Text

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
	ML24330A234
Person / Time
Site:	Grand Gulf
Issue date:	11/25/2024
From:	Couture P; Entergy Operations
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML24330A233	List: ML24330A234;
References
	GNRO2024-00033
	Download: ML24330A234 (1)
	v • d • e

Phil Couture Senior Manager Fleet Regulatory Assurance 601-368-5102 GNRO2024-00033 10 CFR 50.90 November 25, 2024 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

SUBJECT:

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 Grand Gulf Nuclear Station, Unit 1 Docket No. 50-416 License No. NPF-29 In accordance with the provisions of 10 CFR 50.59 and 10 CFR 50.90, Entergy Operations, Inc. (Entergy) is requesting an amendment to Grand Gulf Nuclear Station, Unit 1 (GGNS)

Technical Specifications (TS). The proposed amendment includes, 1) a revision to the criticality safety analysis for the spent fuel storage racks, 2) addition of requirements for the analysis for the fuel pool storage racks as contained in TS 4.3, Fuel Storage; Subpart 4.3.1, Criticality, and

3) the addition of requirements for monitoring of the neutron absorber material in the storage racks in TS 5.5, Programs and Manuals, new Subpart 5.5.15, Spent Fuel Storage Rack Neutron Absorber Monitoring Program.

This amendment is requested to change the neutron absorbing material to be credited for the purpose of criticality control in the spent fuel pool and upper containment pool.

includes a description of the change, no significant hazards consideration determination, and evaluation of environmental impact. Attachments to the enclosure include:, a copy of the marked-up TS pages and Attachment 2, a copy of the clean TS pages. Attachment 3 provides the Non-Proprietary version of the Global Nuclear Fuels (GNF)

- Americas Report NEDO-34125, Grand Gulf Nuclear Station: Fuel Storage Critically Analysis with Rack Inserts. Attachment 4, provides the NEI 12-16 Criticality Analysis Checklist. & 6, provides the GNF - Americas and Curtiss-Wright Nuclear Division (CWND)

Proprietary Information Affidavits of these companies, respectively. Attachment 7 is Proprietary in its entirety, as it contains information that is proprietary to GNF and CWND.

Accordingly, it is respectfully requested that the information proprietary to GNF and CWND be withheld from public disclosure in accordance with 10 CFR 2.390.

Entergy Operations, Inc., 1340 Echelon Parkway, Jackson, MS 39213 Proprietary Information - Withhold from Public Disclosure Under 10 CFR 2.390 The Balance of This Letter May Be Considered Non-Proprietary Upon Removal of Attachment 7 S} entergy

GNRO2024-00033 Page 2 of 2 The proposed change has been evaluated in accordance with 10 CFR 50.91(a)(1) using the criteria in 10 CFR 50.92(c) and it has been determined that the proposed change involves no significant hazards consideration.

Entergy requests review and approval of this license amendment request (LAR) to implement the proposed amendment to the TS by May 20, 2026. Once insert installation and LAR review and approval are complete, whichever date is later, the amendment will be implemented within 120 days.

This letter and its enclosure do not contain any new commitments.

Should you have any questions or require additional information, please contact me at (601) 368-5102.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on November 25, 2024.

Respectfully, Philip Couture Digitally signed by Philip Couture DN: cn=Philip Couture, c=US, o=Entergy, ou=Regulatory Assurance, email=pcoutur@entergy.com Date: 2024.11.25 07:49:07 -06'00' PC/ram : Evaluation of the Proposed Changes Attachments to the

Enclosure:

1. Technical Specification Pages - Marked-up

2. Technical Specification Pages - Clean

3. Global Nuclear Fuels Report NEDO-34125, Rev. 0, Dated July 2024, Grand Gulf Nuclear Station: Fuel Storage Critically Analysis with Rack Inserts (Non-Proprietary version)

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

5. Global Nuclear Fuels - Americas Proprietary Information Affidavits

6. Curtiss-Wright Nuclear Division Proprietary Information Affidavits

7. Global Nuclear Fuels Report NEDC-34125P, Grand Gulf Nuclear Station:

Fuel Storage Critically Analysis with Rack Inserts (Proprietary Version) cc:

NRC Region IV Regional Administrator NRC Senior Resident Inspector - Grand Gulf Nuclear Station NRC Project Manager - Grand Gulf Nuclear Station State Health Officer, Mississippi Department of Health GNRO2024-00033 Evaluation of Proposed Changes (26 pages below)

GNRO2024-00033 Page 1 of 26 Table of Contents Table of Contents.......................................................................................................................... 1

1.

SUMMARY

DESCRIPTION................................................................................................... 3

2.

DETAILED DESCRIPTION.................................................................................................... 3 2.1.

System Design and Operation........................................................................................ 3 2.2.

Current Technical Specifications Requirements............................................................. 4 2.3.

Reason for the Proposed Change.................................................................................. 5 2.4.

Description of the Proposed Change.............................................................................. 6

3.

TECHNICAL EVALUATION................................................................................................... 7 3.1.

Overview......................................................................................................................... 7 3.1.1 Boraflex Degradation.................................................................................................... 7 3.1.2 NETCO-SNAP-IN Rack Inserts Design Description................................................... 8 3.1.3 Demonstration of Proposed Method for Rack Insert Installation................................... 9 3.2.

Criticality....................................................................................................................... 10 3.2.1 Criticality Evaluation for NETCO-SNAP-IN Rack Inserts in GGNS SFP.................. 10 3.2.2 NEI 12-16.................................................................................................................... 11 3.3.

Materials....................................................................................................................... 11 3.3.1 Insert Boron-10 (B-10) Areal Density.......................................................................... 12 3.3.2 Corrosion.................................................................................................................... 13 3.3.3 NETCO-SNAP-IN Rack Insert Dimensions and Physical Properties....................... 13 3.4.

Mechanical................................................................................................................... 14 3.4.1 Fuel Assembly Clearances......................................................................................... 14 3.4.2 Mechanical Wear........................................................................................................ 14 3.4.3 Insertion / Retention Forces and Fuel Assembly Clearance....................................... 14 3.4.4 Stress Relaxation in the Absorber Rack Inserts......................................................... 16 3.5.

Seismic......................................................................................................................... 16 3.6.

Structural...................................................................................................................... 17 3.7.

Thermal-Hydraulic........................................................................................................ 18 3.8.

Accident Conditions...................................................................................................... 18 3.8.1 Accident Considerations Related to Criticality............................................................ 18 3.8.2 Fuel Handling Accident............................................................................................... 18 3.9.

Rack Insert Monitoring Program................................................................................... 19

GNRO2024-00033 Page 2 of 26 3.10.

Summary and Conclusions....................................................................................... 20

4.

REGULATORY EVALUATION............................................................................................ 20 4.1.

Applicable Regulatory Requirements/Criteria............................................................... 20 4.2.

Precedent..................................................................................................................... 20 4.3.

No Significant Hazards Considerations........................................................................ 21 4.4.

Conclusions.................................................................................................................. 24

5.

ENVIRONMENT CONSIDERATION................................................................................... 24

6.

REFERENCES.................................................................................................................... 25

GNRO2024-00033 Page 3 of 26

1.

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, Entergy Operations, Inc, (Entergy) requests an amendment to Facility Operating License No. NPF-29 for Grand Gulf Nuclear Station - Unit 1 (GGNS). The proposed change allows the crediting of NETCO-SNAP-IN neutron absorbing rack inserts in the criticality safety analysis (CSA) for the storage rack cells in the stations fuel building spent fuel storage facility; i.e., the spent fuel pool (SFP) and upper containment pool (UPC). This change is being requested due to the degradation of the Boraflex neutron absorbing material in the GGNS SFP. The change seeks approval of the aforementioned CSA. The change also seeks approval of changes to Technical Specifications (TS) concerning criticality design features of the spent fuel storage racks (TS 4.3.1.1), 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. Finally, the change seeks approval to add a program requirement that implements a monitoring program for the neutron absorbing rack inserts. The addition of this program requirement establishes consistency with Standardized Technical Specification improvement initiatives.

2. DETAILED DESCRIPTION 2.1. System Design and Operation GGNS Updated Final Safety Analysis Report (UFSAR) Section 9.1.2 documents the GGNS spent fuel storage safety design bases as summarized below.

Nuclear - The fuel array in the fully loaded spent fuel racks is designed to be subcritical by at least 5 percent k. Geometrically safe configurations of fuel stored in the spent fuel array are employed to assure the Keff does not exceed 0.95 under all normal and abnormal storage conditions.

Structural - The Unit 1 spent fuel storage racks in the auxiliary building and containment are designed to withstand all credible static and dynamic loadings to prevent damage to the structure of the racks, and therefore the contained fuel, and to minimize distortion of the racks arrangement. The spent fuel storage racks are categorized as safety Class 2 and seismic Category I.

The GGNS SFP 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 Figure 9.1-40a, 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. With current physical, load, and criticality restrictions only 3919 fuel storage locations are available in the spent fuel pool. Currently, in the Spent Fuel Pool, all assemblies are complete with no missing rods, spacers, or other parts. There is a failed fuel basket with 3 rods removed during reconstitution located in the H1 equipment rack. The assemblies which donated these rods had new rods installed.

The upper containment pool contains 7 high density fuel rack modules in 3 different module sizes. The module types are labeled E, F and G on UFSAR Figure 9.1-40a,

GNRO2024-00033 Page 4 of 26 which also shows their relative placement. There are a total of 710 fuel storage locations in the upper containment pool. With current physical and load restrictions only 584 fuel storage locations are available in the upper containment pool.

The spent fuel storage racks consist of individual cells with a 6-inch-square cross section, each of which accommodates a single BWR fuel assembly. The cell walls consist of a neutron absorber (Boraflex) sandwiched between sheets of stainless steel.

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. Fuel elements are limited by rack design to only being top loaded into a fuel storage rack and typical fuel assembly orientation (oriented vertically).

In order to accommodate known and possible future Boraflex degradation and maintain Keff criterion of less than or equal to 0.95, the GGNS 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 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 GGNS 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. This analysis is the basis, in part, for demonstrating compliance with plant TS requirements and U.S. Nuclear Regulatory Commission (NRC) regulations. The CSA methodology and inputs reflect the requirements of 10 CFR 50.68, 10 CFR 50 Appendix A General Design Criterion 62, NUREG-0800 Section 9.1.2 Rev. 3 dated July 1981, Generic Letter 78-11, and ANSI N210-1976. Information regarding the Boraflex and the method of its integration into the GGNS storage racks was provided in the stations response to Generic Letter 2016-01 (Reference 1) 2.2. Current Technical Specifications Requirements The GGNS TS requirements affected by this proposed change are TS Section 4.3.1 Criticality and TS Section 5.5, Programs and Manuals.

TS 4.3.1.1.a and 4.3.1.1.b identify requirements pertaining to the design of the spent fuel storage racks. Specifically, TS 4.3.1.1.a requires Keff 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1.2 of the UFSAR. TS 4.3.1.1.b requires a nominal fuel assembly center to center storage spacing of 6.26 inches in the storage racks.

GNRO2024-00033 Page 5 of 26 TS 4.3.1.1.e requires that Region II racks to be controlled as follows:

Storage cells with any Boraflex panel which has received a gamma dose in excess of 2.3E10 rads or which has a Boron-10 areal density less than 0.0165, which are designated within the Spent Fuel Pool Rack Boraflex Monitoring Program, are treated as Region II panels.

Storage cells face-adjacent to Region II panels are either restricted from fuel storage by physically blocking the isolated cells or are configured to meet, as a minimum (i.e., additional cells may be blocked), the Region II fuel storage configuration requirements.

When a 4x4 array of cell is classified as Region II and face-adjacent to another Region II 4x4 storage array, the new Region II 4x4 array is required to be blocked in the same 8-of-16 pattern and at the same orientation as the adjacent Region II 4x4 storage configuration.

TS Section 5.5, Programs and Manuals, does not contain requirements for a monitoring program for the neutron absorber used in the spent fuel storage racks.

2.3. Reason for the Proposed Change Entergy plans to install NETCO-SNAP-IN rack inserts in the GGNS SFP and UCP storage racks in accordance with the provisions of 10 CFR 50.59. This provides an alternative method of neutron absorption to meet the maximum Keff criticality control requirement without reliance on Boraflex, because the Boraflex has experienced degradation of its neutron absorbing capability as discussed in Reference 1. Entergy is requesting this license amendment to obtain approval for a new CSA that credits the use of the NETCO-SNAP-IN inserts and does not credit Boraflex. The new CSA methodology and inputs reflect the requirements and guidance of 10 CFR 50.68, 10 CFR 50 Appendix A General Design Criterion 62, NUREG-0800, Section 9.1.1 Rev 3 dated March 2007, Nuclear Energy Institute (NEI) 12-16 (Reference 2) and Nuclear Regulatory Commission (NRC) Interim Staff Guidance DSS-ISG-2010-01 (Reference 3).

With the crediting of the neutron absorbing rack inserts for criticality control, it is necessary to change GGNS TS 4.3.1.1 to specifically identify as design features for spent fuel storage the neutron absorbing inserts and fuel-related parameters used in the CSA, as well as remove the need for Region II racks. The proposed change to Section 4.3.1.1 will make the GGNS TS consistent with the Standard Technical Specifications for General Electric BWR/6 Plants, NUREG-1434, Rev 5 (Reference 4).

Finally, with the crediting of the neutron absorbing rack inserts for criticality control of the SFP, Entergy plans to implement a monitoring program consistent with NEI 16-03-A, Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools, Rev 0 (ADAMS ML17263A133) (Reference 5). NEI 16-03-A describes acceptable methods that may be used to monitor fixed neutron absorbers in SFPs to ensure that aging effects, corrosion, and other degradation mechanisms are identified and evaluated prior to loss of the required safety function. Since the GGNS TS do not currently contain any

GNRO2024-00033 Page 6 of 26 requirements regarding the monitoring of fixed neutron absorbers in its SFP, with the addition of the NETCO-SNAP-IN rack inserts into the SFP storage racks, Entergy seeks to establish a standardized TS program requirement that implements the aforementioned monitoring program. The proposed change is consistent with Technical Specifications Task Force (TSTF) traveler TSTF-557, Spent Fuel Storage Rack Neutron Absorber Monitoring Program, Rev 1 (ADAMS Accession ML17353A608) (Reference 6).

