ML25324A332
| ML25324A332 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 12/17/2025 |
| From: | Michael Mahoney Plant Licensing Branch IV |
| To: | Entergy Operations |
| References | |
| EPID L-2024-LLA-0158 | |
| Download: ML25324A332 (0) | |
Text
December 17, 2025 Vice President, Operations Entergy Operations, Inc.
Grand Gulf Nuclear Station P.O. Box 756 Port Gibson, MS 39150
SUBJECT:
GRAND GULF NUCLEAR STATION, UNIT 1 - ISSUANCE OF AMENDMENT NO. 239 - RE: REVISE TECHNICAL SPECIFICATIONS; CRITICALITY SAFETY ANALYSIS,TECHNICAL SPECIFICATION 4.3.1, CRITICALITY AND TECHNICAL SPECIFICATION 5.5.15, SPENT FUEL STORAGE RACK NEUTRON ABSORBER MONITORING PROGRAM (EPID L-2024-LLA-0158)
Dear Sir or Madam:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 239 to Renewed Facility Operating License No. NPF29 for the Grand Gulf Nuclear Station, Unit 1. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated November 25, 2024, as supplemented by letters dated December 16, 2024, and May 5, 2025.
The amendment allows crediting of NETCO-SNAP-IN neutron absorbing rack inserts in the criticality safety analysis (CSA) for the storage rack cells in the stations spent fuel storage facility. The amendment revises the TSs regarding criticality design features of the spent fuel storage racks as contained in TS 4.3, Fuel Storage, Subpart 4.3.1, Criticality, to specifically identify the neutron absorbing inserts, remove requirements for Region II storage racks, and to update the value of K-infinity used in the CSA, consistent with Standard Technical Specifications (STS). Lastly, the amendment adds a new program requirement in TS 5.5, Programs and Manuals, that implements a monitoring program for the neutron absorbing rack inserts, consistent with STS improvement initiatives.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.
Sincerely,
/RA/
Michael Mahoney, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-416
Enclosures:
- 1. Amendment No. 239 to NPF-29
- 2. Safety Evaluation cc: Listserv ENTERGY OPERATIONS, INC.
SYSTEM ENERGY RESOURCES, INC.
COOPERATIVE ENERGY, A MISSISSIPPI ELECTRIC COOPERATIVE ENTERGY MISSISSIPPI, LLC DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 239 Renewed License No. NPF-29 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Entergy Operations, Inc. (the licensee), dated November 25, 2024, as supplemented by letters dated December 16, 2024, and May 5, 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.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-29 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 239 are hereby incorporated into this renewed license. Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented no later than December 31, 2027.
FOR THE NUCLEAR REGULATORY COMMISSION Tony Nakanishi, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License No. NPF-29 and the Technical Specifications Date of Issuance: December 17, 2025 SAMSON LEE Digitally signed by SAMSON LEE Date: 2025.12.17 09:42:55 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 239 RENEWED FACILITY OPERATING LICENSE NO. NPF-29 GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416 Replace the following pages of Renewed Facility Operating License No. NPF-29 and the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Renewed Facility Operating License Remove Insert Technical Specifications Remove Insert 4.0-1 4.0-1 4.0-2 4.0-2 5.0-16d
4 Amendment No. 239 amended, are fully applicable to the lessors and any successors in interest to those lessors, as long as the renewed license of GGNS Unit 1 remains in effect.
(b)
SERI is required to notify the NRC in writing prior to any change in (i) the terms or conditions of any new or existing sale or lease agreements executed as part of the above authorized financial transactions, (ii) the GGNS Unit 1 operating agreement, (iii) the existing property insurance coverage for GGNS Unit 1 that would materially alter the representations and conditions set forth in the Staff's Safety Evaluation Report dated December 19, 1988 attached to Amendment No. 54. In addition, SERI is required to notify the NRC of any action by a lessor or other successor in interest to SERI that may have an effect on the operation of the facility.
C.
The renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1)
Maximum Power Level Entergy Operations, Inc. is authorized to operate the facility at reactor core power levels not in excess of 4408 megawatts thermal (100 percent power) in accordance with the conditions specified herein.
(2)
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 239 are hereby incorporated into this renewed license. Entergy Operations, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
During Cycle 19, GGNS will conduct monitoring of the Oscillation Power Range Monitor (OPRM). During this time, the OPRM Upscale function (Function 2.f of Technical Specification Table 3.3.1.1-1) will be disabled and operated in an indicate only mode and technical specification requirements will not apply to this function. During such time, Backup Stability Protection measures will be implemented via GGNS procedures to provide an alternate method to detect and suppress reactor core thermal hydraulic instability oscillations. Once monitoring has been successfully completed, the OPRM Upscale function will be enabled and technical specification requirements will be applied to the function; no further operating with this function in an indicate only mode will be conducted.
Design Features 4.0 GRAND GULF 4.0-1 Amendment No. 195, 239 4.0 DESIGN FEATURES 4.1 Site Location The site for Grand Gulf Nuclear Station is located in Claiborne County, Mississippi on the east bank of the Mississippi River, approximately 25 miles south of Vicksburg and 37 miles north-northeast of Natchez. The exclusion area boundary shall have a radius of 696 meters from the centerline of the reactor.
