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Issuance of Amendment No. 212 to Revise TS 3.7.15 & 4.3, to Support Updated SFP & New Fuel Criticality Analyses
ML25241A191
Person / Time
Site: Saint Lucie NextEra Energy icon.png
Issue date: 09/26/2025
From: Natreon Jordan
NRC/NRR/DORL/LPL2-2
To: Coffey R
Florida Power & Light Co
Jordan N, NRR/DORL/LPL2-2
References
EPID L-2024-LLA-0057
Download: ML25241A191 (1)
v • d • e


Text

September 26, 2025 Robert Coffey Executive Vice President, Nuclear and Chief Nuclear Officer`

Florida Power & Light Company 700 Universe Blvd.

Mail Stop: EX/JB Juno Beach, FL 33408

SUBJECT:

ST. LUCIE PLANT, UNIT NO. 2 - ISSUANCE OF AMENDMENT NO. 212 TO REVISE TECHNICAL SPECIFICATION 3.7.15, SPENT FUEL POOL STORAGE, AND 4.3, FUEL STORAGE, TO SUPPORT UPDATED SPENT FUEL POOL AND NEW FUEL CRITICALITY ANALYSES FOR PROPOSED TRANSITION TO 24-MONTH FUEL CYCLES (EPID L-2024-LLA-0057)

Dear Mr. Coffey:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 212 to Renewed Facility Operating License No. NPF-16 for the St. Lucie Plant (St. Lucie), Unit No. 2. The amendment is in response to your application dated April 30, 2024, as supplemented by letter dated July 29, 2025.

The amendment revises St. Lucie, Unit No. 2, Technical Specification (TS) 3.7.15, Spent Fuel Pool Storage, and TS 4.3, Fuel Storage, to support updated spent fuel pool and new fuel vault criticality analyses, which account for the impact of a proposed transition to 24-month fuel cycles on fresh and spent fuel storage at St. Lucie Plant, Unit No. 2.

A copy of the Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Natreon Jordan, Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-389

Enclosures:

1. Amendment No. 212 to NPF-16
2. Safety Evaluation cc: Listserv

FLORIDA POWER AND LIGHT COMPANY ORLANDO UTILITIES COMMISSION OF THE CITY OF ORLANDO, FLORIDA AND FLORIDA MUNICIPAL POWER AGENCY DOCKET NO. 50-389 ST. LUCIE PLANT, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 212 Renewed License No. NPF-16

1.

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

A.

The application for amendment by Florida Power & Light Company, et al. (FPL, the licensee), dated April 30, 2024, as supplemented on July 29, 2025, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

Enclosure 1 2.

Accordingly, Renewed Facility Operating License No. NPF-16 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and by amending paragraph 3.B, in part, to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 212, are hereby incorporated in the renewed license. FPL 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 180 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION David Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: September 26, 2025 DAVID WRONA Digitally signed by DAVID WRONA Date: 2025.09.26 13:14:23 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 212 ST. LUCIE PLANT, UNIT NO. 2 RENEWED FACILITY OPERATING LICENSE NO. NPF-16 DOCKET NO. 50-389 Replace the following page of Renewed Facility Operating License No. NPF-16 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert Page 3 Page 3 Replace the following pages of the Appendix A, Technical Specifications, with the attached pages. The revised pages are identified by amendment number and contain a marginal line indicating the areas of change.

Remove Insert 3.7.14-1 3.7.14-1 3.7.15-1 3.7.15-1 3.7.15-2 3.7.15-2 3.7.15-3 3.7.15-3 3.7.15-4 3.7.15-4 3.7.15-5 3.7.15-5 3.7.15-6 3.7.15-6 3.7.15-7 3.7.15-7 3.7.15-8 3.7.15-8 3.7.15-9 3.7.15-9 3.7.15-10 3.7.15-10 3.7.15-11 3.7.15-12 3.7.15-13 4.0.1 4.0.1 4.0.2 4.0.2 Renewed License No. NPF-16 Amendment No. 212 neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required.

D.

Pursuant to the Act and 10 CFR Parts 30, 40, and 70, FPL to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and E.

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

3.

This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission's regulations: 10 CFR Part 20, Section 30.34 of 10 FR Part 30, Section 40.41 of 10 CFR Part 40, Section 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 to 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 FPL is authorized to operate the facility at steady state reactor core power levels not in excess of 3020 megawatts (thermal).

B.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 212, are hereby incorporated in the renewed license. FPL shall operate the facility in accordance with the Technical Specifications.