In Reference 13, Entergy submitted an application for renewal of the operating license for GGNS for an additional 20 years beyond the current expiration date (NRC Safety Evaluation Report is documented in Reference 14). The license renewal application (LRA) credited the Boraflex Monitoring Program, described in Section B.1.4, for managing aging of Boraflex during the period of extended operation. The Boraflex Monitoring Program will be replaced by a Neutron Absorbing Material Monitoring Program, consistent with the program described in NUREG-1801 (Reference 15),

Section XI.M40, Monitoring of Neutron-Absorbing Materials Other than Boraflex, and will follow the industry guidance in NEI 16-03-A (Reference 5) and the neutron absorbing material will be replaced so that the Boraflex material in the spent fuel pool will not be required to perform a neutron absorption function during the period of extended operation.

The proposed change does not apply to the new fuel storage racks. These storage racks do not contain any neutron absorbing material for criticality control and will not have the new NETCO-SNAP-IN rack inserts.

2.4. Description of the Proposed Change The proposed change consists of the following elements:

A new CSA for the GGNS SFP and UCP storage racks that credits the NETCO-SNAP-IN rack inserts for criticality control and does not credit Boraflex; A revision of TS 4.3.1.1.b to specifically identify the neutron absorber inserts as design features of the spent fuel storage racks; A revision of TS 4.3.1.1.c to specifically identify the updated fuel parameter (maximum k-infinity) used in the CSA crediting the NETCO-SNAP-IN rack inserts as design features of the spent fuel storage racks; The deletion of TS 4.3.1.1.e in its entirety to remove Region II as a design feature of the spent fuel storage racks; The addition of a new TS 5.5.15 to TS Section 5.5, Programs and Manuals, to incorporate a program into the TS to monitor the condition of the neutron absorber inserts used in the SFP and UCP storage racks to ensure they will continue to perform their design function.

The addition of TS 5.5.15 is consistent with TSTF-557, Rev. 1 (Reference 6).

GNRO2024-00033 Page 7 of 26 A markup of the proposed TS changes is provided in Attachment 1. The clean TS pages, incorporating these changes, are provided in Attachment 2. The UFSAR will also be revised, upon implementation of the approved amendment, as part of Entergys configuration control process.

3. TECHNICAL EVALUATION 3.1. Overview The following discussion will show that NETCO-SNAP-IN rack inserts are a safe and effective replacement for Boraflex to ensure continued compliance with TS requirements.

The proposed change will credit NETCO-SNAP-IN rack inserts for criticality control in individual SFP and UCP storage rack cells to ensure that the requirements of TS 4.3.1, Criticality, are maintained; specifically, The spent fuel storage racks are designed and shall be maintained with Keff 0.95 if fully flooded with unborated water The proposed change also includes changes to TS regarding design features and monitoring program requirements which are related to the analysis which credits these inserts.

The installation of the NETCO-SNAP-IN rack inserts is being controlled as a design change implemented under the provisions of 10 CFR 50.59 from a structural, seismic, and thermal-hydraulic perspective. As such, Entergy is not seeking NRC review and approval for installation of the inserts, only review and approval of the new CSA for crediting the inserts for criticality control in the GGNS SFP and UCP. Therefore, Sections 3.1.1 through 3.1.3, Sections 3.3 through 3.7, and Section 3.8.2 are provided for information only.

Entergy will not credit the neutron absorbing capability of the inserts for criticality control under the new methodology until and unless this proposed change is approved. The Boraflex material is contained within the GGNS spent fuel storage racks as part of their original fabrication and will remain in place and not be altered by installation of the NETCO-SNAP-IN rack inserts. The rack inserts installation began in the Fall of 2023 and is projected to be completed during the Summer of 2027.

3.1.1 Boraflex Degradation Boraflex is used in the GGNS SFP and UCP as a neutron-absorbing material and is credited in the CSA analysis of record (AOR) for the spent fuel storage racks. The condition of the Boraflex and the monitoring program used to measure changes in the material was documented in the stations response to Generic Letter 2016-01 (Reference 1). Consistent with the concern expressed in NRC Generic Letter 96-04, Boraflex Degradation in Spent Fuel Pool Storage Racks, the GGN monitoring program has identified degradation in the material, with an estimated areal density of 0.0184 g/cm2 in the peak Region I panel at the time the Generic Letter response was submitted.

While this is below the minimum certified Boraflex sheet areal density of 0.0190 g/cm2

GNRO2024-00033 Page 8 of 26 specified by Joseph Oats Corporation, the GGNS storage rack vendor, it remains above the credited areal density of 0.0133 g/cm2.

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. Region II storage locations are grouped in a minimum 4x4 arrays which shall have selected storage locations physically and administratively blocked in a 8 of 16 blocked configuration.

3.1.2 NETCO-SNAP-IN Rack Inserts Design Description This proposed change credits NETCO-SNAP-IN rack inserts for criticality control in SFP and UCP storage rack cells to ensure that the requirements of TS 4.3.1, Criticality, are maintained; specifically, The spent fuel storage racks are designed and shall be maintained with Keff 0.95 if fully flooded with unborated water The GGNS NETCO-SNAP-IN rack inserts will be fabricated from a homogeneous aluminum boron-carbide metal matrix material called BORALCAN (formerly called ALCAN), supplied by Rio Tinto Alcan. The NRC has approved this material for use in spent fuel racks at LaSalle County Station (LSCS), Peach Bottom Atomic Power Station, Units 2 & 3 (PBAPS), Quad Cities Nuclear Power Stations, Units 1 & 2 (QCNPS), River Bend Station (RBS), and Enrico Fermi Nuclear Power Plant, Unit 2 (References 7-11).

The NETCO-SNAP-IN rack inserts design that will be used at GGNS has been employed in the installation and successful operation of a combined total of over 24,000 NETCO-SNAP-IN inserts at these stations.

While the basic design of the GGNS inserts, and the material used in them, is the same as that used at LSCS, QCNPS, PBAPS, RBS, and Fermi, the GGNS inserts are fabricated from material with a B4C neutron absorber content of 23% by volume. The dimensions of the GGNS inserts are also slightly different because they are designed to fit into the GGNS SFP and UCP storage racks, as determined by the performance of confirmatory dimensional sizing measurements in the GGNS racks using non-borated test inserts of different wing widths and bend angles (see Section 3.4.3). A comparison of the insert dimensions and properties is provided in Section 3.3.3.

The NETCO-SNAP-IN rack insert is designed to become an integral part of the rack upon installation, and does not require any modification to the spent fuel storage rack.

The rack inserts slide into the rack and stay in place via friction with enough clearance still available for movement of fuel assemblies into and out of the storage cells. The insert is nominally the same length as a storage rack cell (approximately 169 inches),

thereby spanning the full length of the active fuel region of the fuel assembly when

GNRO2024-00033 Page 9 of 26 installed. Each GGNS insert is formed with a slightly greater than 90-degree bend angle, so that it is L-shaped (chevron shaped). This requires compression of the rack insert to install it into the spent fuel storage rack cell. After installation, the insert will conform to the 90-degree angle between adjacent spent fuel storage rack cell walls.

When installed, the insert sides (or wings) abut against the two adjacent faces of the spent fuel storage rack cell wall. The force exerted due to this deformation is determined by the material properties of the insert. The force between the wings of the insert and the spent fuel storage rack cell walls in conjunction with the static friction between these surfaces serves to retain the NETCO-SNAP-IN insert within the cell during normal fuel movement activities and under seismic events.

Entergy plans to install a NETCO-SNAP-IN insert with the same orientation in every usable (due to travel limitations of the fuel bridge and refueling bridge, certain periphery spent fuel storage rack cells are physically prevented from receiving a fuel bundle) spent fuel storage location within the GGNS SFP and UCP. Also, the H1 rack (SFP) and the J1 rack (UCP) are not receiving inserts. These racks are intended to store control rod blades, control rod guide tube, and/or defective fuel containers, and therefore, are not part of the normal fuel storage cell locations, and are prohibited by station procedure for use as fuel storage locations. Installation of a NETCO-SNAP-IN insert in every usable storage location and with the same orientation ensures that neutron absorption and criticality control by the rack inserts is uniform across the SFP and UCP. A criticality analysis crediting the NETCO-SNAP-IN inserts has been performed for the GGNS SFP and UCP to support this design change. This analysis is discussed in Section 3.2.

The NETCO-SNAP-IN inserts designed for GGNS spent fuel storage racks are fabricated with the top (approximately 4 inches by 1 inch) of the insert bent edges removed or coped, to reduce any potential interference between an insert and a fuel channel spacer. The required NETCO-SNAP-IN insert orientation during installation, insert coping, and the administratively required fuel bundle orientation, reduce the potential for fuel bundle / insert interference.

3.1.3 Demonstration of Proposed Method for Rack Insert Installation To verify the mechanical compatibility of the NETCO-SNAP-IN insert with the GGNS SFP and UCP storage racks and compatibility of the fuel stored therein, an insert demonstration program (i.e., the prototype installation and testing program) was performed at GGNS in October 2023. The mechanical feasibility of using NETCO-SNAP-IN inserts at GGNS was verified by installing fifty-four (54) prototype inserts into randomly selected storage cells within the SFP and by installing two (2) prototype inserts into randomly selected storage cells within the UCP. After installation, retention load testing was performed on all fifty-six (56) of the prototype inserts using the insert removal tool. Additionally, 29 of the SFP storage cells and 2 of the UCP storage cells, containing prototype inserts, were tested using a dummy fuel assembly, which has a cross-sectional dimension of a channeled fuel assembly, to verify adequate dimensional

GNRO2024-00033 Page 10 of 26 clearances between the insert and a fuel assembly during fuel handling. The NETCO-SNAP-IN rack inserts used in the GGNS prototype program were designed, fabricated, tested, and inspected under the NETCO quality assurance program to ensure they meet the design requirements for permanent inserts. In summary, the key insert parameters validated during the demonstration program were: 1) insertion installation success; 2) lack of fuel interference; and 3) retention force (i.e. greater than 150 lbf). These parameters are discussed in further detail below in Section 3.4.3, Insertion / Retention Forces and Fuel Assembly Clearance.

3.2. Criticality 3.2.1 Criticality Evaluation for NETCO-SNAP-IN Rack Inserts in GGNS SFP In accordance with the requirements of 10 CFR 50.68, a CSA was performed to support the storage of spent fuel in the GGNS SFP and UCP with credit for the NETCO-SNAP-IN rack inserts installed. All necessary requirements as outlined in NUREG-0800, Section 9.1.1 Rev 3 March 2007, have been met. Nuclear Energy Institutes (NEI) NEI 12-16, Rev 4 (Reference 2) was used as a guidance document for this analysis. The analysis, described in Attachment 7, demonstrates that the maximum Keff (kmax(95/95)) is substantially less than the 10 CFR 50.68 limit of 0.95 for normal and credible abnormal operation with tolerances and computational uncertainties taken into account. The analysis assumptions included:

Uniform pool storage configuration with all usable fuel storage locations loaded with a NETCO-SNAP-IN insert in the same orientation and a fuel bundle with the highest rack efficiency; A NETCO-SNAP-IN insert Boron-10 (B-10) areal density of 0.0139 g B10/cm2 (which is less than the minimum certified areal density of 0.0141 g B10 /cm2) to account for potential manufacturing uncertainties; No credit for neutron absorption by the Boraflex material installed between the SFP storage rack cells, which has been modeled as water; and, The SFP fully flooded with unborated water.

The CSA covers all legacy fuel in storage at GGNS and the current fuel product line in use at GGNS, GNF3. The description of these product lines is provided in Section 4.0 of, while the disposition for all legacy fuel is provided in Appendix B of.

The reactivity of the GGNS SFP storage rack containing NETCO-SNAP-IN inserts was calculated using the computer codes TGBLA06 and MCNP-05P. In this evaluation, in-core k and exposure dependent, pin-by-pin isotopic specifications were generated using TGBLA06, the NRC-approved Global Nuclear Fuel (GNF) BWR lattice physics code.

The fuel storage criticality calculations were then perform using MCNP-05P, the GNF proprietary version of the Las Alamos National Laboratory Monte Carlo neutron transport code MCNP5. TGBLA06 uses ENDF/B-V cross-section data to perform coarse-mesh,

GNRO2024-00033 Page 11 of 26 broad-group, diffusion theory calculations. MCNP-05P used ENDF/B-VII.0 point-wise (i.e continuous) cross-section data, and all reactions in the cross-section evaluation are considered. MCNP-05P has been validated and verified for spent fuel pool storage rack evaluations in accordance with the NUREG/CR-6698 guidance (included as part of ). The Method of Analysis is discussed in greater detail in Section 3.0 of. Validation of the codes and libraries is described in Section 3.4 and Appendix A of Attachment 7.

The use of TGBLA06 (Reference 16) for BWR core depletion calculations has been reviewed and accepted by the NRC as part of the approval of Reference 17. The NRC has also approved the MCNP/TGBLA06 code package for use in similar fuel pool criticality analyses, as documented in Reference 18. Finally, the NRC has approved use of these codes in the criticality analysis for previous applications of the NETCO-SNAP-IN inserts in the PBAPS spent fuel pools, as documented in Reference 8. For the analysis of the GGNS SFP, TGBLA06 was used in the manner allowed by the NRC approvals (Reference 16, 17, 18). In addition to the request for approval of Attachment 7, which credits the NETCO-SNAP-IN inserts for criticality control in the GGNS SFP, there are two other related elements of the proposed change:

A maximum cold, uncontrolled peak in-core k-infinity of 1.29 was set as the limit for the analysis. In the proposed TS 4.3.1.1.c, this value is incorporated into the GGNS Design Features section on spent fuel storage criticality, consistent with Reference 4.

In the proposed TS 4.3.1.1.b, the description of the neutron absorber inserts within the spent fuel storage racks is incorporated into the GGNS Design Features section on spent fuel storage criticality, consistent with Reference 4.