4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 800 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material, and water rods. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used.
Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.
4.2.2 Control Rod Assemblies The reactor core shall contain 193 cruciform shaped control rod assemblies. The control material shall be boron carbide or hafnium metal, or both.
4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
a.
keff LIIXOO\\IORRGHGZLWKXQERUDWHGZDWHUZKLFKLQFOXGHVDQ
allowance for uncertainties as described in Section 9.1.2 of the UFSAR; b.
A nominal fuel assembly center to center storage spacing of 6.26 inches, with a neutron absorber insert within the storage cells, in the spent fuel storage pool and in the upper containment pool.
c.
Fuel assemblies having a maximum K-infinity of 1.29 in the normal reactor core configuration at cold conditions; (continued)
Design Features 4.0 GRAND GULF 4.0-2 4.0 DESIGN FEATURES 4.3.1.1 (continued) d.
Fuel assemblies having a maximum nominal U-235 enrichment of 4.9 weight percent
(continued)
Amendment No. 195, 239
GRAND GULF 5.0-16d Amendment No. 239 Programs and Manuals 5.5 5.0 Programs and Manuals (continued) 5.5.15 Spent Fuel Storage Rack Neutron Absorber Monitoring Program This program provides controls for monitoring the condition of the neutron absorber inserts used in the high density spent fuel storage racks to verify the Boron-10 areal density is consistent with the assumptions in the spent fuel pool criticality analysis. The program shall be in accordance with NEI 16-03-A, Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools, Revision 0, May 2017.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 239 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-29 ENTERGY OPERATIONS, INC., ET AL.
GRAND GULF NUCLEAR STATION, UNIT 1 DOCKET NO. 50-416
1.0 INTRODUCTION
By application dated November 25, 2024 (Reference 1), as supplemented by letters dated December 16, 2024 (Reference 2), and May 5, 2025 (Reference 3), Entergy Operations, Inc.
(the licensee), requested changes to the Technical Specifications (TSs) for Grand Gulf Nuclear Station, Unit 1 (Grand Gulf).
The proposed changes would allow crediting of NETCO-SNAP-IN neutron absorbing rack inserts in the criticality safety analysis (CSA) for the storage rack cells in the stations spent fuel storage facility. The amendment revises the TSs regarding criticality design features of the spent fuel storage racks as contained in TS 4.3, Fuel Storage, Subpart 4.3.1, Criticality, to specifically identify the neutron absorbing inserts, remove requirements for Region II storage racks, and to update the value of K-infinity used in the CSA, consistent with Standard Technical Specifications (STS). Lastly, the amendment adds a new program requirement in TS 5.5, Programs and Manuals, that implements a monitoring program for the neutron absorbing rack inserts, consistent with STS improvement initiatives.
The supplemental letters dated December 16, 2024, and May 5, 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 March 25, 2025 (90 FR 13635).
2.0 REGULATORY EVALUATION
2.1
System Description
The licensee provided the following system description in its license amendment request (LAR) dated November 25, 2024, which states in part:
The [Grand Gulf] SFP [spent fuel pool] contains 16 high density fuel rack modules in 5 different module sizes. The module types are labeled A, B, C, D and H on UFSAR [Updated Final Safety Analysis Report] Figure 9.1-40a
[Reference 4], which also shows their relative placement. The storage rack cells with a center-to-center spacing of 6.26 inches (nominal). There are a total of 4348 fuel storage locations within the spent fuel pool.
.Criticality in new and spent fuel storage is prevented by the geometrically safe configuration of the storage rack combined with the use of neutron absorber (Boraflex) material in the high-density storage racks. There is either sufficient spacing or neutron poison material between the assemblies to assure that the array, when fully loaded, is substantially subcritical.
In order to accommodate known and possible future Boraflex degradation and maintain Keff [reactivity coefficient k-effective] criterion of less than or equal to 0.95, the [Grand Gulf] fuel pool racks are allocated into Region I and Region II locations. The Region I rack locations are those locations which are above the Boraflex panel areal density limit and below the dose threshold for accelerated gapping and are bounded by the EPRI [Electric Power Research Institute] model for shrinkage. The Region II rack locations are those locations which are below the Boraflex panel areal density limit or at or above the dose threshold for accelerated gapping and no credit is taken for the Boraflex panels in the criticality analysis in these locations.
Each [Grand Gulf] storage rack unit employs Boraflex as a fixed neutron absorber for criticality control, to ensure that the effective neutron multiplication factor (Keff) does not exceed the values and assumptions used in the CSA.
2.2 Proposed Changes The licensee proposed the following changes to the TSs. (deletions are shown in double strike through and additions in bold underline).
The licensee proposes the following changes to TS 4.3.1.1.b:
A nominal fuel assembly center to center storage spacing of 6.26 inches, with a neutron absorber insert within the storage cells, in the spent fuel storage pool and in the upper containment pool in the storage racks.