Appendix B, the Environmental Protection Plan (Non-Radiological), contains environmental conditions of the renewed license. If significant detrimental effects or evidence of irreversible damage are detected by the monitoring programs required by Appendix B of this license, FPL will provide the Commission with an analysis of the problem and plan of action to be taken subject to Commission approval to eliminate or significantly reduce the detrimental effects or damage.

C.

DELETED D.

Antitrust Conditions FPL shall comply with the antitrust conditions in Appendices C and D to this renewed license.

E.

Fire Protection Florida Power & Light Company (FPL) St. Lucie Plant Unit 2 shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated March 22, 2013, and May 2, 2017, and supplements dated June 14, 2013, February 24, 2014, March 25, 2014, April 25, 2014, July 14,

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 212 TO RENEWED FACILITY OPERATING RENEWED FACILITY OPERATING LICENSE NO. NPF-16 FLORIDA POWER & LIGHT COMPANY (FPL)

ST. LUCIE PLANT, UNIT NO.1 DOCKET NO. 50-389

1.0 INTRODUCTION

By letter dated April 30, 2024, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24122A689) Florida Power and Light Company (FPL) (the licensee) submitted a license amendment request (LAR) to change the St. Lucie Plant, Unit No. 2 (SL2)

Technical Specifications (TS) 3.7.15, Spent Fuel Pool Storage, and TS 4.3, Fuel Storage, to support updated spent fuel pool (SFP) and new fuel vault (NFV) criticality analyses which account for the impact of a proposed transition to 24-month fuel cycles on fresh and spent fuel storage at SL2. By letter dated July 29, 2025 (ML25211A263), FPL submitted updates to the initial LAR. The letter dated July 29, 2025, contained the licensees responses to the NRC staffs request for additional information (RAI), revised proposed TS changes, and provided proprietary and non-proprietary versions of revised Holtec report Hl-2230346, Revision 2, Criticality Safety Analysis of SFP for SL2.

The supplement dated July 29, 2025, requested changes to the proposed TS in their entirety, The specific changes are detailed in Enclosure 1 to the July 29, 2025, supplement.

The supplement dated July 29, 2025, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or the Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on July 9, 2024 (89 FR 56438).

2.0 REGULATORY EVALUATION

In accordance with the licensees amendment request, the regulatory requirements and guidance that the NRC staff considered in assessing the proposed TS changes are as follows:

The regulations in Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix A, 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.

Per 10 CFR 50.68(a), each holder of an operating license shall comply with either 10 CFR 70.24 or the requirements in 10 CFR 50.68(b). The licensee has elected to meet 10 CFR 50.68(b).

Accordingly, and as relevant to this license amendment request, the licensee must comply with the following 50.68(b) requirements:

(1) 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.

(2) 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.

(3) 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.

(4) 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 categories of items required to be in the TSs are provided in 10 CFR 50.36(c). As required by 10 CFR 50.36(c)(4), the TSs will include design features which 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 10 CFR 50.36.

Per 10 CFR 50.9(a), Information provided to the Commission by an applicant for a license or by a licensee or information required by statute or by the Commissions regulations, orders, or license conditions to be maintained by the applicant or the licensee shall be complete and accurate in all material respects.

3.0 TECHNICAL EVALUATION

3.1 Method of Review There is no generic or standard NRC-approved methodology for performing nuclear criticality safety (NCS) analyses for fuel storage and handling at commercial power reactors. Each analysis is unique to its analyzed system.

The LAR specifies that the analysis was performed following the guidance in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants Sections 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling (ML070570006), as well as Regulatory Guide (RG) 1.240, Revision 0, Fresh and Spent Fuel Pool Criticality Analyses (ML20356A127). RG 1.240 endorses Nuclear Energy Institute (NEI) guidance document NEI 12-16, Guidance for Performing Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants, Revision 4 (ML19269E069), with clarifications and exceptions.

This safety evaluation (SE) involves a review of the licensees NCS analysis for the SL2 NFV and SFP. A summary of the licensees NCS analyses for the SFP was provided in Enclosures of the licensees July 29, 2025, letter. A summary of the licensees NCS analyses for the NFV was provided in the letter dated April 30, 2024, as Enclosures 3 and 6, Holtec report HI-2230455, Revision 2, Criticality Safety Analysis of NFV for St. Lucie Unit 2 (Proprietary/Non-Proprietary).