3.2.2 NEI 12-16 NEI 12-16 (Reference 2) was used as the guidance documents for this analysis.

Guidance pertaining to soluble boron in the SFP is not applicable because GGNS is a BWR plant and has no soluble boron in the SFP. Attachment 4 includes the Criticality Analysis Checklist from NEI 12-16 to identify the areas of the analysis that conform or do not conform to the guidance in NEI 12-16.

3.3. Materials The NETCO-SNAP-IN Rio Tinto Alcan composite rack inserts must ensure that the neutron absorber remains in place over the lifetime of the SFP and UCP storage racks during normal operation and abnormal events. Reference 12 provides a detailed evaluation of the Rio Tinto Alcan composite material. This report demonstrates that the material is suitable as a neutron absorber to maintain the SFP and UCP within design and regulatory limits over the life of the SFP and UCP storage racks. Qualification testing has been performed to confirm its acceptability and the monitoring program

GNRO2024-00033 Page 12 of 26 discussed in Section 3.9 will confirm its continued acceptability to perform its required design function in the GGNS SFP and UCP.

The production process for manufacturing the rack inserts is described in detail in Reference 12. The technique developed by Rio Tinto Alcan to produce the aluminum/boron carbide metal matrix composite results in a homogeneous distribution of the B4C in a rolled sheet, which is trimmed to produce rack insert blanks. Insert flats are then cut from the blanks and bent on a press brake to an angle somewhat larger than 90° to provide the chevron shaped insert and the long edges of the insert are roll formed to establish the winglets. Additionally, test coupons are cut from each of the blanks and used to confirm acceptable minimum areal density and material properties.

3.3.1 Insert Boron-10 (B-10) Areal Density The insert manufacturing quality assurance testing lower limit for the areal density of boron in the Rio Tinto Alcan composite is given in terms of B-10, and is 0.0141 g B10 /cm2for GGNS. Verification of the minimum certified areal density of B-10 in the rack inserts (i.e., pre-characterization) is performed for 100 percent of the material used for the inserts. Each blank (from which the insert flats are cut) will have a traceable test coupon removed and subjected to neutron attenuation testing.

For each coupon, a specific areal density value is obtained, to which a 3-sigma (99.7%)

uncertainty is applied, to confirm that the measured areal density exceeds the minimum certified areal density before the corresponding inserts are accepted. Given 100 percent sampling and the 3-sigma uncertainty applied to the measurement, GGNS is assured that none of the inserts have an areal density below the minimum certified value. The CSA, discussed in Section 3.2.1, assumes an insert B-10 areal density of 0.0139 g B10 /cm2, which is significantly less than the minimum certified areal density of 0.0141 g B10 /cm2.

Reference 12, Section 3.4 (Table 3.1), refers to a B-10 areal density limit of 0.0087 g B10 /cm2for the quality assurance test program. This value is for the NETCO-SNAP-IN rack inserts manufactured for LSCS. All of the NETCO-SNAP-IN rack inserts manufactured for a particular user have the same minimum certified B-10 areal density, but that value may be different user-to-user. The 0.0087 g B10 /cm2 is an example value used in the NETCO material qualification report and is not the minimum certified B-10 areal density in all NETCO-SNAP-IN rack inserts for all customers. The B-10 areal density in the inserts for a given plant is customized for each users needs based on the criticality analysis and rack design. Each user specifies the minimum certified B-10 areal density for their plants inserts in the procurement specification. For GGNS, the minimum certified manufactured B-10 areal density is 0.0141 g B10 /cm2.

Verification of the areal density of B-10 over the lifetime of the racks will be performed through the rack insert monitoring program discussed in Section 3.9.

GNRO2024-00033 Page 13 of 26 3.3.2 Corrosion Resistance to material loss, pitting, cracking, and blistering is important to ensuring that the B-10 will not be lost, and that distortion of the rack insert will not interfere with fuel movement. Therefore, an accelerated corrosion test program was performed to determine the susceptibility of the Rio Tinto Alcan composite to general (i.e. uniform) and localized (i.e. pitting) corrosion in BWR SFPs. This program is described in detail in Section 5.0 of Reference 12. The material qualification program included material at 16 volume percent and 25 volume percent loadings of boron carbide (B4C). This range of as-tested boron carbide loadings of the test coupons bounds the loading to be used at GGNS (23 volume percent B4C).

In summary, the material qualification test program concluded that the A1100 aluminum boron carbide composite produced by Rio Tinto Alcan is a highly suitable neutron absorber for use in spent fuel storage racks. The program determined that general corrosion of the material would occur at an extremely low rate (approximately 0.02 mils/year); no local corrosion (pitting) or cracking was detected; and there was no measurable change in the B-10 areal density. The program also determined, through a review of pertinent literature, that the aluminum alloy used to make the inserts is not susceptible to stress corrosion cracking (SCC). Verification that unexpected material degradation is not occurring, over the lifetime of the racks, will be performed through the rack insert monitoring program discussed in Section 3.9.

3.3.3 NETCO-SNAP-IN Rack Insert Dimensions and Physical Properties The NETCO-SNAP-IN rack inserts to be used in the GGNS spent fuel storage pools are dimensionally and physically similar to those already in use at other BWR stations -

RBS, LSCS, PBAPS, FERMI, and QCNPS, as shown in Table 3.3-1 Table 1: Insert Dimension/Property Comparison Property GGNS RBS FERMI LSCS PBAPS QCNPS Length (in.)

169 169 175 167.75 169 Style 1 -

165.25 Style 2 -

165.00 Thickness (in.)

0.080 0.080 Proprietary 0.065 0.075 0.085 B-10 Min Areal Density (g B10 /cm2) 0.0141 0.0129 0.0157 0.0087 0.0105 0.0116 B4C Density (vol %)

23 21 23 17 19 17

GNRO2024-00033 Page 14 of 26 3.4. Mechanical 3.4.1 Fuel Assembly Clearances Placement of the rack insert in a SFP or UCP storage rack cell slightly reduces the cell inside dimension available for fuel assembly insertion. The prototype installation and testing program (Sections 3.1.3 and 3.4.3) confirmed adequate clearance between a fuel assembly and rack cells containing prototype inserts by inserting and removing a dummy fuel bundle that is dimensionally the same as a channeled fuel assembly.

The NETCO-SNAP-IN inserts designed for GGNS spent fuel storage racks are fabricated with the top (approximately 4 inches by 1 inch) of the insert bent edges removed or coped, to reduce any potential interference between an insert and a fuel channel spacer. The required NETCO-SNAP-IN insert orientation during installation, insert coping, and the administratively required fuel bundle orientation, reduce the potential for fuel bundle / insert interference.

If there is unexpected warping or bowing of the rack insert after installation that reduces the fuel assembly-to-spent fuel storage rack insert clearance, then the fuel handler would notice increased force indicated on the hoist load cell when attempting to raise (i.e., remove) an assembly. If the rack insert would inadvertently come out of a spent fuel storage rack cell with an assembly, this condition is bounded by the missing rack insert evaluation in the criticality analysis (see Section 5.5.2 of Attachment 7).

If a channeled spent fuel assembly cannot fit into the spent fuel storage rack cells containing rack inserts due to mechanical clearances, the fuel assembly may be de-channeled and stored. The new criticality analysis demonstrates that this is a conservative configuration compared to storing fuel assemblies with the channel (see Section 5.4.2 of Attachment 7) 3.4.2 Mechanical Wear Minimal insert material wear is expected within the active fuel region due to adequate clearance between the fuel assembly and rack insert. The clearance between the fuel and insert has been verified using a dummy fuel assembly, as part of the prototype testing (see Sections 3.1.3 and 3.4.3). The combined effects of adequate clearance and infrequent fuel assembly movement will preclude significant wear of the rack insert.

3.4.3 Insertion / Retention Forces and Fuel Assembly Clearance Dimensional Sizing Testing Past experience from installing the NETCO-SNAP-IN inserts in other spent fuel storage racks has shown that the manufactured dimensions for the rack cells do not always

GNRO2024-00033 Page 15 of 26 match the tolerances shown on design drawings. Because the NETCO-SNAP-IN insert relies heavily on the spring force of the insert obtained when compressing the insert into the cell, even small deviations of the cell dimensions can have a large impact on how an insert fits into a rack cell. In order to determine the optimal wing width and initial bend angle needed for an insert to successfully fit into the GGNS spent fuel storage racks, test inserts made from non-borated, 3000 series aluminum were installed into and removed from fifty-seven (57) randomly selected storage cells within the GGNS SFP and seven (7) randomly selected storage cells within the GGNS UCP in February 2023. The main purpose of these test installations was to provide a basis for determining the appropriate size of the wing width and initial bend angle needed for the final insert design that will be installed in the GGNS SFP and UCP. Load tests were also performed during the removal of these test inserts to determine the force required to remove the insert. Due to slight differences in mechanical properties of the materials, the load test results for the aluminum test inserts were not expected to be identical to those of the inserts made from BORALCAN. However, the results were useful as a guide to ensure the final design of the absorber inserts will provide the minimum force required for insert removal.

Prototype Installation and Testing A demonstration program using prototype NETCO-SNAP-IN rack insert was completed at GGNS in October 2023, as described in Section 3.1.3 above. The prototype installation and testing provided a confirmation that BORALCAN inserts, made to the final design, meet the interference and retention load testing requirements. The GGNS specific parameters observed during the demonstration program were: (1) installation force; (2) retention force (greater than 150 lbs.); and (3) fuel assembly clearance.

Additional detail is provided below.

Insertion Force - The insertion or installation force is produced by the installation tool, through the use of an impact mechanism at the top of the tool and the weight of the tool itself. The combined weight of the installation tool and insert is less than 1000 pounds to maintain a load under the hoist limit for the refueling bridge auxiliary hoist. It is also less than the heavy load limit of GGNS of 1140 pounds. Some of the installation tool weight is due to the external frame that is part of the tool design that helps to guide the insert into place, and therefore the full weight of the tool is not applied to seat the insert. Most of the time, the weight provided is sufficient. But in some instances, the insert may stop just before it is fully seated into the storage rack cell. In those cases, a separate insert setting tool, which does not have the external frame of the installation tool, is used to provide additional force to fully seat the insert the last few inches. The yield stress of the aluminum-boron carbide composite material is less than the yield stress of the SFP and UCP storage rack material (i.e., stainless steel); therefore, the applied stress on the SFP and UCP storage rack is significantly less than the allowable stress for the stainless steel SFP and UCP storage racks and will not damage the existing racks.

GNRO2024-00033 Page 16 of 26 Retention Force - Acceptance testing was performed to measure the force required to remove an insert from a fuel storage rack cell once installed (i.e., the retention force).

The minimum acceptable force was 150 lbf, which meets the GGNS specific design criteria for seismic accelerations and stress relaxation (see Section 3.4.4 below). It also provides a significant margin in retention force to reduce the possibility that the insert will move during normal fuel movement operations due to drag force, if the fuel were to contact the insert during removal from a storage cell.

Fuel Assembly Clearance - During the prototype installation and testing program, a dummy fuel assembly was inserted and then removed from 29 SFP test locations in which a prototype insert was installed, with no indication of clearance issues. For the UCP, 2 test locations were tested with no indication of clearance issues. The dummy fuel assembly used has a cross-sectional dimension of a channeled fuel assembly. This testing was performed to confirm that the installed inserts would not interfere with fuel movement.

In summary, the results of the prototype installation and testing program demonstrated the mechanical compatibility with the fuel stored therein. The results provide reasonable assurance that NETCO-SNAP-IN inserts will perform their intended safety function when installed in the GGNS SFP and UCP.

3.4.4 Stress Relaxation in the Absorber Rack Inserts During installation, the NETCO-SNAP-IN rack inserts are compressed from an initial bend angle of greater than 90 degrees to fit in the square dimensions of the spent fuel storage rack cell interior. Once installed, the internal stresses in the rack inserts may be susceptible to relaxation over time. This relaxation would result in less force against the spent fuel storage rack cell wall and lower retention force. An analysis of stress relaxation in aluminum alloys has been performed to establish the expected performance of the rack inserts in this regard (Reference 12).

The GGNS insert design has an assumption of approximately 50% stress relaxation during the course of its service life (Reference 12). This assumption is conservative due to the reinforcing properties of the boron carbide particles. This assumption was used to determine the minimum retention force requirements of the inserts during installation, discussed in Section 3.4.3, that would hold the inserts in place during a seismic event even after relaxation has occurred.

3.5. Seismic A reconciliation of the seismic AOR for GGNS was performed to demonstrate that the conclusions developed in the original analysis remain valid with the inserts installed in the GGNS fuel storage racks. The reconciliation considered the added mass and the effect on the natural frequency of the fuel storage racks due to the addition of the

GNRO2024-00033 Page 17 of 26 inserts. The reconciliation evaluation determined the effect of the additional mass of the inserts (17 lbm each) on the maximum displacements and stress factors due to the slightly greater kinetic energy for the same seismic inputs. The assessment determined that the changes to displacement are proportional to the percent increase in kinetic energy as a result of the addition of the inserts and would not produce displacements that exceed the gaps between storage racks or between the racks and walls. It was also shown that a proportional increase in the value of each stress factor equal to the percent increase in kinetic energy does not significantly reduce the available margin reported in the AOR for the calculated stresses based on a time history analysis.

For all conditions, it was concluded that the allowable limits were not exceeded as a result of the addition of the inserts. Finally, the concern that an insert may slide upwards out of the rack cell during a seismic event is precluded by the low seismic g-level in the vertical direction, and the total retention friction force between the insert and the cell wall. The prototype installation and testing program confirmed that sufficient retention force exists to prevent the insert from moving upward during a seismic event (see Section 3.4.3).

3.6. Structural A reconciliation of the structural AOR for the GGNS spent fuel storage racks was performed to demonstrate that the conclusions developed in the analysis remain valid with the NETCO-SNAP-IN inserts installed. The margins of safety calculated in the AOR were used as a basis for reconciliation. Each of the components analyzed in the AOR were evaluated for changes in the margin of safety, to determine if the addition of the inserts will significantly change the stresses under normal and seismic conditions. The evaluated components included:

The fuel storage cell assemblies; The fuel storage rack support assemblies; The spent fuel pool structure was evaluated for the additional mass added by the installation of the NETCO-SNAP-IN inserts.