The licensee proposes the following changes to TS 4.3.1.1.c:
Fuel assemblies having a K-infinity of 1.26 1.29 in the normal reactor core configuration as cold conditions; The licensee proposes to delete TS 4.3.1.1.e and Figure 4.3.1 Region II 4x4 Storage Configuration in its entirety.
The licensee proposes adding a new program requirement in TS 5.5, Programs and Manuals, as TS 5.5.15, Spent Fuel Storage Rack Neutron Absorber Monitoring Program, which will state, as follows:
This program provides controls for monitoring the condition of the neutron absorber inserts used in the high density spent fuel storage racks to verify the Boron-10 areal density is consistent with the assumptions in the spent fuel pool criticality analysis. The program shall be in accordance with NEI [Nuclear Energy Institute] 16-03-A, Guidance for Monitoring Fixed Neutron Absorbers in Spent Fuel Pools, Revision 0, May 2017 [Reference 5].
2.3 Regulatory Requirements 2.3.1 Applicable Regulatory Requirements As required by Title 10 of the Code of Federal Regulations (10 CFR) 50.36(c)(4), Design features, the TSs will 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 regulations in 10 CFR 50.36(c)(5), Administrative controls, state, in part, that the TSs will include provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.
The regulations in 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix A, General Design Criteria for Nuclear Power Plants, General Design Criterion (GDC) 61, Fuel storage and handling and radioactivity control, state, in part, that, These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety and the regulations in GDC 62, Prevention of criticality in fuel storage and handling, state 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 Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,Section V, Instructions, Procedures, and Drawings, state, in part, that, Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances.
The regulations in 10 CFR 50.68, Criticality accident requirements, state, in part, that each holder of an operating license shall comply with either 10 CFR 70.24, Criticality accident requirements, or 10 CFR 50.68(b). The licensee has elected to comply with 10 CFR 50.68(b) and therefore does not need to comply with 10 CFR 70.24.
2.3.2 Applicable Regulatory Guidance The relevant regulatory guidance documents used to assist the NRC staff in its review of compliance with the regulatory requirements listed in section 2.1 are listed below.
NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, (SRP) Section 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling, Revision 3; and Section 9.1.2, New and Spent Fuel Storage, Revision 4 (References 6 and 7, respectively), provide guidance regarding the specific acceptance criteria and review procedures to ensure that proposed changes satisfy the requirements in 10 CFR 50.68, GDC 61 and GDC 62.
NRC Regulatory Guide (RG) 1.240 Fresh and Spent Fuel Pool Criticality Analysis (Reference 8), provides updated guidance to the NRC staff reviewer to address the increased complexity of recent SFP nuclear criticality safety (NCS) analyses and operations.
3.0 TECHNICAL EVALUATION
3.1 Background to the licensees LAR (Reference 1) provides a summary of the NCS analysis for Grand Gulf spent fuel storage racks. It describes the methodology and analytical models used in the NCS analysis to show that the Boraflex spent fuel storage racks maximum reactivity coefficient k-effective (keff) will be no greater than 0.95 if flooded with unborated water. No credit is taken for borated water. It also references a benchmarking evaluation performed for the Monte Carlo N-Particle (MCNP)-05P and TGBLA06 computer codes used in the analysis, to demonstrate the applicability of the codes to geometries and compositions being analyzed and to determine the code bias and uncertainty. The NRC has previously approved the use of this code package for similar SFP NCS analyses involving the use of NETCO-SNAP-IN neutron absorbing inserts for other facilities (e.g., River Bend Station License Amendment No. 201 (Reference 9).
The use of NETCO-SNAP-IN neutron absorbing inserts as a replacement for the Boraflex in serving the neutron absorption function for criticality control is being proposed because Boraflex is known to have severe degradation issues. The licensee would install the inserts in every Boraflex storage cell to qualify all usable cells for storage.
The Grand Gulf SFP currently utilizes storage racks supplied by Joseph Oat Corporation, which utilizes Boraflex as the neutron absorber material (NAM). There are 16 Boraflex racks with a rack pitch of 6.259 inches. Boraflex NAM has a long history of degradation due to dissolution and loss of Boron -10 (B10). This has been a safety-significant concern in a number of facilities that has led to the submission of amendments for those facilities to, like the current Grand Gulf submittal, credit another NAM in the NCS analysis to meet the requirements of 10 CFR 50.68 and GDC 62.
The changes proposed in the LAR would allow the licensee to credit NETCO-SNAP-IN neutron absorbing inserts in the Grand Gulf NCS analysis in place of Boraflex. These inserts are a homogenous mixture of boron carbide and aluminum. They are fashioned by bending a thin sheet of the insert material into a chevron shape with a bend angle slightly greater than 90 degrees. This allows the insert to hold itself in place inside an SFP storage cell by a retention force. The insert spans the entire height of the SFP storage cell. The NETCO-SNAP-IN neutron absorbing inserts would be installed in the Boraflex racks oriented such that the north and east sides of every cell containing an insert would have a wing of the insert. This orientation would place a wing/sheet of the insert between every fuel assembly in the Boraflex racks. Since cells would only contain one NETCO-SNAP-IN neutron absorbing insert, the south and west peripheral border of the Boraflex racks would not have a wing/sheet of the inserts; the reactivity effect of storing along the south and west borders is discussed in section 3.3.4 of this safety evaluation (SE).