3.2 The SL2 SFP NCS Analysis Review 3.2.1 SFP Overview The analysis supporting the applicants request for the modification to its SFP storage controls is contained in the letter dated July 29, 2025, as Enclosures 2 and 3, Holtec report HI-2230346, Revision 2, Criticality Safety Analysis of SFP for St. Lucie Unit 2 (Proprietary/Non-Proprietary).

The SL2 SFP has three rack designs. Region 1 (R1) and Region 2 (R2) are identical except R1 has permanently installed stainless steel inserts. The third region is the cask pit rack (CPR) which is different in several aspects including the permanently installed neutron absorbing material (NAM) Boral'.

The proposal by the licensee is for the SL2 SFP to have a total of 11 distinct storage configurations, each identified by repeating 2x2 arrays of storage cells. R1 is to have two storage configurations. R2 is to have four storage configurations. The CPR is to have five storage configurations. In addition, the proposal includes another five temporary storage configurations that would only be used during Fuel Inspection and Reconstitution (see Figure 3-7 of HI-2230346, Rev 2.).

3.2.2 Computational Methods The depletion calculations for the spent fuel storage racks were performed using CASMO5 Version 2.08.00 (CASMO5), which is a multigroup two-dimensional transport theory code for burnup calculations of fuel assemblies. The code allows modeling of a planar cross section of an individual fuel assembly, including individual pellet and cladding diameters, guide and instrument tube locations, and material compositions. The calculations assume a planar and axially infinite array of fuel assemblies to provide the actinide and fission product densities in the isotopic composition of spent fuel.

For the spent fuel rack and NFV criticality calculations, MCNP5 Version 1.51, was used, which is a Monte Carlo code developed by Los Alamos National Laboratory that relies on repeated random sampling to determine neutron lifecycles. The code offers three-dimensional calculations of the storage racks based on ENDF/B-VII cross-sections. As described in of the licensees letter dated July 29, 2025, the MCNP5 code computes a Shannon entropy of the fission source distribution to account for the convergence of the fission source spatial distribution.

3.2.2.1 Depletion Computer Code Validation The LAR specifies that the analysis of fuel irradiation during core operation two-dimensional transport theory code is based on the Method of Characteristics, using the ENDF/B-VII Library.

In its letter dated July 29, 2025, response to RAI #4, the licensee confirmed that its use of CASMO5 in this analysis is consistent with topical report SSP-14-P01/028-TR-P-A, Generic Application of the Studsvik Scandpower Core Management System to Pressurized Water Reactors, Revision 0 (Proprietary; a Non-Proprietary version is available at ML17279A986).

The guidance in NEI 12-16, Rev. 4, section 4.2.3, PWR Depletion Bias and Uncertainty, states, in part:

In lieu of a formal lattice depletion validation, the licensee may apply an uncertainty equal to 5 [percent] of the reactivity decrement, if the licensee uses the lattice depletion code in a manner that is consistent with nuclear design calculations previously performed for commercial power reactor licensing. This ensures that the depletion code will produce reliable and predictable results for the intended application.

Using CASMO5 consistent with topical report SSP-14-P01/028-TR-P-A provides reasonable assurance it was used in a manner that is consistent with nuclear design calculations previously performed for commercial power reactor licensing. Therefore, use of 5 percent of the delta k between fresh and burnup of interest as the depletion uncertainty is acceptable.

3.2.2.2 SFP keff Computer Code Validation For the spent fuel rack and NFV criticality calculations, MCNP5 Version 1.51 was used. The code is capable of performing three-dimensional calculations of the storage racks based on ENDF/B-VII cross-sections. The NRC staff has accepted properly validated usage of MCNP for performing NFV and SFP nuclear criticality safety analyses.

Holtec report HI-2230346, Revision 2, section 3.2.2.1, MCNP Validation, describes the validation of MCNP Version 1.51 and ENDF/B-VII cross-sections used in the analysis. The applicant states the validation was performed in accordance with NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology (ML050250061). Using NUREG/CR-6698 is consistent with the guidance in NEI 12-16 Rev. 4 and RG 1.240, Rev. 0.

Per Holtec International HI-2230346, Revision 2, the details of the validation are contained in Nuclear Group Computer Code Benchmark Calculations, HI-2104790, Revision 3.

HI-2104790, Revision 3 was not submitted as part of the application. It is not currently on the docket, nor was it part of the audit conducted during the review. Therefore, the NRC staff relied upon the summary information provided in HI-2230346, Revision 2, Tables 3-1, Summary of Area of Applicability of the MCNP Benchmark, Table 3-2, MCNP Benchmark Analysis for Various Fuel and Water Subsets of Experiments, Table 3-3, Significant Trending Analysis for St. Lucie Unit 2 Parameters, Table 3-4, Summary of MCNP Code Validation Bias and Bias Uncertainty, and 10 CFR 50.9(a). The information provides reasonable assurance that MCNP5 Version 1.51, with the ENDF/B-VII cross-sections, was validated appropriately for the SL2 SFP.