The upper containment pool structure was evaluated for the additional mass added by the installation of the NETCO-SNAP-IN inserts.

The evaluation determined that the changes in margins of safety for each component due to the addition of the inserts were not significant. Therefore, it is concluded that the addition of the inserts does not cause an impact that would compromise the structural integrity of the fuel storage racks or the storage pools.

The structural performance of the NETCO-SNAP-IN inserts under GGNS design conditions were also evaluated. The objective of this evaluation was to confirm that the neutron absorber inserts will continue to perform their safety function under the required

GNRO2024-00033 Page 18 of 26 loading conditions. It was concluded that in the installed condition at GGNS, stresses on the inserts will be significantly less than the material yield strength and therefore the inserts will not deform plastically. Additionally, it was concluded that no significant stresses will be produced as a result of thermal expansion.

3.7. Thermal-Hydraulic A reconciliation of the thermal-hydraulic AOR for the GGNS spent fuel storage racks was performed to demonstrate that the conclusions developed in the analysis remain valid with the NETCO-SNAP-IN inserts installed. Changes in the fuel storage cell geometry due to the addition of the inserts were evaluated. The effects of these changes on the thermal-hydraulic analysis were then determined and it was concluded that the addition of the NETCO-SNAP-IN inserts will not adversely affect the existing thermal-hydraulic analysis.

3.8. Accident Conditions 3.8.1 Accident Considerations Related to Criticality As part of the criticality analysis discussed in Section 3.2 and described in Attachment 7, the spent fuel rack configuration was analyzed for credible accident scenarios. The scenarios analyzed are listed below and are discussed in Section 5 of Attachment 7.

Dropped / damaged fuel Abnormal positioning of a fuel assembly outside the fuel storage rack Misplacement of fuel bundles in unpoisoned equipment racks next to the fuel racks In addition, the following scenarios were considered bounded by the analysis, with the justification provided in Section 5.5.3 of Attachment 7.

Dropped fuel assembly on rack Closure of water gap between racks caused by rack sliding due to seismic event Loss of spent fuel cooling The analysis, described in Attachment 7, demonstrates that the maximum k-effective (kmax(95/95)) is less than the 10 CFR 50.68 limit of 0.95 for normal and credible abnormal operation with tolerances and computational uncertainties taken into account.

3.8.2 Fuel Handling Accident A reconciliation review of the fuel drop AOR was performed to verify that the spent fuel storage racks with NETCO-SNAP-IN inserts will continue to accommodate the fuel handling uplift load and impact loadings resulting from the analyzed fuel assembly drop accidents. The evaluation of the Fuel Handling Accident analysis determined that there is no adverse impact on the analysis due to the presence of the inserts.

GNRO2024-00033 Page 19 of 26 The evaluation of a fuel handling uplift load with the inserts installed concluded that addition of the inserts does not impact this analysis. Additionally, insert and insert tool drop accidents were evaluated including (a) the straight drop of an insert and insert tool onto the top of a rack; (b) an inclined drop onto the top of a rack; and (c) a straight drop through the cell to the bottom of the rack. For all cases, the review concluded that the accidental drop of the inserts and insert tool would not adversely affect the results of the AOR.

3.9. Rack Insert Monitoring Program GGNS is committed to the monitoring program for the spent fuel storage racks Boraflex panels described in Section B.1.4 of the UFSAR supplement of the GGNS license renewal application (Reference 13). The current monitoring program is consistent with the NRC-recommended program described in NUREG-1801,Section XI.M22, Boraflex Monitoring (Reference 15).

GGNS will have an updated monitoring program for the spent fuel storage rack neutron absorbing inserts for Section B.1.4 of the UFSAR supplement of the GGNS license renewal application. The program will be consistent with the NRC-recommended program described in NUREG-1801, Revision 2,Section XI.M40, Monitoring of Neutron-Absorbing Materials Other than Boraflex. Upon issuance of the GGNS renewed operating license, the program will become part of the GGNS UFSAR and the licensing basis.

The program will use monitoring coupons and in-situ inspections and will follow the most current industry guidance (Reference 5). Degradation of the neutron absorbing material that could compromise the criticality analysis will be detected to assure that the required 5% sub-criticality margin is maintained during the period of extended operation. The parameters monitored include the physical condition and dimensions (e.g., corrosion, pitting, wear, blisters, and bulges) and areal density (neutron absorber loss). Inspection and test frequencies will be based on plant-specific experience and will be informed by industry operating experience, but will be at least once every 10 years. Test results will be trended and, if necessary, corrective action will be taken to ensure the subcriticality margin is maintained.

Since the GGNS TS do not contain any requirements regarding the monitoring of fixed neutron absorbers in its SFP (and UCP), with the addition of the NETCO-SNAP-IN rack inserts into the SFP and UCP storage racks, Entergy seeks to establish a standardized TS program requirement that implements the aforementioned monitoring program. The proposed change, the addition of TS 5.5.15, is consistent with Reference

6.

GNRO2024-00033 Page 20 of 26 3.10.

Summary and Conclusions The proposed change to credit the NETCO-SNAP-IN rack inserts in the SFP and UCP storage racks for criticality control has been evaluated and shown to be a safe and effective manner in which to resolve the Boraflex degradation issue for the remaining period of time that spent fuel needs to be stored in the GGNS SFP storage racks, ensuring that the plants safety design bases for the SFP continue to be maintained.

Furthermore, the proposed change establishes consistency with Standardized Technical Specification Improvement initiatives and the updated monitoring program satisfies the commitment Entergy made to implement a Neutron Absorbing Material Monitoring Program for license renewal for GGNS.

4. REGULATORY EVALUATION 4.1. Applicable Regulatory Requirements/Criteria 10 CFR 50.68, Criticality accident requirements, paragraph (b)(4) states that the k-eff of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. The GGNS SFP CSA crediting the neutron absorbing rack inserts provided as Attachment 7 to this submittal, demonstrates that this requirement is met.

Paragraph (b)(7) of 10 CFR 50.68 states that the maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to 5.0 percent by weight. The aforementioned CSA assumes a maximum of 4.9 percent by weight of U-235 enrichment for current and future fuel used at GGNS and TS 4.3.1.1.d meets this requirement.

General Design Criteria (GDC) 62 Prevention of criticality in fuel storage and handling states that criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.

The evaluation of conformance with GDC 62 is discussed in Section 0.1.2, Spent Fuel Storage, of the GGNS UFSAR. The NETCO-SNAP-IN rack insert CSA has been performed to demonstrate that Keff will remain less than or equal to 0.95 with no credit taken for the Boraflex neutron poison material in the spent fuel storage racks in the final configuration.

4.2. Precedent The NRC has approved the use of NETCO-SNAP-IN rack inserts as an alternative method of criticality control to address the Boraflex degradation for five other plants as documented in References 7-11. If the proposed change is approved, GGNS would become the sixth boiling water reactor (BWR) nuclear station to credit use of NETCO-SNAP-IN rack inserts for criticality control in the SFP.

GNRO2024-00033 Page 21 of 26 Additionally, the NRC has approved NEI 16-03-A (Reference 5) concerning guidance for monitoring of fixed neutron absorbers in spent fuel storage pools. The requested change to add a new program to the GGNS TS for monitoring of the neutron absorbing rack inserts is consistent with Reference 5. It is also consistent with Reference 6.

4.3. No Significant Hazards Considerations In accordance with 10 CFR 50.90, Entergy Operations, Inc (Entergy) requests an amendment to Facility Operation License (NPF-29) for Grand Gulf Nuclear Power Station (GGNS) - Unit 1. The proposed change requests NRC approval for:

The crediting of NETCO-SNAP-IN neutron absorbing rack inserts in the criticality safety analysis (CSA) for the storage rack cells in the stations fuel building spent fuel storage facility; i.e., the spent fuel pool (SFP) and the stations containment building spent fuel storage facility; i.e. the upper containment pool (UCP). This change is being requested due to the degradation of the Boraflex neutron absorbing material currently being used in the GGNS SFP and UCP.

Changes to the Technical Specifications (TS) concerning criticality design features of the spent fuel storage racks (TS 4.3.1.1), to specifically identify the neutron absorbing inserts and fuel-related parameters used in the CSA, consistent with Standard Technical Specifications (NUREG-1434).

Changes to the Technical Specifications (TS) to remove Region II requirements of the spent fuel storage racks (TS 4.3.1.1), consistent with Standard Technical Specifications (NUREG-1434).

The addition of a TS program requirement (TS 5.5.15) that implements a monitoring program for the neutron absorbing rack inserts. The addition of this program requirement establishes consistency with a Standardized Technical Specification Improvement initiative (TSTF-557, Rev 1).

According to 10 CFR 50.92, a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3. Involve a significant reduction in a margin of safety Entergy has evaluated the proposed change for GGNS using the criteria in 10 CFR 50.92, and has determined that the proposed change does not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.

GNRO2024-00033 Page 22 of 26 Criteria

1. Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change involves a new CSA for the GGNS SFP and UCP to credit the neutron absorbing capability of the NETCO-SNAP-IN rack inserts installed in the SFP and UCP storage rack cells for criticality control. The neutron absorbing capability of the Boraflex material contained in the SFP and UCP storage racks would no longer be credited. The new CSA is not a physical change to the plant and does not affect the ability of any structures, systems or components (SSCs) to perform a design function. The proposed new CSA demonstrates adequate margin to criticality for spent fuel storage rack cells and therefore does not affect the consequences of any accident previously evaluated.

The proposed change also involves changes to the requirements specified in TS 4.3.1.1 for spent fuel storage racks. These changes are consistent with the new CSA and impose additional requirements in the plants Technical Specifications.

These new requirements for the spent fuel storage racks do not involve a physical change to any plant systems and do not affect the ability of any SSCs to perform a design function. The new requirements support the assumptions of the new CSA and therefore do not affect the consequences of any accident previously evaluated.

Finally, the proposed change involves the addition of a new programmatic requirement in TS 5.5 to perform monitoring of the NETCO-SNAP-IN rack inserts to ensure that they continue to perform their design function, consistent with the assumptions of the new CSA. Monitoring of the SFP Neutron absorber does not affect the ability of any SSCs to perform a design function. A SFP storage rack neutron absorber monitoring program is not an initiator to any accident previously evaluated and does not affect the consequences of any accident previously evaluated.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Will operation of the facility in accordance with this proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No Onsite storage of spent fuel assemblies in the GGNS spent fuel pool is a normal activity for which GGNS has been designed and licensed. The new CSA does not involve any physical changes to the plant and does not change the method of spent fuel movement or storage. It only provides an analysis of the existing SFP and UCP storage racks, with credit for the NETCO-SNAP-IN rack inserts, to demonstrate adequate margin to criticality.

GNRO2024-00033 Page 23 of 26 Similarly, the addition of new requirements in TS 4.3.1.1 for the spent fuel storage racks, and the removal of Region I / Region II requirements, and a requirement in TS 5.5 for a new storage rack neutron absorber monitoring program does not involve any physical changes to the plant and does not change the method of spent fuel movement or storage.

Based on the above information, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?

Response: No The safety margin which is relevant to the proposed change is the safety margin for criticality in spent fuel storage racks. This margin is 5% (i.e., Keff less than or equal to 0.95 when fully flooded with unborated water), including a conservative margin to account for engineering and manufacturing uncertainties. The new CSA demonstrates that this margin is maintained when the NETCO-SNAP-IN rack inserts are credited for criticality control in the GGNS SFP and UCP, without credit for Boraflex.

The safety margin is unaffected by the addition of new requirements in TS 4.3.1.1 for the spent fuel storage racks. The new requirements are consistent with the assumptions of the new CSA and therefore support the basis of the safety margin demonstrated in the CSA.

The safety margin is unaffected by the removal of Region I / Region II requirements from TS 4.3.1.1 for the spent fuel storage racks. The new requirements are consistent with the assumptions of the new CSA and therefore support the basis of the safety margin demonstrated in the CSA.

The addition of a new programmatic requirement in TS 5.5 to perform monitoring of the SFP neutron absorber inserts does not affect the margin to safety for criticality.

Performance of monitoring in accordance with this new requirement will support the criticality safety margin as it provides assurance that the inserts continue to perform their assumed design function which is credited in the new CSA.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above evaluation, Entergy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of no significant hazards considerations is justified.

GNRO2024-00033 Page 24 of 26 4.4. Conclusions Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) 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 the health and safety of the public.