The NETCO-SNAP-IN neutron absorbing inserts would be held in place by their own retention force. The licensee has accounted for a missing insert in its NCS analysis to bound incidents such as the inadvertent removal of an insert and to ensure that the limits in the TSs are not exceeded.
Grand Gulf used the peak Standard Cold Core Geometry (SCCG) methodology for its criticality analysis. Due to the usage of burnable absorbers in boiling-water reactor (BWR) fuel rods, the peak reactivity of the fuel increases during the initial irradiation. That peak reactivity in the core then correlates to the infinite multiplication factor (k) in the SFP. The SFP NCS analysis considers three types of fuels: Global Nuclear Fuels (GNF) designed GE14, GNF2, and GNF3.
The SCCG of 1.29 bounds the legacy fuel. The licensee determined GNF2 fuel to be bounding of the other fuel types in the NCS analysis. Consideration was given for gadolinia and reconstituted fuel where the rod contained fuel is replaced with another fuel or non-fueled rod.
3.2 SFP NCS Analysis Methods The methods used for the NCS analysis for fuel in the Grand Gulf SFP are described in section 3.0 of attachment 7 to the licensees LAR (Reference 1). The licensee used Nuclear Energy Institute (NEI) 12-16, Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants (Reference 10), as guidance for this analysis. The Grand Gulf SFP NCS analysis utilizes an MCNP-05P/TGBLA06 code package, which has been previously approved in a similar SFP NCS analysis involving NETCO-SNAP-IN neutron absorbing inserts for Peach Bottom Atomic Power Station, Units 2 and 3 (Peach Bottom) (Reference 11). The area of applicability for the benchmarking analyses was verified to be applicable to the Grand Gulf SFP.
3.2.1 Computational Methods Two computational methods were used in the SFP NCS analysis. The GNF lattice design code TGBLA06 was used to calculate the maximum SCCG, uncontrolled peak in-core keff values.
Due to the presence of burnable absorbers, the maximum keff fuel bundle will have some degree of burnup. The burned fuel compositions were then imported into MCNP-05P, the GNF proprietary version of MCNP, to obtain the fuel storage rack keff values. The TBGLA06 code uses the ENDF/B-V neutron cross section library.
Based on the use of acceptable inputs and appropriate convergence checks to ensure that an accurate keff is determined, the NRC staff finds that the computational methods used for the SFP NCS analysis are acceptable.
3.2.2 Computer Code Validation The SFP NCS analysis utilizes the highest in-core keff at SCCG. Due to integral burnable absorbers such as gadolinia, this point does not occur at the beginning of life. It is necessary to consider the validation of computer codes and data used to calculate burned fuel compositions, and the computer code data that utilize the burned fuel compositions to calculate keff for systems with burned fuel. The purpose of the criticality code validation is to ensure that appropriate code bias and bias uncertainty are determined for use in the criticality evaluation.
RG 1.240 references NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology (Reference 12), which states, in part, that:
In general, the critical experiments selected for inclusion in the validation must be representative of the types of materials, conditions, and operating parameters found in the actual operations to be modeled using the calculation method. A sufficient number of experiments with varying experimental parameters should be selected for inclusion in the validation to ensure an area of applicability as feasible and statistically significant results.
The NRC staff used NUREG/CR-6698 as guidance for its review of the computer code validation methodology provided in the application. The basic elements of validation are outlined in NUREG/CR-6698, including identification of operating conditions and parameter ranges to be validated, selection of critical benchmarks, modeling of benchmarks, statistical analysis of results, and determination of the area of applicability (AOA).
The licensee validated the code by comparing the calculated keff values with the measured keff values of a set of critical experiments. The critical experiments should result in an AOA that bound all parameters found in the Grand Gulf SFP. The critical experiments used by the licensee bounded all the Grand Gulf SFP parameters, except for SFP water temperature, as both ends of the range of Grand Gulf SFP water temperatures are outside the AOA created by the validation critical experiments. As part of the analysis, the licensee performed a sensitivity analysis based on the range of SFP water temperature. The sensitivity analysis determined that the Grand Gulf SFP has a negative moderator temperature coefficient. The sensitivity analysis also determined that there is essentially no difference between the nominal temperature used in the analysis, which is within the AOA, and the lowest temperature in the range.
The licensee identified the applicable operating conditions for the validation (e.g., fuel assembly materials and geometry, enrichments, fuel properties, neutron absorbers, moderators, etc.). The licensee compared the spectral parameters (e.g., energy of average lethargy causing fission (EALF)) between the benchmarks and the Grand Gulf SFP conditions to demonstrate that the selected benchmarks are applicable.
Based on its review of the validation database and its applicability to the compositions, geometries, and methodologies used in the licensees SFP NCS analysis, the NRC staff finds that the code validation was acceptable and that all identified biases and uncertainties were propagated appropriately.