3.2.3 SFP and Fuel Storage Racks 3.2.3.1 SFP Water Temperature Guidance in NEI 12-16, Rev. 4 and RG 1.240, Rev. 0, specifies that the NCS analysis should be done at the temperature corresponding to the highest reactivity. Holtec report HI-2230346, Revision 1, sections 3.3.7, Spent Fuel Pool Water Temperature, and 8.7, Reactivity Effect of SFP Water Temperature, discuss the applicants sensitivity analyses on SFP water temperature. It appears the applicant found the most reactive temperature and either applied a bias or used the most reactive temperature as the temperature in the analysis. Therefore, the staff finds that the water temperature was handled appropriately in the licensees criticality analysis.

3.2.3.2 SFP Storage Rack Models Holtec report HI-2230346, Revision 2, section 3.3, Analysis Methods, summarizes the SFP storage rack models.

3.2.3.3 SFP Storage Rack Models Manufacturing Tolerances and Uncertainties For the Region 1 SFP rack, the licensee included uncertainties for the following SFP rack manufacturing tolerances: storage cell inner diameter (ID), storage cell wall thickness, storage cell pitch, stainless steel L-insert thickness, width, and gap.

For the Region 2 SFP rack, the licensee included uncertainties for the following SFP rack manufacturing tolerances: storage cell ID, storage cell wall thickness, and storage cell pitch.

For the CPR rack, the licensee included uncertainties for the following SFP rack manufacturing tolerances: storage cell ID, storage cell wall thickness, storage cell pitch, neutron absorbing material thickness, width, pocket thickness, and sheathing thickness.

In its July 29, 2025, letter response to RAI #5, the licensee confirmed that the minimum certified 10B Areal Density was used for the BORAL' and Metamic' neutron absorbing materials (NAMs). With respect to the Control Element Assembly (CEA)-type inserts, the applicant used a conservative 10B Areal Density.

The NRC staff concludes that the licensees treatment of SFP storage rack manufacturing tolerance and uncertainties is generally consistent with the guidance in NEI 12-16, Rev. 4 and RG 1.240, Rev. 0, and is, therefore, acceptable.

3.2.4 Eccentric Positioning Holtec report HI-2230346, Revision 1, section 3.3.8.1, Fuel Assembly Radial Positioning, and section 8.8, Reactivity Effect of Fuel Assembly Radial Positioning, describe the applicants initial sensitivity analyses on fuel assembly positioning within SFP storage cells. HI-2230346, Revision 1, considered three eccentric position scenarios for each of the 11 distinct storage configurations. Those were a 2x2 array of all fuel assemblies centered in their respective storage cells, a 2x2 array with all fuel assemblies in the innermost corner of their respective storage cells, and a 2x2 array with all fuel assemblies in the same relative (unidentified) corner.

Additionally, an 8x8 array with all fuel assemblies in the innermost corner of their respective storage cells was also analyzed. In all cases the reactivity of the 8x8 array was significantly higher than the smaller 2x2 array version. The guidance in RG 1.240 and NEI 12-16 calls for eccentric positioning to be analyzed with a 4x4 array of storage cells. That is because in some scenarios, the boundary conditions on smaller models may mask reactivity increases, as demonstrated in the cases with all fuel assemblies positioned in the innermost corner. In the licensees letter dated July 29, 2025, the licensee analyzed additional scenarios with fuel assemblies in two corners, the southwest and northeast corners, and in a non-symmetric configuration. All the new analyses were performed in 8x8 arrays. Revisions to the eccentric positioning analyses are described in the response to RAI # 1 and incorporated into Holtec report HI-2230346, Revision 2.

The CPR has NAM Boral' permanently installed with an areal density of 0.030 B10 gm/cm2.

With respect to the CPR, the applicants sensitivity study indicates the most reactive position is with the fuel assemblies centered in their respective storage cells. The model with the fuel assemblies centered in their respective storage cells was the base condition, no bias was used.

This is consistent with the guidance in NEI 12-16, Rev. 4, and RG 1.240, Rev. 0, for strong NAM and is acceptable to the NRC staff.