5. ENVIRONMENT CONSIDERATION The proposed change does not change any requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or does not change an inspection or surveillance requirement. The proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

GNRO2024-00033 Page 25 of 26

6. REFERENCES

1. Entergy Letter: Response to Generic Letter 2016-01, Monitoring of Neutron Absorbing Materials in Spent Fuel Pools. (GNRI-2016/00059 dated November 1, 2016) (ADAMS Accession No. ML16306A433)

2. NEI 12-16, Revision 4, Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants, September 2019. (ADAMS Accession Number ML19269E069)

3. NRC Interim Staff Guidance DSS-ISG-2010-01, Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools, Revision 0 (ADAMS Accession Number ML110620086)

4. NUREG-1434, Rev. 5.0, Standard Technical Specifications for General Electric BWR/6 Plants

5. NEI 16-03-A, Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools, Revision 0, March 2017 (ADAMS Accession No. ML17263A133)

6. TSTF-557, Spent Fuel Storage Rack Neutron Absorber Monitoring Program, Rev. 1, dated December 19, 2017 (ADAMS Accession No. ML17353A608)

7. LaSalle County Station, Units 1 and 2 - Issuance of Amendments Concerning Spent Fuel Neutron Absorbers (TAC Nos. ME2376 and ME2377), dated January 28, 2011 (ADAMS Accession No. ML110250051)

8. 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. ME7537 and ME7539), dated May 21, 2013 (ADAMS Accession No. ML13114A929)

9. Quad Cities Nuclear Power Station, Units 1 and 2 - Issuance of Amendments Regarding NETCO Inserts (TAC Nos. MF2489 and MF2490) (RS-13-148), dated December 31, 2014 (ADAMS Accession No. ML14346A306)

10. 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 (ADAMS Accession No. ML19357A009)

11. FERMI 2 - Issuance of Amendment No. 220 Re: Revision to the Renewed Facility Operating License Including the Technical Specifications to Utilize Neutron Absorbing Inserts in Criticality Safety Analysis for Spent Fuel Pool Storage Racks (EPIC L-2019-LLA-0199), dated May 24, 2021 (ADAMS Accession No. ML21029A254)

12. NETCO Report NET-259-03, Revision 5, Material Qualification of Alcan Composite for Spent Fuel Storage, August 2008 (ADAMS Accession Number ML13199A039)

13. Entergy Letter: License Renewal Application Grand Gulf Nuclear Station, Unit 1, Docket No. 50-416, (GNRO-2011/00093 dated October 28, 2011)

14. NRC Letter and

Enclosure:

Updated Safety Evaluation Report Related to the Grand Gulf Nuclear Station License Renewal Application (TAC No. ME7493), October 18, 2016 (ADAMS Accession Number ML16288A185)

GNRO2024-00033 Page 26 of 26

15. NUREG-1801, Generic Aging Lessons Learned (GALL) Report, Revision 2, December 2010 (ADAMS Accession Number ML103490041)

16. General Electric Company, Steady-State Nuclear Methods, NEDE-30130-P-A, April 1985 (Non-Proprietary Version - ADAMS Accession No. ML14104A064)

17. NRC Letter: Amendment 26 to GE licensing Topical Report NEDE-24011-P-A, GESTAR II - Implementing Improved GE Steady-State Methods (TAC No. MA6481), November 1999 (ADAMS Accession Number ML993230387)

18. NRC Letter: Final Safety Evaluation for GE Hitachi Nuclear Energy Licensing Topical Report NEDC-33374P, Revision 3, Safety Analysis Report for Fuel Storage Racks Criticality Analysis for ESBWR Plants., September 2010 (ADAMS Accession Number ML102430582)

GNRO2024-00033 Technical Specification Pages - Marked-up (3 pages below)

Design Features 4.0 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 (U02 ) 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 maintained with:

are designed and shall be

a. ketf s 0.95 if fully flooded wi h unborated water, which includes an allowance for unce tainties as described in Section 9.1.2 of the UFSAR;

b.

c.

Fuel assemblies having a maximum K-infinity of -1. in

{

the normal reactor core configuration at cold conditions; 1.29 (continued)

GRAND GULF 4.0-1 Amendment No.

, H-4, -1/4

}'. nominal fuel a

/

semu 1y cent~

to center stora~e of 6. 26 inches in the stora~e racks.

spacin§

Design Features 4.o 4.3.1.1 (continued)

d.

Fuel assemblies having a maximum nominal U-235 enrichment of 4.9 weight percent;

e.

Region II racks are controlled as follows:

1. Storage cells with any Boraflex panel which has received a gamma dose in excess of 2.3E10 rads or which has a Boron-10 areal density less than 0.0165, which are designated within the Spent Fuel Pool Rack Boraflex Monitoring Program, are treated as Region II panels.

2. Storage cells face-adjacent to Region II panels are either restricted from fuel storage by physically blocking the isolated cells or are configured to meet, as a minimum (i.e., additional cells may be blocked),

the Region II fuel storage configuration requirements in Figure 4.3-1.

3. When a 4x4 array of cells is classified as Region II and face-adjacent to another Region II 4x4 storage array, the new Region II 4x4 array is required to be blocked in the same 8-of-16 pattern and at the same orientation as the adjacent Region II 4x4 storage configuration.

Figure 4.3.1 Region II 4x4 Storage Configuration Fuel Assembly Storage Location Location Physically Blocked to Prevent Storage (continued)

Amendment No. 4 -0, 195 4.0 DESIGN FEATURES Delete 4.0-2 GRAND GULF

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 16d 5.B Programs and Manuals GRANO GULF 5.o Programs and Manuals 5.5 Amendment No. -234"-t

GNRO2024-00033 Technical Specification Pages - Clean (3 pages below)

Design Features 4.0 (continued)

GRAND GULF 4.0-1 Amendment No. 195, 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 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1.2 of the UFSAR;

b.

A nominal 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;

Design Features 4.0 (continued)

GRAND GULF 4.0-2 Amendment No. 195, 4.0 DESIGN FEATURES (continued) 4.3.1.1 (continued)

d.

Fuel assemblies having a maximum nominal U-235 enrichment of 4.9 weight percent;

GRAND GULF 5.0-16c Amendment No.

5.5 Programs and Manuals Programs and Manuals 5.5 5.5.14 Risk Informed Completion Time Program (continued)

c.

When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.

1.

For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2.

For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e.,

not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

3.

Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.

d.

For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1.

Numerically accounting for the increased possibility of CCF in the RICT calculation; or

2.

Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the functions(s) performed by the inoperable SSCs.

e.

The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods approved for use with this program in Amendment No. 234, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

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.

GNRO2024-00033 Global Nuclear Fuels Report NEDO-34125, Rev. 0, Dated July 2024, Grand Gulf Nuclear Station: Fuel Storage Critically Analysis with Rack Inserts (Non-Proprietary Version)

(65 pages below)

Global Nuclear Fuel NEDO-34125 Revision 0 July 2024 Non-Proprietary Information Grand Gulf Nuclear Station:

Global Nuclear Fuel

NEDO-34125 Revision 0 Non-Proprietary Information ii INFORMATION NOTICE This is a non-proprietary version of the document NEDC-34125P Revision 0, which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here ((

)).

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The design, engineering, and other information contained in this document is furnished for the purpose of providing the results of the fuel storage rack criticality analysis for Grand Gulf Nuclear Station. The only undertakings of GNF with respect to information in this document are contained in the contracts between Entergy and GNF, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than Entergy, or for any purpose other than that for which it is furnished by GNF is not authorized; and with respect to any unauthorized use, GNF makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

NEDO-34125 Revision 0 Non-Proprietary Information iii Revision Status Revision Number Date Description of Change 0

July 2024 Initial issue.

NEDO-34125 Revision 0 Non-Proprietary Information iv Table of Contents 1.0 Introduction........................................................................................................................1 2.0 Requirements..................................................................................................................... 1 3.0 Method of Analysis............................................................................................................ 1 3.1 Cross-Sections..................................................................................................................... 2 3.2 Geometry Treatment........................................................................................................... 2 3.3 Convergence Checks........................................................................................................... 2 3.4 Validation and Computational Basis................................................................................... 3 3.5 In-Core k Methodology..................................................................................................... 6 3.6 Definitions........................................................................................................................... 7 3.7 Assumptions and Conservatisms........................................................................................ 8 4.0 Fuel Design Basis............................................................................................................. 10 4.1 GE14 Fuel Description...................................................................................................... 10 4.2 GNF2 Fuel Description..................................................................................................... 13 4.3 GNF3 Fuel Description..................................................................................................... 15 4.4 Fuel Model Description.................................................................................................... 19 5.0 Criticality Analysis of Fuel Storage Racks................................................................... 20 5.1 Description of Fuel Storage Racks.................................................................................... 20 5.2 Fuel Storage Rack Models................................................................................................ 21 5.3 Design Basis Lattice Selection.......................................................................................... 25 5.4 Normal Configuration Analysis........................................................................................ 28 5.4.1 Analytical Models..................................................................................................... 28 5.4.2 Normal Configuration Results...................................................................................29 5.5 Bias Cases......................................................................................................................... 29 5.5.1 Depletion Bias Cases................................................................................................ 29 5.5.2 Normal Bias Cases.................................................................................................... 30 5.5.3 Abnormal/Accident Bias Cases................................................................................ 32 5.5.4 Results....................................................................................................................... 39 5.6 Uncertainties..................................................................................................................... 40 5.6.1 Tolerance Analytic Models....................................................................................... 40 5.6.2 Results....................................................................................................................... 41 5.7 Maximum Reactivity........................................................................................................ 43 6.0 Interfaces Between Areas with Different Storage Conditions......................................43 7.0 Conclusions...................................................................................................................... 43 8.0 References........................................................................................................................ 44 Appendix A - MCNP-05P Code Validation.............................................................................. 45 A.1 - Trend Analysis................................................................................................................. 49 A.2 - Bias and Bias Uncertainty Calculation - Single Sided Tolerance Limit......................... 54 Appendix B - Legacy Fuel Storage Justification...................................................................... 57

NEDO-34125 Revision 0 Non-Proprietary Information v

List of Tables Table 1 - Summary kmax(95/95) Result.......................................................................................... 1 Table 2 - Summary of the Critical Benchmark Experiments..........................................................4 Table 3 - Area of Applicability Covered by Code Validation........................................................ 5 Table 4 - GE14 Fuel Stack Density as a Function of Gadolinia Concentration........................... 10 Table 5 - Nominal Dimensions for GE14 Fuel Lattice..................................................................12 Table 6 - Nominal Dimensions for GNF2 Fuel Lattice................................................................. 14 Table 7 - Nominal Channel Dimensions for GNF2 Lattice......................................................... 14 Table 8 - Fuel Stack Density as a Function of Gadolinia Concentration...................................... 15 Table 9 - Lattice Dimensions....................................................................................................... 17 Table 10 - Cell Dimensions.......................................................................................................... 17 Table 11 - Channel Dimensions................................................................................................... 18 Table 12 - Storage Rack Model Dimensions................................................................................ 25 Table 13 - Fuel Parameter Ranges Studied in Fuel Rack..............................................................26 Table 14 -Fuel Storage Rack In-Rack k Results - Normal Configurations............................... 29 Table 15 - Rack Periphery Study Results......................................................................................31 Table 16 - Results for a Misplaced Assembly Outside of Rack................................................... 36 Table 17 - Results for a Misplaced Assembly Against Corner of Rack....................................... 37 Table 18 - Results for Bundles in Unpoisoned Equipment Racks............................................... 38 Table 19 -Fuel Storage Rack Abnormal Bias Summary.............................................................. 38 Table 20 -Fuel Storage Rack Bias Summary................................................................................39 Table 21 -Fuel Storage Rack Tolerance and Uncertainty k Results.......................................... 42 Table 22 - Fuel Storage Rack Results Summary.......................................................................... 43 Table 23 - MCNP-05P Results for the Benchmark Calculations..................................................45 Table 24 - Trending Parameters................................................................................................... 49 Table 25 - Trending Results Summary..........................................................................................54 Table 26 - Bias and Bias Uncertainty for MCNP-05P with ENDF/B-VII.................................... 56 Table 27 - Recommended Bias and Bias Uncertainty in Criticality Analyses for MCNP-05P with ENDF/B-VII....................................................................................56 Table 28 - Limiting SCCG In-Core Eigenvalue of all Legacy GGNS Bundles............................57

NEDO-34125 Revision 0 Non-Proprietary Information vi List of Figures Figure 1 - GE14 Fuel Lattice Configuration................................................................................ 11 Figure 2 - GNF2 Fuel Lattice Configuration............................................................................... 13 Figure 3 - Channel Dimensions.................................................................................................... 14 Figure 4 - GNF3 Lattice Configuration........................................................................................ 16 Figure 5 - Channel 1/8 Cross-Sections........................................................................................... 18 Figure 6 - GNF2 VAN1 Lattice in MCNP-05P............................................................................ 20 Figure 7 - 4x4 Fuel Storage Rack Model..................................................................................... 22 Figure 8 - Cell Identifiers in the 4x4 Fuel Storage Rack Model.......................................................... 23 Figure 9 - Storage Rack Model Schematic................................................................................... 24 Figure 10 - Zoomed Storage Rack Model Schematic.................................................................. 24 Figure 11 - In-Rack Fuel Eigenvalue as a Function of In-Core Eigenvalue................................ 28 Figure 12 - Finite Misplaced Assembly Outside of Rack............................................................ 34 Figure 13 - Finite Misplaced Assembly Pushed Against Corner of Rack.................................... 35 Figure 14 - Storage of Fuel in Unpoisoned Equipment Racks..................................................... 37 Figure 15 - Scatterplot of EALF versus knorm........................................................................................................................ 50 Figure 16 - Scatterplot of wt.% 235U versus knorm............................................................................................................... 51 Figure 17 - Scatterplot of wt.% 239Pu versus knorm............................................................................................................. 52 Figure 18 - Scatterplot of H/X versus knorm............................................................................................................................. 53 Figure 19 - Normality Test of knorm Results................................................................................. 55

NEDO-34125 Revision 0 Non-Proprietary Information vii ACRONYMS Term Definition 2D Two-Dimensional AOA Area of Applicability BAF Bottom of Active Fuel BASE Base Lattice BOL Beginning-of-Life BWR Boiling Water Reactor CFR Code of Federal Regulations CW Curtiss-Wright Flow Control Service, LLC DOM Dominant Lattice EALF Energy of the Average Lethargy Causing Fission

((

))

GE General Electric GEH GE-Hitachi Nuclear Energy Americas, LLC GGNS Grand Gulf Nuclear Station GNF Global Nuclear Fuel - Americas, LLC H/X Hydrogen-to-Fissile HTC Haut Taux de Combustion MID Mid Lattice MOX Mixed Uranium-Plutonium Oxide NCA Nuclear Critical Assembly NEI Nuclear Energy Institute NRC Nuclear Regulatory Commission SCCG Standard Cold Core Geometry TCH Corner Thickness TCM Side Thickness TCS Groove Thickness TCS Tank Critical Assembly (as used in Table 2)

UO2 Uranium Dioxide

NEDO-34125 Revision 0 Non-Proprietary Information viii Term Definition US United States VAN Vanished Lattice WCS Groove Width WML (Major) Side 1/2-Width WMS Minor Side Width WRL Corner Ramp Width WRS Groove Ramp Width

NEDO-34125 Revision 0 Non-Proprietary Information 1

1.0 INTRODUCTION

This report describes the criticality analysis and results for the Grand Gulf Nuclear Station (GGNS) fuel pool and upper containment pool with credit for Curtiss-Wright Flow Control Service, LLC (CW) NETCO-SNAP-IN neutron absorbing inserts in each usable rack cell. No credit for the Boraflex neutron absorber is taken in this analysis. This analysis includes sufficient detail on the methodology and analytical models utilized in the criticality analysis to verify that the storage rack systems have been accurately and conservatively represented. This analysis covers the current GNF2, GNF3 and GE14 fuel product lines and all legacy fuel stored in GGNSs fuel pool and upper containment pool.