3.2.3 Trend Analysis As part of the statistical analysis of the results of the SFP NCS analysis, the licensee provided trending analysis between the ratio of calculated keff to measured keff of the critical experiments to the parameters found in table 23 of attachment 7 to the LAR (Reference 1). The purpose of a trend analysis is to determine if any adverse trending is identified in the pool of experiments.
Adverse trending may bring the accuracy of the NCS methodology into question.
The licensee utilized two methods to determine if there was any trending between the parameters: (1) the linear correlation coefficient (r2) of a linear regression fitted equation and (2) a chi squared test. In both cases, the data showed no significant trending.
Based on a review of the trend analysis, the NRC staff finds that the trend analysis is acceptable.
Based on the above, the NRC staff finds that the SFP NCS analysis methods are consistent with the guidance in RG 1.240 and therefore are acceptable.
3.3 SFP NCS Analysis The licensee determined the maximum reactivity of the fuel storage racks considering all biases, tolerances, and uncertainties. Grand Gulf did this by summing the normal reactivity (knormal), the reactivity uncertainties (kuncertainty), and the reactivity biases (kbias). Knormal was determined by analyzing all nominal conditions and using the highest value of reactivity. The positive reactivity increase from the normal and abnormal cases was included in kbias. Kuncertainty was determined by taking the root sum square of the reactivity effects due to uncertainties. Kbias was determined by evaluating the reactivity effects of different depletion parameters and abnormal configurations.
When cases for kbias or kuncertainty were mutually exclusive, the larger positive effect was used (e.g., high and low SFP moderator temperature are considered to be the same term, but only low moderator temperature contributes to kbias because it is the largest positive value).
3.3.1 SFP Water Temperature The guidance in NEI 12-16, Revision 4, which is endorsed by RG 1.240, states that the NCS analysis should be done at the temperature corresponding to the highest reactivity. The licensees analysis calculated a base keff using a nominal temperature. As part of the analysis, the licensee performed a sensitivity analysis based on the range of SFP water temperature. The sensitivity analysis determined that the Grand Gulf SFP has a negative moderator temperature coefficient. The sensitivity analysis also determined that there is essentially no difference between the nominal temperature used in the analysis, which is within the AOA, and the lowest temperature in the range. In this analysis, the temperature corresponding to the highest reactivity is essentially equivalent to the nominal temperature used in the analysis. Therefore, the NRC staff finds that the NCS analysis is acceptable with respect to SFP water temperature.
3.3.2 SFP Storage Rack Models The licensee used a 2-dimensional (2D) infinite SFP storage rack array to conservatively estimate the reactivity of the system. Every fuel assembly position contains the maximum reactivity design basis fuel assembly as determined in section 5.3 of attachment 7 to the licensees LAR (Reference 1). This analysis does not credit Boraflex as a neutron absorbing material and, therefore, the licensee chose to model the Boraflex inside the storage rack as water. This substitution of materials is acceptable because the wrappers around the Boraflex do not necessarily provide a watertight seal and modeling water instead of Boraflex material is conservative with respect to reactivity. The dimensions and uncertainties of the storage racks can be found in table 12 of attachment 7 to the licensees LAR.
The inserts are positioned such that the insert wings occupy the north and east sides of each storage cell. The inserts span the entire height of the fuel assembly. The purpose of the rack inserts is to absorb neutrons, which is accomplished by the boron carbide in the inserts. The rack inserts are modeled such that the boron areal density is 0.0139 boron-10 atoms in atoms per square centimeter (g B10/cm2). This is slightly less than the minimum certified acceptable boron areal density to account for potential manufacturing uncertainties.
3.3.2.1 SFP Storage Rack Model Manufacturing Tolerances and Uncertainties The manufacturing tolerances of the storage racks and rack inserts are found in table 12 of attachment 7 to the licensees LAR. The manufacturing tolerances of the storage racks can contribute to SFP reactivity. According to the guidance in NEI 12-16, Revision 4, as endorsed by RG 1.240, the determination of the maximum keff should consider either: (1) a worst-case combination with mechanical and material conditions set to maximize keff or (2) a sensitivity study of the reactivity effects of tolerance variations. In section 5.6.1 of attachment 7 to the LAR, the licensee stated the parameters that were varied to account for manufacturing tolerances.
The parameters were varied by at least 2 sigma ().
The licensee included tolerances for the rack insert thickness. The variation in thickness does not necessarily impact the ability of the insert to absorb neutrons because there is a minimum certified boron areal density for all inserts, regardless of thickness. The licensee also conservatively modeled the insert by not modeling the additional insert material which is bent at a 90-degree angle at the end of each wing.
The NRC staff finds the treatment of the storage rack tolerances and uncertainties acceptable based on the following discussion. The rack insert is modeled using a conservative boron areal density and a reduced wing length, resulting in a conservative estimate of the negative reactivity contribution of the insert.