The R1 rack has permanently installed stainless steel inserts as a reactivity control device. The NRC staff consider stainless steel to be a weak NAM. R1 has two storage configurations; 2F which is a repeating 2x2 array checkerboard pattern of fresh and empty cells, and 3E which is a repeating 2x2 array with three irradiated fuel assemblies and one empty cell. With respect to the R1 2F storage configuration, the applicants sensitivity study indicates it has an eccentric positioning bias of 0.0076 k. With respect to the R1 3E storage configuration, the applicants sensitivity study indicates it has an eccentric positioning bias of 0.0015 k. While the scenarios analyzed still represent a small fraction of those possible, the NRC staff considered the margin available in the R1 storage configurations and accepted the applicants R1 eccentric modeling.

The R2 rack does not have any permanently installed NAM. R2 has four storage configurations; 1C which is a repeating 2x2 array of four irradiated fuel assemblies with one having an insert, 2C which is a repeating 2x2 array of irradiated fuel assemblies with two having an insert, 3C which is a repeating 2x2 array of irradiated fuel assemblies with three having an insert, and 3E which is a repeating of 2x2 array with three irradiated fuel assemblies and one empty cell. The insert may be either a control element assembly, or a Dream insert. The applicants sensitivity analysis on fuel assembly positioning within R2 found eccentric position scenarios that were more reactive than the nominal all fuel assemblies centered in their respective storage cells, but rather than calculate and apply a bias, the applicant then used the more reactive eccentric position scenarios as the reference positioning. While the scenarios analyzed still represent a small fraction of those possible, the NRC staff considered the margin available in the R1 storage configurations and accepted the applicants R2 eccentric modeling.

3.2.5 Fuel Assembly 3.2.5.1 Fuel Assembly Design HI-2230346, Revision 2, section 3.3.1, Design Basis Fuel Assembly Design, describes the selection of the fuel assembly design used in the analysis.

3.2.5.1.1 Fuel Assembly Physical Changes with Depletion HI-2230346, Revision 2, section 3.3.9.5, Reactivity Effect of Depletion Related Fuel Assembly Geometry Changes, describes how the physical changes of the fuel assembly are modeled.

The licensee included a fuel depletion related geometry change bias determined in a way consistent with NEI 12-16, Rev. 4, and RG 1.240, Rev. 0; therefore, the NRC staff finds the bias to be appropriate for this application.

3.2.5.2 Fuel Assembly Manufacturing Tolerances and Uncertainties The licensee included uncertainties for the following fuel assembly manufacturing tolerances:

fuel rod pitch, fuel pellet diameter, fuel cladding ID, fuel cladding outer diameter (OD), guide tube ID, guide tube OD, fuel enrichment, and fuel pellet density. For the storage configuration that credits Gadolinia, uncertainties for Gadolinia manufacturing tolerances were determined as well. The licensees treatment of fuel assembly manufacturing tolerance and uncertainties is generally consistent with the guidance in NEI 12-16, Rev. 4, and RG 1.240, Rev. 0, and is, therefore, acceptable.

3.2.5.3 Spent Fuel Characterization Characterization of fresh fuel is based primarily on U-235 enrichment and various manufacturing tolerances. The manufacturing tolerances are typically manifested as uncertainties, as discussed above, or are bounded by values used in the analysis. These tolerances and bounding values would also apply to the spent nuclear fuel. Common industry practice has been to treat the uncertainties as unaffected by the fuel depletion. The characterization of spent nuclear fuel is complex. Its characterization is based on the specifics of its initial conditions and its operational history in the reactor. That characterization has three main areas: depletion uncertainty, the axial and radial apportionment of the burnup, and the core operation that achieved that burnup. These characteristics are evaluated in the following sections.

3.2.5.3.1 Depletion Uncertainty See section 3.2.2.1, Depletion Computer Code Validation above.

The licensee used 5 percent of the delta k between fresh and Burnup of interest as the depletion uncertainty.

3.2.5.3.2 Axial Apportionment of the Burnup or Axial Burnup Profile Another important aspect of fuel characterization is the selection of the axial burnup profile. At the beginning of life, a PWR fuel assembly will be exposed to a near-cosine axial-shaped flux, which will deplete fuel near the axial center at a greater rate than at the ends. As the reactor continues to operate, the cosine flux shape will flatten because of the fuel depletion and fission-product buildup that occurs near the center. Near the fuel assembly ends, burnup is suppressed due to neutron leakage. If a uniform axial burnup profile is assumed, then the burnup at the ends is over predicted. Analysis discussed in NUREG/CR-6801, Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analysis (ML031110292), has shown that, at assembly burnups above about 10 to 20 gigawatt-days per metric ton of uranium (GWd/MTU), the use of a uniform axial burnup profile results in an under prediction of Keff; generally, the under prediction becomes larger as burnup increases. This is what is known as the end effect. Proper selection of the axial burnup profile is necessary to ensure Keff is not under-predicted due to the end effect.