The racks are analyzed using the MCNP-05P Monte Carlo neutron transport program and ENDF/B-VII.0 cross-section library. The methodology used in this analysis is the peak Standard Cold Core Geometry (SCCG) in-core eigenvalue (k) criterion. A maximum cold, uncontrolled peak in-core k of 1.29 as defined by the lattice physics code TGBLA06 (Reference 1) is set as the limit for this analysis. As demonstrated in Table 1, the analysis resulted in a storage rack maximum k-effective (kmax(95/95)) less than 0.95 for normal and credible abnormal operation with tolerances and uncertainties taken into account.

Table 1 - Summary kmax(95/95) Result Region kmax(95/95)

Fuel Pool and Upper Containment Pool 0.92632 2.0 REQUIREMENTS Title 10 of the Code of Federal Regulations (CFR) Part 50 defines the requirements for the prevention of criticality in fuel storage and handling at nuclear power plants. 10 CFR 50.68 details specifically that the storage rack kmax(95/95) for fuel storage racks must be demonstrated to be 0.95 for normal and credible abnormal operation with tolerances and computational uncertainties taken into account. The Standard Review Plan (Reference 2) outlines the standards that must be met for these analyses. All necessary requirements are met in this analysis. Nuclear Energy Institute (NEI) 12-16 (Reference 3), endorsed by Regulatory Guide 1.240 (Reference 4) is used as the guidance documents for this analysis.

3.0 METHOD OF ANALYSIS In this evaluation, in-core k values and exposure dependent, pin-by-pin isotopic specifications are generated using the GE-Hitachi Nuclear Energy Americas, LLC (GEH)/GNF lattice physics production code TGBLA06. TGBLA06 solves Two-Dimensional (2D) diffusion equations with diffusion parameters corrected by transport theory to provide system multiplication factors and perform burnup calculations.

The fuel storage criticality calculations are then performed using MCNP-05P, the GEH/GNF proprietary version of MCNP5 (Reference 5). MCNP-05P is a Monte Carlo program for solving the linear neutron transport equation for a fixed source or an eigenvalue problem. The code implements the Monte Carlo process for neutron, photon, electron, or coupled transport involving

NEDO-34125 Revision 0 Non-Proprietary Information 2

all these particles, and computes the eigenvalue for neutron-multiplying systems. For the present application, only neutron transport is considered.

3.1 Cross-Sections TGBLA06 uses ENDF/B-V cross-section data to perform coarse-mesh, broad-group, diffusion theory calculations. It includes thermal neutron scattering with hydrogen using an S(,) light water thermal scattering kernel.

MCNP-05P uses pointwise (i.e., continuous) cross-section data, and all reactions in a given cross-section evaluation (e.g., ENDF/B-VII.0) are considered. For the present work, thermal neutron scattering with hydrogen is described using an S(,) light water thermal scattering kernel.

The cross-section tables include all details of the ENDF representations for neutron data. The code requires that all the cross-sections be given on a single union energy grid suitable for linear interpolation; however, the cross-section energy grid varies from isotope to isotope. The libraries include very little data thinning and utilize resonance integral reconstruction error tolerances of 0.001%.

3.2 Geometry Treatment TGBLA06 is a 2D lattice design computer program for Boiling Water Reactor (BWR) fuel bundle analysis. It assumes that a lattice is uniform and infinite along the axial direction and that the lattice geometry and material are reflecting with respect to the lattice boundary along the transverse directions.

MCNP-05P implements a robust geometry representation that can correctly model complex components in three dimensions. An arbitrary three-dimensional configuration is treated as geometric cells bounded by first and second-degree surfaces and some special fourth-degree elliptical tori. The cells are described in a cartesian coordinate system and are defined by the intersections, unions and complements of the regions bounded by the surfaces. Surfaces are defined by supplying coefficients to the analytic surface equations or, for certain types of surfaces, known points on the surfaces. Rather than combining several pre-defined geometrical bodies in a combinatorial geometry scheme, MCNP-05P has the flexibility of defining geometrical shapes from all the first and second-degree surfaces of analytical geometry and elliptical tori and then combining them with Boolean operators. The code performs extensive checking for geometry errors and provides a plotting feature for examining the geometry and material assignments.

3.3 Convergence Checks The use of TGBLA06 as a depletion code in this criticality analysis is consistent with its use for BWR fuel design and its associated users manual. Convergence checks are encoded in the standard error routines and the absence of error messages is confirmed in all code output.

In this analysis, the following criticality code parameters are specified. At a minimum, all MCNP-05P cases are run with 20,000 neutrons per generation, 200 cycles skipped, and 500 total cycles run. Some cases are run for more cycles skipped and more total cycles to meet all the

NEDO-34125 Revision 0 Non-Proprietary Information 3

convergence checks. For this analysis, the following MCNP-05P convergence checks are reviewed and confirmed passed for each case:

Sampling of all cells that contain fissionable material Matching of first and second half eigenvalue Fission source entropy check 3.4 Validation and Computational Basis MCNP-05P has been compared to ((

)) critical experiments for validation purposes using ENDF/B-VII.0 nuclear cross-section data. The experiments cover a number of moderator-to-fuel ratios and poison materials that represent material and geometric properties similar to that of BWR fuel lattices both in and out of fuel racks. The critical experiments to which MCNP-05P has been compared are provided in Table 2. All are either low-enriched Uranium Dioxide (UO2) or Mixed Uranium-Plutonium Oxide (MOX) pin lattice in water experiments. The Area of Applicability (AOA) considered covered by this validation is listed in Table 3, along with the parameters which characterize the fuel rack system for comparison. The critical experiment modeling results, along with the calculation of the associated bias and bias uncertainty terms at the 95/95 confidence level using NUREG/CR-6698 (Reference 6) guidance is provided in Appendix A. The study concluded that the appropriate bias to apply to systems covered by this AOA is (( )), and the appropriate uncertainty of that bias is ((

)).

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Table 2 - Summary of the Critical Benchmark Experiments Experiment Experiments Year Where

((

))

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Table 3 - Area of Applicability Covered by Code Validation Parameters Validation Area of Applicability Fuel Rack Characteristics Fissionable Material Uranium, Plutonium Uranium, Actinides Chemical Form UO2, MOX UO2, MOX Enrichment (wt.% 235U) wt.% 235U 4.9 wt.% 235U 4.9 Enrichment (wt.% 239Pu) wt.% 239Pu 5.3 wt.% 239Pu 4.9 Physical Form Solid Compound Solid Compound Temperature

~20°C up to ~100°C 4-126°C Moderator (in fuel region)

H2O H2O Physical Form Solution Solution Temperature

~20°C up to ~100°C 4-126°C Reflector (in fuel region)

H2O H2O Physical Form Solution Solution Temperature 20°C 4-126°C Absorbers None/Boron/Gadolinium Stainless Steel /Copper Boron/Gadolinium/

Fission Products Neutron Energy Spectrum Thermal Thermal Energy of Average Lethargy Causing Fission (MeV) 6.8E 8.6 E-7 3.53E-7 (Limiting In-rack k Case)

Table 3 demonstrates that the AOA of this validation encompasses the majority of storage characteristics of new fuel in the fuel storage racks. ((

))

For the storage of fuel, however, it is appropriate to add additional uncertainty terms to the kmax(95/95) result. Specifically, these items are:

a. Uncertainty in fuel depletion calculations Consistent with NEI 12-16 (Reference 3), a conservative approximation of the fuel depletion uncertainty is quantified by assessing the reactivity difference between a Beginning-of-Life (BOL) system and the exposure dependent, peak reactivity system of interest. Specifically, the cold, in-core, BOL reactivity of the fuel rack design basis lattice with no gadolinium present is compared to the reactivity of the exposed design basis lattice at its cold, in-core, peak reactivity statepoint. Both reactivities are calculated for comparison in the rack system. Five percent of the difference in reactivities between these two cases is included as an uncertainty to the fuel rack studies in Table 21 to cover the depletion isotopic benchmarking gap, including the gap for minor actinides and fission products.

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b. TGBLA06 eigenvalue uncertainty An additional uncertainty is also added to the fuel rack studies related to eigenvalue calculations performed using TGBLA06. A bias of (( )) and a 95/95 bias uncertainty of ((

)) This uncertainty is applied to the fuel racks kmax(95/95) value to cover uncertainty in the assignment of in-core k values to fuel lattices.

3.5 In-Core k Methodology The design of the fuel storage racks provides for a subcritical multiplication factor for both normal and credible abnormal storage conditions. In all cases, the storage rack eigenvalue must be 0.95.

To demonstrate compliance with this limit, the peak in-core k method is utilized.

The peak in-core k criterion method relies on a well-characterized relationship between infinite lattice k (in-core) for a given fuel design and a specific fuel storage rack k (in-rack) containing that fuel. The use of an infinite lattice k criterion for demonstrating compliance to fuel storage criticality criteria has been used for all General Electric (GE)-supplied storage racks and is currently used for re-rack designs at a number of plants. This report demonstrates that the methodology is also appropriate for use at GGNS by presenting the following:

a. A well-characterized, linear relationship between infinite lattice k (in-core) and fuel storage rack k (in-rack)

b. The use of a design basis lattice with a conservative rack efficiency and in-core k for all criticality analyses The analysis is performed to calculate the lattice k to confirm compliance with the above criterion by utilizing the Nuclear Regulatory Commission (NRC)-approved lattice physics methods encoded into the TGBLA06 engineering computer program. One of the outputs of the TGBLA06 solution is the lattice k of a specific nuclear design for a given set of input state parameters (e.g., void fraction, control state, fuel temperature).

Compliance of fuel with specified k limits will be confirmed for each new lattice as part of the bundle design process. Documentation that this has been met will be contained in the fuel design information report, which defines the maximum lattice k for each assembly nuclear design. The process for validating that specific assembly designs are acceptable for storage in the GGNS fuel storage racks is provided below.

1.

Identify the unique lattices in each assembly design.

2.

Deplete the lattices in TGBLA06 using the following conditions:

a. Assembly aligned according to GGNS specific lattice spacing and zero leakage

b. ((

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))

3.

Ensure that the k values obtained from Step 3 for each lattice are less than or equal to the k limit of 1.29.

Documentation that all legacy fuel types currently in the GGNS fuel storage racks comply with this in-core limit is found in Appendix B.

3.6 Definitions Fuel Assembly - A complete fuel unit consisting of a basic fuel rod structure that may include large central water rods. Several shorter rods may be included in the assembly. These are called part-length rods. A fuel assembly includes the fuel channel.

Fuel Storage Rack - An array of usable rack cells, which refers to both the spent fuel pool and the upper containment pool. Both the spent fuel pool and upper containment pool have the same usable rack cell configurations.

Usable Rack Cell - A rack cell containing a neutron absorbing insert that is accessible by fuel handling equipment where a fuel bundle can be physically placed within.

Gadolinia - The compound Gd2O3. The gadolinium content in integral burnable absorber fuel rods is usually expressed in weight percentage gadolinia.

Lattice - An axial zone of a fuel assembly within which the nuclear characteristics of the individual rods are unchanged.

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Base Lattice (BASE) - An axial zone of a GNF2 or GNF3 fuel assembly located in the bottom half of the bundle within which all possible fuel rod locations for a given fuel design are occupied.

Dominant Lattice (DOM) - An axial zone of a GE14 fuel assembly typically located in the bottom half of the bundle within which all possible fuel rod locations for a given fuel design are occupied.

Mid Lattice (MID) - ((

))

Vanished Lattice (VAN) - An axial zone of a fuel assembly typically in the upper half of the bundle within which a number of possible fuel rod locations are unoccupied.

Rack Efficiency - The ratio of a particular lattice statepoint in-rack eigenvalue (k) to its associated lattice nominal in-core eigenvalue (k). This value allows for a straightforward comparison of a racks criticality response to varying lattice designs within a particular fuel product line. A lower rack efficiency implies increased reactivity suppression capability relative to an alternate design with a higher rack efficiency.

Design Basis Lattice - The lattice geometry, exposure history, and corresponding fuel isotopics for a fuel product line that result in the highest rack efficiency in a sensitivity study of reasonable fuel parameters at the desired in-core reactivity. This lattice is used for all normal, abnormal, and tolerance evaluations in the fuel rack analysis.

3.7 Assumptions and Conservatisms The fuel storage rack criticality calculations are performed with the following assumptions to ensure the true system reactivity is always less than the calculated reactivity:

1. ((

))

3. Design basis lattices with in-core k values greater than the proposed 1.29 in-core k limit is used for all criticality analyses.

4. ((

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)) Sensitivity studies of the storage system reactivity to these depletion parameters are presented in Section 5.5. ((

))

5. For conservatism, only positive reactivity differences from nominal conditions determined from depletion sensitivity and abnormal configuration analyses are added as biases to the final storage rack kmax(95/95).

6. Neutron absorption in spacer grids, concrete, activated corrosion and wear products (CRUD) and axial blankets is ignored to limit parasitic losses in non-fuel materials.

7. TGBLA06 defined lumped fission products and Xe-135 are both conservatively ignored for MCNP-05P in-rack k calculations.

8. ((

))

9. The neutron absorber inserts are modeled with nominal minimum wing width of

((

)) inches and nominal wing thickness of ((

)) inches. The wing length does not include the insert material which is bent at a 90-degree angle at the end of each wing.

Including this material, the total unbent insert length is greater than (( )) inches. Each wing is modeled at a wing length of ((

)) inches to represent all inserts in the rack which is an equivalent ((

)) inches total unbent insert length. Because the analysis models less material than is actually present in the insert, this approach is conservative. Modeling the inserts in this way minimizes thermal neutron absorption in the inserts.

10. Only B10 is modeled in the rack inserts. The minimum certified areal density is

((

)) g B10/cm2. Each insert is assumed to contain an areal density of 0.0139 g B10/cm2 to account for potential manufacturing uncertainties. All other insert material is ignored. Ignoring the other materials conservatively limits neutron absorption in the inserts.