3.3.3 Fuel Assembly Models 3.3.3.1 Bounding Fuel Assembly Design The Grand Gulf SFP may store GE14, GNF2, GNF3, and legacy fuel types previously used in Grand Gulf. Grand Gulf SFP NCS analysis includes analyses for GNF2 fuel. The analysis evaluated the different lattice designs, with some fuel assemblies having multiple lattice designs. The analysis included a determination of the most limiting lattice from the current and anticipated future fuel designs and then used that limiting lattice in the remainder of the analysis. The analysis used the rack efficiency method of evaluating the lattices. The rack efficiency is defined as the ratio of the in-rack keff to the peak in-core keff. The fuel assembly with the highest rack efficiency at the proposed TS limiting SCCG is determined to be the limiting lattice. The NRC previously approved this method for determining the limiting lattice in Amendment No. 201 for River Bend Station (Reference 9). Because the Grand Gulf analysis determined its limiting lattice consistent with a previously approved method, the NRC staff finds it acceptable.
3.3.3.2 Fuel Assembly Manufacturing Tolerances and Uncertainties According to the guidance in NEI 12-16, Revision 4, as endorsed by RG 1.240, the determination of the maximum keff should consider either: (1) a worst-case combination with mechanical and material conditions set to maximize keff or (2) a sensitivity study of the reactivity effects of tolerance variations. In its analysis, the licensee selected option 2 and performed a sensitivity study of 50 different lattice types found in table 13 of attachment 7 to its LAR. The selected lattice was used in the MCNP-05P/TGBLA06 code package with no further mechanical tolerances accounted for. Additional tolerances were considered in a tolerance study in which fuel assembly design parameters were varied by at least 2. The parameters that were varied can be found in section 5.6.1 of attachment 7 to the licensees LAR.
The NRC staff finds the treatment associated with the lattice design and depletion calculations acceptable based on the following discussion. The licensee performed a sensitivity study to find the most reactive fuel lattice. The licensee performed a tolerance study on the fuel assembly design parameters to calculate kuncertainty and only included positive reactivity contributions.
3.3.3.3 Spent Fuel Characterization To determine the effects of varied configurations on the depletion conditions, and therefore reactivity, Grand Gulf analyzed several cases where the power and environmental histories of the rods were varied. The cases analyzed can be found in section 5.5.1 of attachment 7 to the licensees LAR. These cases bound the conditions the fuel in the reactor would experience. The positive reactivity effects of these analyzed cases were added to the kbias term.
The highest reactivity of a BWR fuel assembly occurs with some degree of burnup due to burnable poisons in the assembly. Therefore, it is necessary to account for the treatment of burned fuel in the SFP. The licensee performed depletion calculations in TGBLA06 with parameters set to maximize burned fuel reactivity. Additionally, the TGLBLA06-defined lumped fission products are conservatively ignored in the in-rack keff calculations in MCNP-05P. The NRC staff notes that the licensee also did not include the effects of Xenon (Xe)-135 but does not necessarily consider this to be a conservative estimate. Xe-135 has a half-life of 9.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />; therefore, after several days it is reasonable to assume that no Xe-135 is present in the spent fuel. The time needed to transport fuel from the reactor core to the SFP may be enough to account for this decay time. Therefore, the NRC staff finds that not including the effects of Xe-135 reflects reality and is not necessarily a conservative approach.
The NRC staff finds the characterization of spent fuel in the NCS analysis acceptable because the analysis is conservative and accounts for varying depletion conditions and only included positive reactivity contributions.
3.3.3.4 Integral Burnable Absorbers The fuel types stored in the Grand Gulf SFP utilize gadolinia poison to help control reactivity and peaking within fuel assemblies. The concentration and position of gadolinia loading can have a significant impact on the reactivity of the assembly. The licensee considers the effects of gadolinia and variation in BWR fuel designs as it pertains to the SCCG method. A tolerance study was considered in the NCS analysis with respect to the gadolinia concentration.
The NRC staff finds the treatment of integral burnable absorbers in the NCS analysis acceptable based on the following discussion. The sensitivity study found in table 13 of attachment 7 to the licensees LAR is performed to select the lattice with the highest rack efficiency with an in-core keff greater than 1.29. This sensitivity study accounts for the gadolinia weight percent. The selected design basis lattice does not contain the least amount of gadolinia, but nonetheless represents the most reactive lattice.
Based on the above, the NRC staff finds that the NCS analysis is acceptable with respect to fuel assembly models.
3.3.4 Analysis of Normal and Abnormal Conditions The licensee considered the criticality effects of normal and abnormal conditions in calculating the kbias. The normal conditions can be found in section 5.4.1 of attachment 7 to the licensees LAR (Reference 1), and the abnormal conditions can be found in section 5.5.3 of attachment 7 to the LAR. Consistent with the rest of the analysis, only positive contributions to reactivity that are the largest for their respective term are considered. The case of eccentric loading was not considered. Consistent with NEI 12-16 (Reference 10), when a fuel assembly is surrounded by inserts with an areal density above 0.01 g B10/cm2, the most reactive configuration is a centrally located assembly.