NEI 12-16, Rev. 4, section 5.1.4, Axial Burnup Distribution, contains guidance on using appropriate axial burnup distributions for use in SFP criticality analyses. The applicant used a variation of Option 2 from NEI 12-16, Rev. 4, section 5.1.4 to develop its axial burnup distributions. The variation is consistent with the intent of Option 2. Consequently, the staff finds that the treatment of axial burnup distribution by the licensee is acceptable.

3.2.5.3.3 Radial Burnup Distribution The applicants analysis did not consider radial burnup distribution. Due to the neutron flux gradients in the reactor core, assemblies can show a radially tilted burnup distribution (i.e.,

differences in burnup between portions or quadrants of the cross section of the assembly). The licensees analysis did not consider the effect of planar burnup distribution on reactivity.

NUREG/CR-6800, Assessment of Reactivity Margins and Loading Curves for PWR Burnup-Credit Cask Designs (ML031110280), estimates the effect of a radial burnup profile.

NUREG/CR-6800 estimates the effect to be approximately 0.002 k. The licensees analysis has sufficient margin to accommodate the potential radial burnup distribution. The staff finds the omission of the planar burnup distribution on reactivity to be acceptable because the expected reactivity impact is accommodated by the available margin in the licensees analysis.

3.2.5.3.4 Burnup History/Core Operating Parameters NEI 12-16, Rev. 4, section 4.2, Reactivity Effects of Depletion for PWRs, contains guidance on modeling reactor operating parameters to determine the post irradiation isotopic content of used nuclear fuel.

HI-2230346, Revision 2, section 3.3.2, Design Basis Core Operation Parameters, indicates the applicant considered three sets of nominal core operating parameters: Pre-Extended Power Uprates (EPUP, EPU, and 24-month-cycle for the following core operating parameters: core moderator temperature, fuel temperature, reactor specific power, and cycle average soluble boron concentration.

HI-2230346, Revision 2, section 3.3.9.1, Core Operating Parameters, performed a sensitivity analysis on those parameters and included resultant increases as biases.

The NRC staff considers this treatment of the core operating parameters as conservative and therefore acceptable.

3.2.5.3.5 Integral and Fixed Burnable Absorbers HI-2230346, Revision 2, section 3.3.9.3.3, Integral Burnable Absorbers, addresses the effect of integral burnable absorbers (IBA) currently in use at SL2 on the post-irradiated reactivity of the fuel assemblies. SL2 uses two types of IBAs: Gadolinium and absorber rods containing Boron Carbide (B4C) replacing fuel rods. HI-2230346, Revision 2, section 3.3.9.3.3, cites previous work indicating its conservative to model the SL2 IBAs as fuel rods having the same nominal enrichment as the rest of the fuel assembly. The NRC staff accepts this modeling of SL2 IBAs.

3.2.5.3.6 Control Element Assembly Usage HI-2230346, Revision 2, section 3.3.9.3.2, Reactivity Control Devices, addresses the effect of operation with CEAs inserted. The section states, The SL2 reactor always operates ARO (All Rods Out) at full power operations. However, the section calculates a penalty to be applied to fuel assemblies that should experience operation with CEAs inserted. The NRC staff accepts this modeling of CEA usage.

3.2.5.3.7 Credited Nuclides HI-2230346, Revision 2, section 3.3.9, Spent Fuel Reactivity Calculation, addresses spent fuel isotopes modeled in the analysis. The applicant removed all volatile and gaseous isotopes from the model. The NRC staff determined that this is more conservative than the guidance in NEI 12-16, Rev. 4, and RG 1.240, Rev. 0, and, therefore, is acceptable.

3.2.6 Non-Standard Fuel Configurations/Reconstituted Fuel In the LAR the licensee indicates it does not have non-standard fuel configurations.

HI-2230346, Revision 2, section 3.5.5, Fuel Assembly Inspection and Reconstitution, addresses fuel assembly inspection, testing, and reconstitution. The licensee has identified five temporary storage configurations that it can use if fuel assembly inspection, testing, or reconstitution is necessary. The application analyzed each to demonstrate they meet 10 CFR 50.68(b)(4).