11. No credit is taken for the Boraflex in the storage racks in the analysis, and all material between the inner cell wall and outer wrapper of the fuel rack is modeled as water. Modeling this material as water is reasonable, as the outer wrapper does not provide a watertight seal between the Boraflex and pool environment, and therefore any significant gap formations within the poison material will be filled with water.

NEDO-34125 Revision 0 Non-Proprietary Information 10 4.0 FUEL DESIGN BASIS The rack criticality analysis covers the GE14, GNF2, and GNF3 fuel product lines as well as all legacy fuel stored at GGNS. Justification for the storage of all legacy fuel is provided in Appendix B. The description of the fuel product lines, GE14, GNF2 and GNF3, are found in Sections 4.1, 4.2, and 4.3. All of these product lines are investigated to determine the design basis lattice in Section 5.3.

All fuel is UO2 with some fuel rods containing gadolinia, Gd2O3.

This criticality analysis covers reconstituted fuel where a rod containing fuel is replaced with another fueled or non-fueled rod. This analysis does not cover reconstituted fuel where there are missing rod locations that are not part of the normal fuel product line design.

This criticality analysis also bounds the storage of non-fuel items such as channels in fuel rack locations because this analysis utilizes peak reactivity fuel in every rack cell location.

4.1 GE14 Fuel Description The GE14 fuel lattice configuration is a 10x10 fuel rod array ((

)), as shown in Figure 1. Figure 1 also demonstrates the part-length rod locations, which cannot be changed for this fuel design. ((

)) Information regarding the GE14 pellet stack density is provided in Table 4. The corresponding dimensions of Figure 1 are provided in Table 5.

Table 4 - GE14 Fuel Stack Density as a Function of Gadolinia Concentration Gadolinia Concentration (wt. fraction)

((

Pellet Density (g/cc)

))

NEDO-34125 Revision 0 Non-Proprietary Information 11

((

))

Figure 1 - GE14 Fuel Lattice Configuration

NEDO-34125 Revision 0 Non-Proprietary Information 12 Table 5 - Nominal Dimensions for GE14 Fuel Lattice Features Reference (mm)

(inches)

Channel Dimensions:

((

))

Fuel Rod Dimensions:

((

))

Water Rod Dimensions:

((

))

Bundle Lattice Dimensions:

((

))

((

)) The full lattice, also referred to in this report as the dominant lattice (DOM), ((

)) The vanishing rod lattice, or vanished lattice (VAN), ((

)) Variation in axial height of these regions is irrelevant to this analysis due to the fact that all criticality calculations are performed with a single lattice design and burnup that corresponds to the highest rack efficiency.

NEDO-34125 Revision 0 Non-Proprietary Information 13 4.2 GNF2 Fuel Description Criticality safety analyses to determine storage system reactivity are performed using the GNF2 fuel design. The GNF2 fuel lattice configuration is a 10x10 fuel rod array ((

)) as shown in Figure 2 with corresponding dimensions in Table 6. Figure 2 also demonstrates the part-length rod locations, which cannot be changed for this fuel design. The references in Figure 3 correspond to Table 7. GNF2 pellet stack density is provided in Table 8. ((

))

Figure 2 - GNF2 Fuel Lattice Configuration

NEDO-34125 Revision 0 Non-Proprietary Information 14 Table 6 - Nominal Dimensions for GNF2 Fuel Lattice Features Reference (mm)

(inches)

Channel Dimensions:

((

))

Fuel Rod Dimensions:

((

))

Water Rod Dimensions:

((

))

Bundle Lattice Dimensions:

((

))

((

))

Figure 3 - Channel Dimensions Table 7 - Nominal Channel Dimensions for GNF2 Lattice Dimension mm inches

((

))

NEDO-34125 Revision 0 Non-Proprietary Information 15 Table 8 - Fuel Stack Density as a Function of Gadolinia Concentration Gadolinia Concentration (wt. fraction)

((

Pellet Density (g/cc)

))

((

)) The full lattice, also referred to in this report as the base lattice (BASE), ((

)) The first vanishing rod lattice, or vanished one lattice (VAN1), ((

)) The second vanished rod lattice (VAN2) ((

)) Variation in axial height of these regions is irrelevant to this analysis because all criticality calculations are performed assuming a single lattice design.

4.3 GNF3 Fuel Description The GNF3 fuel lattice configuration is a 10x10 fuel rod array ((

)) as shown in Figure 4 with corresponding dimensions in Table 9 and Table 10. Figure 4 also demonstrates the part-length rod locations. Fuel channel dimensions are provided in Figure 5 and Table 11. The pellet stack density is in Table 8. ((

))

NEDO-34125 Revision 0 Non-Proprietary Information 16

((

))

Figure 4 - GNF3 Lattice Configuration

NEDO-34125 Revision 0 Non-Proprietary Information 17 Table 9 - Lattice Dimensions Item Dimension mm in Channel

((

Fuel Rod Pellet

((

))

Bundle Lattice

))

Table 10 - Cell Dimensions Lattice Type Channel Name 1/2 Wide Gap, Q 1/2 Narrow Gap, R Control Blade Pitch, S mm in mm in mm in

((

))

NEDO-34125 Revision 0 Non-Proprietary Information 18

((

))

Figure 5 - Channel 1/8 Cross-Sections Table 11 - Channel Dimensions Channel Name 93AV Channel Section Zone 1 Zone 2 Dimension mm in mm in

((

))

NEDO-34125 Revision 0 Non-Proprietary Information 19 4.4 Fuel Model Description The fuel models considered include 2D geometric modeling of all fuel material, cladding, water rods, and channels. In the depletion model, appropriate depletion time steps are used consistent with depletion timesteps used in BWR core design analyses. ((

)) Pin specific isotopic modeling as a function of exposure is performed based on the lattice physics code TGBLA06. To obtain the isotopic composition of the fuel pins, each lattice design considered is burned at reactor operating conditions ((

)) and depleted through to a final exposure of ((

Xe-135 and TGBLA06 defined lumped fission products ((

)) The isotopics utilized exclude

)) An example of a GNF2 VAN1 lattice model in MCNP-05P (Case 12 from Table 13) is depicted in Figure 6. The black pins are the gadolinia rods. ((

))

NEDO-34125 Revision 0 Non-Proprietary Information 20

((

))

Figure 6 - GNF2 VAN1 Lattice in MCNP-05P The fuel loadings considered for each lattice span a range of exposures, average enrichments, number of gadolinia rods, gadolinia concentration, and void histories considered to be reasonably representative of any GGNS fuel design. The lattice type and exposure history that result in the worst-case rack efficiency for an in-core k greater than the proposed limit is then used to define the design basis lattice. This lattice is assumed to be stored in every location in the rack being analyzed. Details on the determination of the design basis lattice using the process outlined above are presented in Section 5.3.

5.0 CRITICALITY ANALYSIS OF FUEL STORAGE RACKS 5.1 Description of Fuel Storage Racks The GGNS high-density fuel storage rack is a Joseph Oat design and uses Boraflex as a neutron absorber. Racks of this design are present in the fuel pool and upper containment pool. This analysis applies to racks in both locations. The rack design uses L and T shaped sub-elements to assemble each fuel storage array. Each sub element is composed of stainless-steel sheets, Boraflex inserts, and stainless steel edge strips.

Originally, the storage racks at GGNS employed thermal neutron absorption in the B10 of the Boraflex as the primary mechanism of reactivity control; however, the Boraflex has been demonstrated to be degrading over time. Therefore, no credit is taken for the Boraflex in this analysis, and all material between the inner cell wall is modeled as water. Modeling this material as water is reasonable, as the outer wrapper does not provide a watertight seal between the Boraflex and pool environment.

Therefore, any significant gap formations within the poison material will be filled with water.

To supplement the reactivity suppression capability of the rack, neutron absorbing inserts are installed in each of the usable storage cells in the storage rack module. In this analysis, a lower B10 areal density of 0.0139 g B10/cm2 is used in the base model instead of the certified minimum B10 areal

NEDO-34125 Revision 0 Non-Proprietary Information 21 density of ((

)) g B10/cm2 to account for potential manufacturing uncertainties. The minimum designed wing length for these inserts is ((

)) inches. This length does not include the insert material which is bent at a 90-degree angle at the end of each wing. Including this material, the total unbent insert length is greater than (( )) inches. For simplicity, each wing is modeled with a

((

))-inch wing length to conservatively represent all inserts in the rack. Each insert is installed with the same north-east orientation with respect to the cell. In this way, one leg of an insert exists between each bundle in the storage rack assembly.

Based on the insert configuration, peripheral storage cells on two sides of the storage pools will not be surrounded by four wings of the absorbing insert. The reactivity effect of this storage limitation is assessed in Section 5.5.

5.2 Fuel Storage Rack Models A 2D infinite 4x4 array was constructed to analyze the fuel storage rack. ((

)) Figure 7 displays a simplified 2D layout of the rack cells. The numbers 1-16 are unit identifiers, and each unit includes a fuel assembly. The i and ii identify the unit type, which is shown in Figure 8. Boraflex is not credited in this analysis, and all Boraflex is modeled as water.

Neutron absorbing inserts are positioned in a north-east orientation relative to each rack cell (see Figure 7). The fuel pool and the upper containment pool storage racks are the same dimensions and are denoted as fuel storage racks in this analysis.

NEDO-34125 Revision 0 Non-Proprietary Information 22 Figure 7 - 4x4 Fuel Storage Rack Model 13 11 9

l 5

II I

I

NEDO-34125 Revision 0 Non-Proprietary Information 23 Figure 8 - Cell Identifiers in the 4x4 Fuel Storage Rack Model To simulate an infinite array, periodic boundary conditions were specified in the X and Y dimensions and reflective boundary conditions were specified in the Z dimension. An image demonstrating the inner four bundles of the 4x4 infinite array model is provided in Figure 9 with a zoomed in view in Figure 10. Storage rack dimensions and tolerances are presented in Table 12.

((

)) cm Insert Thickness

((

)) cm Insert Length U.

91 c:ttdk*a:llocit Wi ID Insert

NEDO-34125 Revision 0 Non-Proprietary Information 24

((

))

Figure 9 - Storage Rack Model Schematic

((

))

Figure 10 - Zoomed Storage Rack Model Schematic

NEDO-34125 Revision 0 Non-Proprietary Information 25 Table 12 - Storage Rack Model Dimensions Storage Rack Component Nominal (inch)

Tolerance (inch)

Rack Pitch 6.259 0.062 Primary Fuel Box Inner Width 6.063 Rack Wall Thickness 0.063 0.006 Rack Insert Wing Length

((

Rack Insert Thickness

))

Modeled wing length of ((

)) inches. See Section 3.7 for modeling assumptions.

5.3 Design Basis Lattice Selection Table 13 defines the lattice designs and exposure histories that are explicitly studied in the fuel storage rack to determine the geometric configuration and isotopic composition that results in the worst rack efficiency. Note that void state is not a relevant parameter for zero exposure peak reactivity cases, and, therefore, only a single result is presented for these fuel loadings. The highest rack efficiency with an in-core k greater than the proposed limit of 1.29 is found to result from the parameters defined in Case 12 from Table 13. The geometry and isotopics defined for this case are used to define all bundles in the remaining fuel rack analyses.

Figure 11 presents a graph that demonstrates the linear nature of the in-core to in-rack results over all rack efficiency cases studied in the rack system. Figure 11 provides infinite in-core and in-rack eigenvalue pairs for GE14, GNF2, and GNF3 lattices at ((

)) to allow for the linear relationship to be demonstrated over a large range of exposures and fuel lattice designs.

NEDO-34125 Revision 0 Non-Proprietary Information 26 Table 13 - Fuel Parameter Ranges Studied in Fuel Rack Case Lattice Type Void Average Lattice Enrichment (235U wt.%)

Number of Gadolinia Rods Gadolinia Concentration (Gd wt. %)

Peak-Reactivity Exposure (GWD/ST)

TGBLA06 Defined In-Core k MCNP-05P Defined In-Rack k Rack Efficiency 1

((

0.8806

((

2 0.8791 3

0.8743 4

0.8913 5

0.8830 6

0.8725 7

0.8500 8

0.8849 9

0.8843 10 0.8818 11 0.8624 12 0.8961 13 0.8908 14 0.8807 15 0.8589 16 0.8878 17 0.8846 18 0.8792 19 0.8747 20 0.8758 21 0.8720 22 0.8765 23 0.8732 24 0.8521 25 0.8457 26 0.8860 27 0.8780 28 0.8676 29 0.8831 30 0.8812 31

))

0.8622

))

NEDO-34125 Revision 0 Non-Proprietary Information 27 Table 13 - Fuel Parameter Ranges Studied in Fuel Rack Case Lattice Type Void Average Lattice Enrichment (235U wt.%)

Number of Gadolinia Rods Gadolinia Concentration (Gd wt. %)

Peak-Reactivity Exposure (GWD/ST)

TGBLA06 Defined In-Core k MCNP-05P Defined In-Rack k Rack Efficiency 32

((

0.8615

((

33 0.8949 34 0.8883 35 0.8783 36 0.8755 37 0.8718 38 0.8357 39 0.8443 40 0.8882 41 0.8833 42 0.8766 43 0.8889 44 0.8829 45 0.8759 46 0.8746 47 0.8729 48 0.8677 49 0.8439 50 0.8850 51 0.8845 52 0.8842 53 0.8935 54 0.8915 55 0.8856 56 0.8726 57 0.8742 58

))

0.8753

))

Six gadolinia rods at 3 wt.% concentration and one gadolinia rod at 4.0 wt.% concentration.

NEDO-34125 Revision 0 Non-Proprietary Information 28

((

))

Figure 11 - In-Rack Fuel Eigenvalue as a Function of In-Core Eigenvalue 5.4 Normal Configuration Analysis 5.4.1 Analytical Models The most reactive normal configuration is determined by studying the reactivity effect of the following credible normal scenarios:

1. Storage of non-channeled assemblies

2. Eccentric loadings o When neutron absorber inserts with an areal density above 0.01 g 10B/cm2 are present on all four sides of the fuel assembly, a centrally located positioning of the fuel assembly in the storage cell is the most reactive configuration. Therefore, no eccentric loading cases are performed, which is consistent with NEI 12-16 (Reference 3).