The licensee analyzed credible normal scenarios, such as temperature variation and storage of non-channeled assemblies in determining the most reactive configuration. An additional normal scenario considered by the licensee was SFP moderator temperatures beyond the AOA. As stated previously, the licensee performed a sensitivity analysis based on the range of SFP water temperature, which determined that the Grand Gulf SFP has a negative moderator temperature coefficient. The results of this analysis are found in table 14 of attachment 7 to the licensees LAR.
The dropped or damaged fuel assembly abnormal condition assumes that a bundle is dropped, with resultant damage modeled as described in section 5.5.3 of attachment 7 to its LAR. The licensee also considered a scenario in which a dropped bundle comes to rest horizontally on top of the SFP storage rack. The licensee did not model this scenario given the following justification. The minimum separation of the fissile material between the dropped bundle and the stored fuel assemblies will be at least 12 inches. This provides enough neutron mean free paths to preclude neutron interactions that increase keff.
The licensee considered abnormal positioning of a fuel assembly outside of the fuel storage rack. There are numerous locations in which it is feasible for a bundle to be misplaced. The licensee analyzed three abnormal storage configurations: (1) when the bundle was placed between the storage rack and the wall, (2) when the bundle was pushed against the fuel rack corner, and (3) when the bundle was placed in unpoisoned equipment racks next to the fuel racks. For the case with the bundle outside the SFP, the licensee modeled this case as described in section 5.5.3 of attachment 7 to its LAR. The model for the misplaced bundle is shown in figure 12 of attachment 7 to the LAR. Different configurations were analyzed, and there is little sensitivity to the placement of the misplaced bundle. For the case with the bundle placed against the corner of the rack, the licensee modeled this case as described in section 5.5.3 of attachment 7 to the LAR. The model for the misplaced bundle is shown in figure 13 of attachment 7 to the LAR. For the case with the bundle in the unpoisoned equipment racks, the licensee modeled this case, as described in section 5.5.3 of attachment 7 to the LAR (Reference 1). Three configurations were considered, with all being bounded by the case where a bundle is misplaced against the corner. As per the double contingency principle, dropped/damaged fuel and a misplaced bundle are considered to be the same term and only the most limiting case is included in the kbias calculation.
The licensee considered the possibility of a non-conservative estimate of keff due to the lack of inserts along the rack periphery. The orientation of the inserts is chosen such that all interior storage cells will have an insert along every edge. This is not the case for storage cells along the south and west edges of the racks adjacent to the pool walls. These storage cells will only have insert material along two edges of the cell. To analyze the effects of missing inserts on the rack periphery, the licensee used a model as described in section 5.5.2 of attachment 7 to the LAR. The results of this analysis are shown in table 15 of attachment 7 to the LAR. The results show a small increase in keff as a result of missing inserts and is accounted for in the calculation of kbias.
The licensee considered a missing insert condition. This analysis is necessary due to the possibility of inadvertent removal of an insert during fuel assembly movement and insert removal for periodic inspections.
The licensee considered the effects of rack sliding. The licensee stated that the nominal model is infinite in extent with no inter-module gaps; therefore, all racks are close fitting and in the highest reactivity configuration. Therefore, sliding of the Boraflex racks at their interface is bounded by the infinite model of the individual rack designs.
Lastly, the licensee considered the effects of loss of spent fuel cooling. Loss of cooling would result in higher than normal moderator temperatures, reducing the reactivity of the system.
Therefore, the licensee stated that this abnormal condition is bounded by the normal configuration.
The NRC staff finds that the licensees evaluation of the normal and abnormal conditions considered all positive reactivity impacts and is performed consistent with RG 1.240. Therefore, the calculation of kbias is acceptable.
3.3.5 Disposition of Non-Conservatisms Several potential non-conservative assumptions were identified as part of the NRC staff review of the application. A potential non-conservatism is the inclusion of gadolinia loading in the NCS analysis. While the selected design basis lattice corresponds to the most reactive lattice, it does not correspond to the lowest gadolinia concentration. Despite this, the NRC staff finds the treatment of integral burnable absorbers acceptable because a sensitivity study for the selection of the design basis lattice was conducted and accounted for differences in gadolinia concentration.
Additionally, the NETCO-SNAP-IN rack inserts are only modeled as B10. This accounts for the neutron absorption capability of the inserts but does not account for the moderation capability.
The inserts are a homogenous mixture of boron carbide and aluminum. Carbon is a potential neutron moderator with a low absorption cross-section. This may correspond to an unanalyzed positive reactivity contribution. Despite this, the NRC staff finds the material modeling of the rack inserts acceptable based on the following discussion. Carbon is not nearly as effective as a moderator as hydrogen as the moderation capability of an element depends on the mass number of the element. Boron carbide accounts for 23 volume percent of the rack inserts, therefore carbon accounts for roughly 4 to 5 volume percent. There is significantly less carbon compared to hydrogen in a single storage cell and, therefore, the staff finds that the moderation capability of carbon is negligible compared to hydrogen in water.
Based on the above and given the significant margin of the kmaximum (95/95) to the limit, the NRC staff finds the potential non-conservatisms in the licensees NCS analysis acceptable.
Therefore, the NRC staff concludes that the proposed changes are performed consistent with RG 1.240 and therefore, are acceptable.
