the SFP. This bounds all other accident conditions.

The NRC staff has determined that the licensees disposition of accident conditions is acceptable because bounding credible accident conditions are used in the nominal cases for each region of the SFP.

3.8.3 Boron Dilution A boron dilution accident is one in which there is a leak of borated water from the SFP. The water added to the SFP that might compensate for such a leak may not be borated, thus reducing the soluble boron concentration to below the TS minimum concentration of 2300 ppm.

To accommodate a boron dilution accident, the licensee must maintain a minimum boron concentration of 350 ppm, as determined by the minimum boron concentration necessary for normal conditions.

The licensees analysis must demonstrate that there is adequate time to detect and mitigate a dilution event before keff exceeds 0.95. The analysis considers a dilution event starting at the TS minimum soluble boron concentration of 2300 ppm diluting to 400 ppm. The licensees analysis demonstrates that the maximum time to identify a boron dilution event may be up to 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />, which is significantly less than the 41 hours4.74537e-4 days <br />0.0114 hours <br />6.779101e-5 weeks <br />1.56005e-5 months <br /> necessary to dilute the SFP to a soluble boron concentration of 400 ppm under bounding circumstances. Therefore, it is demonstrated that a boron dilution accident that dilutes the soluble boron concentration to below 400 ppm is not credible. The NRC staff has determined that the licensees boron dilution accident analysis is acceptable because there is significant margin between the operator response and mitigate time and the minimum dilution to 400 ppm boron time.

3.8.4 Cask Drop A cask drop accident is the bounding accident condition for Region 1 of the Surry SFP. The application does not attempt to accurately predict and model the damage that a dropped cask will cause, but instead comes up with three bounding conditions. These conditions are (1) all fuel stays within the storage cells; (2) the fuel falls into ordered stacks beneath the storage rack; (3) the fuel falls between cell walls. In all of these cases, the fuel stack or pin pitch is optimized to maximize reactivity. One of these cases is bounding for each burnup and enrichment combination and is used in the nominal model for that specific case for Region 1.

When analyzing the cask drop accident, the licensees model uses a soluble boron concentration of 2250 ppm. There is no need to consider unborated water due to the double contingency principle which does not require the licensee to model multiple simultaneous independent accidents.

Therefore, the NRC staff has determined that the licensees cask drop accident analysis is acceptable because bounding conditions for each burnup and enrichment combination are used in the nominal SFP keff calculation for Region 1 of the SFP.

3.9 NCS Analysis Results An acceptable NFSA and SFP NCS analysis must demonstrate compliance with the requirements of 10 CFR 50.68(b). The licensee has demonstrated in the NCS analysis that the maximum keff will remain below the regulatory limits. The results of the NFSA and NCS analysis are summarized in Table 3, below.

Table 3 shows that the licensee has demonstrated compliance with the requirements of 10 CFR 50.68(b) in all cases. These cases include:

(1) The keff of the NFSA must remain below 0.95 when fully flooded with unborated water (10 CFR 50.68(b)(2)).

(2) The keff of the NFSA must remain below 0.98 when filled with low-density hydrogenous fluid if optimum moderation occurs when filled with low-density hydrogenous fluid.

(10 CFR 50.68(b)(3)).

(3) The keff of the SFP must remain below 0.95 when fully flooded with borated water with consideration of a limiting accident scenario (10 CFR 50.68(b)(4)).

(4) The keff of the SFP must remain below 1.0 when fully flooded with unborated water (10 CFR 50.68(b)(4)).

Section 3.3.6, New Fuel Storage Area NCS Analysis Biases and Uncertainties, of this SE describes how the licensee meets the requirements of 10 CFR 50.68(b)(2) and 50.68(b)(3).

Section 3.5.7, Spent Fuel Storage NCS Analysis Biases and Uncertainties, of this SE describes how the licensee meets the requirements of 10 CFR 50.68(b)(4).

The licensees analyses consider all significant reactivity increasing effects. Therefore, there are no known adverse conditions that have not been analyzed which could result in a violation of

these requirements. The licensee provided the NRC staff with up to 0.01 k review margin to account for potential non-conservatisms in the NCS analysis. Overall, the NRC staff accounted for 0.00197 k toward the review margin due to grid modeling. However, the full 0.01 k is accounted for in Table 3. The NRC staff recognizes that the licensees modeling decisions related to grid modeling are realistic and not non-conservative but decided to make use of the review margin anyway.

Therefore, the NRC staff has determined that the licensees NCS analysis is acceptable because the requirements of 10 CFR 50.68(b) are met, as summarized in Table 3, below.

Table 3: Summary of NFSA and SFP Maximum keff and Regulatory Limits New Fuel Storage Spent Fuel Storage Storage 4.35 wt% U-235 5 wt% U-235 Region 1 Region 2 Pattern 4-out-of-4 3-out-of-4 Optimum Fully Optimum Fully Condition Cask Drop Unborated Moderation Flooded Moderation Flooded Maximum 0.9735 0.9112 0.8879 0.9316 0.9403 0.9905 keff 10 CFR 0.9800 0.9500 0.9800 0.9500 0.9500 1.000 50.68 Limit Margin to 0.0065 0.0388 0.0921 0.0184 0.0097 0.0095 50.68 Limit 3.10 TECHNICAL CONCLUSION The NRC staff concludes that there is reasonable assurance that the Surry NFSA and SFP meet the applicable regulatory requirements in 10 CFR 50.68 and GDC 62.

Additionally, the NRC staff determined that the proposed TSs would continue to be based on the analyses and evaluations included in the UFSAR and amendments thereto in accordance with 10 CFR 50.36(b). The NRC staff also determined that the proposed TSs will continue to include required design features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety, in accordance with 10 CFR 50.36(c)(4).

Therefore, the NRC staff concludes that the proposed changes are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Commonwealth of Virginia official was notified of the proposed issuance of the amendments on July 19, 2023. On July 19, 2023, the State official confirmed that the Commonwealth of Virginia had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the 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 published in the Federal Register on November 8, 2022, 87 FR 67506, and there has been no public comment on such finding. 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 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 Contributor: Brandon Wise Date: November 2, 2023

ML23200A262 *by email OFFICE DORL/LPL2-1/PM DORL/LPL2-1/LA DSS/SFNB/BC* DSS/STSB/BC*

NAME JKlos KGoldstein SKrepel VCusumano DATE 7/18/2023 09/27/2023 5/2/2023 08/09/2023 OFFICE OGC* DORL/LPL2-1/(A) BC DORL/LPL2-1/PM NAME MASpencer MMarkley (EMiller for) JKlos DATE 9/13/2023 11/2/2023 11/2/2023