3.2.7 Abnormal and Accident Conditions The regulations in 10 CFR 50.68(b)(4) require that the Keff of the SL2 SFP 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. This requirement applies to all normal and abnormal/accident conditions.

HI-2230346, Revision 2, section 3.6, Abnormal and Accident Conditions, addressing accident conditions are considered for the SL2 SFP. Some of the abnormal conditions considered, such as the Dropped Assembly - Horizontal (HI-2230346, Revision 2, section 3.6.2), are dispositioned using engineering judgement. Others are dispositioned using calculations. Those are discussed further in HI-2230346, Revision 2, section 8.13, Accident Conditions. The analysis determined the limiting accident, of those considered, was what the analysis labeled the Incorrect Loading Curve (Multiple Misload). The results are shown in HI-2230346, Revision 2, Table 8-27, Maximum keff Calculation for the Incorrect Loading Curve (Multiple Misload)

Accident. The calculated SFP soluble boron requirement is 1882.7 ppm. The current SL2 TS 3.7.14, Spent Fuel Storage Pool Boron Concentration, requires the SFP soluble boron concentration to be greater than or equal to 1900 ppm. If the effect of fuel assembly radial burnup distribution (see Section 3.2.5.3.3 above) is included, then the required soluble boron concentration would exceed 1900 ppm. In its letter dated July 29, 2025, in response to RAI #3, the licensee increased the SL2 TS 3.7.14 SFP required soluble boron concentration to 2000 ppm. The NRC staff determined that this change provides margin to cover the potential for the small non-conservatism and is therefore acceptable.

HI-2230346, Revision 1, did not consider the misloading of multiple fresh fuel assemblies.

HI-2230346, Revision 1, Section 3.6.8, Incorrect Burnup (Multiple Misload), indicated that the analysis considering the misloading of multiple fresh fuel assemblies was not credible at SL2. In response to an NRC RAI, the licensee made changes to the LAR as detailed in the response to RAI # 2 and revised Holtec report HI-2230346, Revision 2. Those changes include the following:

Designating specific CPR and R1 areas that can only have fresh fuel.

Specific training for fuel handing operators and Reactor Engineers that only fresh fuel can be stored in those areas.

Enhanced procedural requirement These changes, coupled with the applicants robust set of administrative controls for developing the fuel movement instructions for fuel handling operators, align with NRC staffs past acceptance of overall controls necessary to minimize the probability of misloading multiple fresh fuel assemblies. Therefore, the disposition of accident conditions is acceptable.

3.3 The SL2 NFV NCS Analysis 3.3.1 NFV Overview The analysis supporting the applicants request for the modification to its NFV storage controls is contained in the Enclosure of the April 30, 2024, letter: Holtec report HI-2230455, Revision 2, Criticality Safety Analysis of NFV for St. Lucie Unit 2. The NFV storage controls were not modified by the applicants July 29, 2025, letter. Analysis of the NFV considers both fully flooded and optimum moderation scenarios and involves only fresh fuel.

The SL2 NFV has one rack design and is described in Holtec report HI-2230455, Revision 2, section 5.2, New Fuel Vault Specifications.

3.3.2 Computational Methods As stated in section 3.2.2 above, MCNP5 Version 1.51, was used to perform the calculations estimating keff in the SL2 NFV NCS analysis.

3.3.2.1 SFP keff Computer Code Validation Holtec report HI-2230455, Revision 2, section 3.2.2, MCNP Validation, describes the validation of MCNP Version 1.51 and ENDF/B-VII cross-sections used in the analysis. The applicant states the validation was performed in accordance with NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology (ML050250061). Using NUREG/CR-6698 is consistent with the guidance in NEI 12-16, Rev. 4, and RG 1.240, Rev. 0.

Per Holtec International HI-2230455, Revision 2, the details of the MCNP validation are based on results of Nuclear Group Computer Code Benchmark Calculations, HI-2104790, Revision

3. HI-2230455, Revision 2, added eight more criticality experiments with low moderator density to cover the optimum moderation scenarios. HI-2104790, Revision 3, was not submitted as part of the application. It is not currently on the docket, nor was it part of the audit conducted during the review. Therefore, the NRC staff relied upon the summary information provided in HI-2230455, Revision 2, Appendix C Table C-1, Description of the New Experiments, Table C-2, Summary of Area of Applicability of the MCNP Benchmark, Table C-3, MCNP Benchmark Analysis for Various Fuel and Water Subsets of Experiments, Table C-4 Significant Trending Analysis for St Lucie Unit 2 NFV Parameters, Table C-5, Summary of MCNP Code Validation Bias and Bias Uncertainty, and 10 CFR 50.9(a). The NRC staff concludes that the information provides reasonable assurance that MCNP5 Version 1.51, with the ENDF/B-VII cross-sections, was validated appropriately for the SL2 NFV analysis.