3.4 Evaluation of TS Changes 3.4.1 Changes to TS 4.3.1.1 Section 2.2 of this SE describes the proposed changes to TS 4.3.1.1.
The NRC finds the proposed changes to TS 4.3.1.1 are consistent with Grand Gulfs criticality analysis with the NETCO-SNAP-IN rack inserts installed. Therefore, the NRC staff finds that the proposed TS changes are acceptable, and 10 CFR 50.36(c)(4) will continue to be met.
3.4.2 Changes to TS 5.5 Section 2.2 of this SE describes the proposed changes to TS 5.5.
The NRC finds the proposed change to TS 5.5 is consistent with Technical Specifications Task Force (TSTF) Traveler TSTF-557, Revision 1 (Reference 13), and the licensees criticality analysis with the NETCO-SNAP-IN rack inserts installed. In addition, the NRC staff finds that the new TS 5.5.15 provides for monitoring the condition of the neutron absorber inserts.
Therefore, the NRC staff finds that the proposed change to TS 5.5 is acceptable, and 10 CFR 50.36(c)(5) will continue to be met.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Mississippi State official was notified of the proposed issuance of the amendment on November 17, 2025. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement 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 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, published in the Federal Register on March 25, 2024 (90 FR 13633), 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 amendment.
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 amendment will not be inimical to the common defense and security or to the health and safety of the public.
7.0 REFERENCES
1.
Couture, P. Entergy Operations, Inc., letter to NRC, Application to Revise Technical Specifications; Criticality Safety Analysis, Technical Specification 4.3.1, Criticality and Technical Specification 5.5.15, Spent Fuel Storage Rack Neutron Absorber Monitoring Program, dated November 25, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. (Attachments 1 through 6 to the Enclosure) and ML24330A235 (Attachment 7 to the Enclosure; not publicly available, proprietary information).
2.
Couture, P. Entergy Operations, Inc., letter to NRC, Supplement to Entergys Application to Revise Technical Specifications, Criticality Safety Analysis, Technical Specification 4.3.1, Criticality and Technical Specification 5.5.15, Spent Fuel Storage Rack Neutron Absorber Monitoring Program, dated December 16, 2024 (ML24351A148).
3.
Couture, P. Entergy Operations, Inc., letter to NRC, Supplement to Entergys Application to Revise Technical Specifications, Criticality Safety Analysis, Technical Specification 4.3.1, Criticality and Technical Specification 5.5.15, Spent Fuel Storage Rack Neutron Absorber Monitoring Program, dated May 5, 2025 (ML25126A111, letter; ML25126A112, not publicly available, proprietary information) 4.
Entergy Operations, Inc., Updated Final Safety Analysis Report, GGNS-USFAR (2024),
Nuclear Regulatory Commission (Package ML24341A171) 5.
Nuclear Energy Institute, NEI 16-03-A, Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools, Revision 1, May 2024 (ML24054A079).
6.
NRC, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, NUREG-0800, Section 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling, Revision 3, March 2007 (ML070570006).
7.
NRC, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, NUREG-0800, Section 9.1.2, New and Spent Fuel Storage, Revision 4, March 2007 (ML070550057).
8.
NRC, Fresh and Spent Fuel Pool Criticality Analyses, Regulatory Guide 1.240 March 2021 (ML20356A127).
9.
OBanion, M. W., NRC, letter to Vice President, Operations, Entergy Operations, Inc.
River Bend Station, Unit 1, Issuance of Amendment No. 201 Re: Change to the Neutron Absorbing Material Credited in Spent Fuel Pool for Criticality Control (EPID L-2018-LLA-0298), dated December 31, 2019 (ML19357A009).
10.
Nuclear Energy Institute, Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants, NEI 12-16, Revision 4, September 2019 (ML19269E069).
11.
Ennis, R. B., NRC, letter to Mr. M. J., Pacilio, Exelon Nuclear, Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Amendments Re: Use of Neutron Absorbing Inserts in Spent Fuel Pool Storage Racks (TAC Nos ME7538 and ME7539), dated May 21, 2013 (ML13114A929).
12.
NRC and Science Applications Internation Corporation, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, January 2001 (ML050250061).
13.
Technical Specifications Task Force, Transmittal of TSTF-557, Revision 1, Spent Fuel Storage Rack Neutron Absorber Monitoring Program, dated December 19, 2017 (ML17353A608).
Principal Contributors: J. Vande Polder, NRR K. Wood, NRR J. Wilson, NRR M. Mahoney, NRR Date: December 17, 2025
- via eConcurrence NRR-058 OFFICE NRR/DORL/LPL4/PM*
NRR/DORL/LPL4/LA*
NRR/DSS/SFNB/BC NAME MMahoney PBlechman SKrepel DATE 11/20/2025 12/1/2025 8/25/25 OFFICE NRR/DSS/STSB/BC*
NRR/DORL/LPL4/BC*
NRR/DORL/LPL4/PM*
NAME SMehta TNakanishi (SLee for)
MMahoney DATE 12/3/2025 12/17/2025 12/17/2025