3.3.3 NFV Storage Racks The SL2 NFV storage racks consist of 80 square cavities, each capable of containing one new 16x16 array fresh fuel assembly and are designed as seismic Category I. The storage cavities are comprised of four stainless steel angles connected by horizontal ties and a hinged plate cover.

3.3.3.1 NFV Storage Racks Manufacturing Tolerances and Uncertainties For the SL2 NFV rack, the licensee included uncertainties for the following SFP rack manufacturing tolerances: storage cell ID and storage cell pitch. The NRC staff concludes that the licensees treatment of SFP storage rack manufacturing tolerance and uncertainties is generally consistent with the guidance in NEI 12-16, Rev. 4 and RG 1.240, Rev. 0, and therefore, acceptable.

3.3.4 Fuel Assembly 3.3.4.1 Fuel Assembly Design Holtec report HI-2230455, Revision 2, section 3.3.1, Design Basis Fuel Assembly Design, describes the selection of the fuel assembly design used in the analysis.

3.3.4.2 Fuel Assembly Manufacturing Tolerances and Uncertainties The licensee included uncertainties for the following fuel assembly manufacturing tolerances:

fuel rod pitch, fuel pellet diameter, fuel cladding ID, fuel cladding OD, guide tube thickness, fuel enrichment, and fuel pellet density. The NFV analysis does not credit Gadolinia, so no uncertainties for Gadolinia manufacturing tolerances were calculated. NRC staff concludes that the licensees treatment of fuel assembly manufacturing tolerance and uncertainties is generally consistent with the guidance in NEI 12-16, Rev. 4 and RG 1.240, Rev. 0, and therefore, acceptable.

3.3.5 NFV 3.3.5.1 NFV Water Temperature Guidance in NEI 12-16, Rev. 4 and RG 1.240, Rev. 0, states the NCS analysis should be done at the temperature corresponding to the highest reactivity. Holtec report HI-2230455, Revision 2, section 3.3.2, Reactivity Effect of Water Temperature and Density, discusses the applicants sensitivity analyses on NFV water temperature. It appears the applicant found the most reactive temperature and either applied a bias or used the most reactive temperature as the temperature in the analysis. Therefore, the NRC staff finds that the water temperature was handled appropriately in the licensees criticality analysis.

3.3.5.2 Eccentric Positioning Holtec report HI-2230455, Revision 2, section 3.3.5, Eccentric Fuel Positioning, discusses the applicants sensitivity analyses on fuel assembly positioning within NFV storage cells. While the analysis only considered two positioning scenarios, the margin shown in HI-2230455, Revision 2, Table 8-6, Summary of the Analysis for NFV, makes it unlikely a significantly more reactive eccentric positioning would be found to overcome the margin. Therefore, the NRC staff finds that the Eccentric Fuel Positioning analysis is acceptable.

3.4 Technical Specification Changes The staff reviewed the proposed changes to TS 3.7.15 and TS 4.3 and finds that the changes conform with the SL2 NFV and SFP NCS analyses evaluated above therefore meet 10 CFR 50.36(c) requirements for safe operation of the facility.

3.5 Conclusions The NRC staff has completed its review of the SL2 NFV and SFP NCS analyses, which are documented in the licensees April 30, 2024, and July 29, 2025, letters. The NRC staff finds that the licensee has demonstrated that the requested changes will meet 10 CFR 50, Appendix A, Criterion 62, 10 CFR 50.68(b)(2), 10 CFR 50.68(b)(3), 10 CFR 50.68(b)(4) and 10 CFR 50.36(c) requirements.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Florida State official was notified of the proposed issuance of the amendment on August 29, 2025. The State official provided no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR, Part 20. The NRC staff has determined that the amendment involves 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 amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on July 9, 2024 (89 FR 56438). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

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

Principal Contributor: K. Wood, NRR Date of Issuance: September 26, 2025

ML25241A191 NRR-058 OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DSS/STSB/BC NRR/DSS/SFNB/BC NAME NJordan ABaxter SMehta SKrepel DATE 09/02/25 09/04/25 09/15/2025 09/15/2025 OFFICE NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME DWrona NJordan DATE 09/26/2025 09/26/2025

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