3. ((

))

4. Pool moderator temperature variation As the non-channeled assembly evaluation demonstrates a decrease in reactivity when compared to nominal, channeled storage conditions, the remaining normal configuration studies are performed with channeled bundles.

NEDO-34125 Revision 0 Non-Proprietary Information 29 5.4.2 Normal Configuration Results The results of the normal configuration study are provided in Table 14. This information demonstrates that none of the normal configurations analyzed increase the system reactivity by a statistically significant amount over the nominal loading pattern. The in-rack k associated with this nominal combination of conditions is 0.89613 and is hereafter referred to as kNormal. This configuration will be used for all abnormal and tolerance studies that are performed on an infinite basis. Any positive reactivity differences from this nominal condition are included in the calculation of the system bias in Section 5.5.2.

Table 14 -Fuel Storage Rack In-Rack k Results - Normal Configurations Term Configuration In-Rack k MCNP-05P Uncertainty (1)

Base Nominal - Centered, channeled, ((

))

0.89613

((

kN1 Non-channeled assemblies 0.89185 kN2a ((

0.89636*

kN2b

))

0.89628 kN3a Moderator Temperature decrease to 4oC (=1.000 g/cc) 0.89655*

kN3b Moderator Temperature increase to 126oC with 20% void

(=0.7508 g/cc) 0.86062

))

Largest positive reactivity increase from nominal case for each term is included in roll-up of kBias 5.5 Bias Cases 5.5.1 Depletion Bias Cases The following configurations related to the depletion conditions of the stored bundles are explicitly considered, where each description defines a condition all bundles in storage experience over their entire exposure histories. These bound the conditions the bundles actually experience.

1. ((

NEDO-34125 Revision 0 Non-Proprietary Information 30

))

The following potential reactivity effect of changes that occur during depletion are considered:

a. Fuel rod changes (clad creep, fuel densification/swelling)

Clad Creep - ((

))

Fuel Pellet Densification - ((

))

b. Material dependent grid growth

((

))

5.5.2 Normal Bias Cases The following bias cases are included for normal conditions. As seen in Table 14, ((

)) and moderator temperature decrease cases resulted in positive reactivity increases from the nominal case. Therefore, these cases are included in the roll-up of kBias in Table 20.

1. No inserts on rack periphery There may be assemblies loaded in storage cells on two sides that will not be surrounded by neutron absorbing inserts. ((

)) Results are provided in Table 15. The reactivity increase from this study is included in the final kBias term.

NEDO-34125 Revision 0 Non-Proprietary Information 31 Table 15 - Rack Periphery Study Results Description keff MCNP-05P Uncertainty (1) k

((

))

((

No Inserts on Rack Periphery

))

2. Missing rack insert A missing insert from the ((

)) was analyzed to cover the periodic removal of an insert from a cell for inspection purposes or an insert being accidently removed during fuel movements. Thus, ((

)) The relative reactivity increase from this condition is included in the bias table in Table 20.

3. Fuel out of rack during normal fuel handling/inspections Several fuel assembly geometric configurations are possible in the fuel storage racks and fuel transfer area during fuel handling activities such as fuel stored in the fuel prep machines. ((

NEDO-34125 Revision 0 Non-Proprietary Information 32

))

5.5.3 Abnormal/Accident Bias Cases Additionally, perturbations of the normal fuel rack configuration were considered for credible accident scenarios. The scenarios considered are presented in the lists that follow, with explanations of the abnormal condition provided below each listing of similar configurations.

Results of these abnormal/accident conditions is included in the final kBias term in Table 20.

1. Dropped/damaged fuel

a. Justification - The dropped/damaged fuel scenario ((

))

The relative reactivity change from this abnormal condition is included in Table 19 and included in the in the final kBias term in Table 20.

2. Abnormal positioning of a fuel assembly outside the fuel storage rack

a. Justification - There is not enough space for a bundle to fit between racks in the fuel storage racks; however, there is space for a misplaced bundle outside the fuel racks and between the fuel storage rack walls. ((

)) as shown in Figure 12. ((

)) in the described directions provided in Table 16). Results from these analyses are presented in Table 16. As seen in Table 16, there is little sensitivity in the placement of the misplaced bundle. Therefore, the cases analyzed here are sufficient.

NEDO-34125 Revision 0 Non-Proprietary Information 33

b. Abnormal positioning of a fuel bundle pushed against corner of fuel rack Justification - A bundle could be misplaced at the edge of the fuel racks against the corner of the rack as shown in Figure 13. ((

)) Several orientations of the misplaced bundle were performed to assess the most reactive location as shown in Table 17. As seen in Table 17, there is little sensitivity in the placement of the misplaced bundle. Therefore, the cases analyzed here are sufficient.

NEDO-34125 Revision 0 Non-Proprietary Information 34

((

))

Figure 12 - Finite Misplaced Assembly Outside of Rack

NEDO-34125 Revision 0 Non-Proprietary Information 35

((

))

Figure 13 - Finite Misplaced Assembly Pushed Against Corner of Rack

NEDO-34125 Revision 0 Non-Proprietary Information 36 Table 16 - Results for a Misplaced Assembly Outside of Rack Description keff MCNP-05P Uncertainty (1) k

((

))

NEDO-34125 Revision 0 Non-Proprietary Information 37 Table 17 - Results for a Misplaced Assembly Against Corner of Rack Description keff MCNP-05P Uncertainty (1) k

((

))

3. Misplacement of fuel bundles in unpoisoned equipment racks next to the fuel racks Justification - It is possible for fuel bundles to be placed in unpoisoned equipment racks next to the fuel racks. The configuration of this study is a ((

)) This scenario was explicitly considered by studying three possible configurations with two bundles, as depicted in Figure 14. ((

)) The calculation was reperformed with bundles in the unpoisoned equipment racks to determine the limiting configuration. Results are provided in Table 18. The results of this study are bounded by the results of the misplaced bundle against the corner of rack study.

((

))

Figure 14 - Storage of Fuel in Unpoisoned Equipment Racks

NEDO-34125 Revision 0 Non-Proprietary Information 38 Table 18 - Results for Bundles in Unpoisoned Equipment Racks Description keff MCNP-05P Uncertainty (1) k

((

))

The following abnormal configurations are also considered bounded, with the justification provided:

4. Dropped bundle on rack Justification - For a drop on the rack, the fuel assembly may come to rest horizontally on top of the rack with a minimum separation distance from the fuel in the rack of more than 12 inches. At this separation distance, the fissile material will be separated by enough neutron mean free paths to preclude neutron interactions that increase keff, and the overall effect on reactivity will be insignificant. Therefore, no case was performed for this analysis consistent with NEI 12-16 (Reference 3).

5. Rack sliding due to seismic event which causes water gap between racks to close Justification - The racks modeled in this analysis are infinite in extent with no inter-module water gaps. This essentially assumes all racks are close-fitting and bounds possible reactivity effects of rack sliding.

6. Loss of fuel pool cooling Justification - Normal sensitivity analysis results demonstrate that system reactivity decreases as moderator density decreases and pool temperature increases; therefore, reactivity effects of loss of fuel pool cooling are bounded by the nominal reactivity results.

Table 19 -Fuel Storage Rack Abnormal Bias Summary Description keff MCNP-05P Uncertainty (1) k k Uncertainty (2)

Dropped/Damaged Fuel 0.89665 ((

)) 0.00052

((

))

((

))

Per the double contingency principle (Reference 3), only the most limiting of the misplaced bundle cases is included in the bias roll-up in Table 20.

NEDO-34125 Revision 0 Non-Proprietary Information 39 5.5.4 Results The results of the bias studies are provided in Table 20. The k term in this table represents the difference between the system reactivity with the specified bias case and kNormal. kB6 is the MCNP-05P bias from Section 3.4. The total contribution from these independent conditions to the kmax(95/95) of the fuel rack is calculated using Equation 1. In this equation, a kBi value must be both positive and the largest for its respective term to be considered.

n kbias= L kbi i=1 Table 20 -Fuel Storage Rack Bias Summary (1)

Term Description keff MCNP-05P Uncertainty (1) k k Uncertainty (2) kB1 ((

0.88088

((

-0.01525

((

kB2a 0.89621 0.00008 kB2b 0.89747 0.00134*

kB3a 0.89737 0.00124*

kB3b 0.89572

-0.00041 kB4a 0.89711 0.00098*

kB4b

))

0.89649 0.00036 kB5 Depleted with clad creep 0.89717 0.00104 kB6 MCNP-05P bias

((

))

kB7 Dropped/damaged fuel 0.89665 0.00052 kB8 No inserts on rack periphery

((

))

kB9 Missing insert 0.90165 0.00544 kB10 ((

((

))

kN2a

))

0.89636 0.00023 kN3a Moderator temperature decrease to 4°C

(=1 g/cc) 0.89655

)) 0.00042

))

kBias

((

))

For conservatism, only positive values that are the largest for their respective term are considered.

- ((

))

NEDO-34125 Revision 0 Non-Proprietary Information 40 5.6 Uncertainties 5.6.1 Tolerance Analytic Models The following tolerance study configurations were explicitly considered for the fuel rack:

1. Fuel enrichment increases by ((

)) 235U

2. Fuel pellet density increased by ((

)) of nominal value

3. Gadolinia concentration decreased by ((

))

4. Rod cladding thickness decreased by ((

)) and rod cladding outer diameter decrease by ((

))

5. Rod cladding thickness increased by ((

)) and rod cladding outer diameter increase by ((

))

6. Channel thickness increase by ((

))

7. Channel thickness decrease by ((

))

8. Fuel pellet outer diameter increase by ((

))

9. Fuel pellet outer diameter decrease by ((

))

10. Fuel rod pin pitch increase by ((

))

11. Fuel rod pin pitch decrease by ((

))

12. Rack wall thickness increase by 0.006 inches

13. Rack wall thickness decrease by 0.006 inches

14. Rack pitch decrease by 0.062 inches

15. Rack pitch increase by 0.062 inches

16. Rack insert thickness decrease by ((

))

17. Rack insert wing length decrease by ((

))

All the tolerances used in these analyses are at least 2 design limits. The models developed for these studies were all based on the normal configuration presented in Section 5.4.

The inner width tolerance case is covered by the rack pitch tolerance case because the rack pitch tolerance bounds the inner cell width tolerance. Because there is no tolerance on the rack wall thickness, the only way to change the inner box width is by changing the pitch.

Because the Boraflex is modeled as water in this analysis, no tolerance cases are performed on the Boraflex thickness or width.

NEDO-34125 Revision 0 Non-Proprietary Information 41 Ti Ui 5.6.2 Results The results of the tolerance studies and uncertainties are provided in Table 21. The values are summed using Equation 2 which is adopted from NEI 12-16 (Reference 3).

The kTi term in this table represents the difference between the system reactivity with the specified tolerance perturbation and kNormal. In Equation 2, a kTi value must be both positive and the largest for its respective term to be considered.

The kUi terms in the table represent the uncertainty contributions to kmax(95/95) of the fuel rack and from the problem and code specific uncertainties which are combined with the tolerance contributions (kTi) using Equation 2.

n kuncertainty= L k2 i1 n

L k2 i1 (2)

NEDO-34125 Revision 0 Non-Proprietary Information 42 Table 21 -Fuel Storage Rack Tolerance and Uncertainty k Results Term Description keff MCNP-05P Uncertainty (1) k k

Uncertainty (2) kT1 Fuel enrichment increase 0.90023

((

0.00410

((

kT2 Fuel pellet density increase 0.89719 0.00106 kT3 Gadolinia wt.% decrease 0.90244 0.00631 kT4 Rod clad thickness/outer diameter increase 0.89034

-0.00579 kT4b Rod clad thickness/outer diameter decrease 0.90181 0.00568*

kT5a Channel thickness increase 0.89619 0.00006 kT5b Channel thickness decrease 0.89619 0.00006*

kT6a Pellet outer diameter increase 0.89713 0.00100*

kT6b Pellet outer diameter decrease 0.89611

-0.00002 kT7a Fuel rod pin pitch increase 0.89778 0.00165*

kT7b Fuel rod pin pitch decrease 0.89503

-0.00110 kT8a Rack wall thickness increase 0.89736 0.00123*

kT8b Rack wall thickness decrease 0.89557

-0.00056 kT9a Rack pitch decrease 0.90024 0.00372*

kT9b Rack pitch increase 0.89246

-0.00406 kT11a Poison inserts thickness decrease 0.89603

-0.00049 kT11b Poison inserts thickness increase 0.89654 0.00002*

))

kT12 Poison inserts wing length decrease 0.89636

))

-0.00016 kU1 Critical benchmark bias uncertainty (95/95) (MCNP-05P versus critical experiments)

((

kU2 TGBLA06 eigenvalue uncertainty (95/95) kU3 Uncertainty on kNormal (2 x 1 value for base term in Table 14) kU4 Uncertainty of k bias contributors (2) kU5 Uncertainty of k tolerance contributors (2) kU6 Uncertainty in fuel depletion kUncertainty

))

For conservatism, only positive values that are the largest for their respective term are considered.

NEDO-34125 Revision 0 Non-Proprietary Information 43 5.7 Maximum Reactivity The maximum reactivity of the fuel storage racks without crediting Boraflex and with rack inserts installed, considering all biases, tolerances, uncertainties, is calculated using Equation 3. The final values are presented in Table 22.

kmax(95/95)=knormal+kbias+kuncertainty (3)

Table 22 - Fuel Storage Rack Results Summary Term Value kNormal 0.89613 kBias

((

kUncertainty

))

kmax(95/95) 0.92632

((

))

6.0 INTERFACES BETWEEN AREAS WITH DIFFERENT STORAGE CONDITIONS The fuel pool and upper containment pool are neutronically decoupled because the pools are not connected. A scenario was investigated in Section 5.5.3 to determine the reactivity effect in the fuel storage racks as a result of placing two bundles in unpoisoned equipment racks next to the fuel racks. This effect was found to be negligible to the final result of the storage rack maximum k-effective (kmax(95/95)).

7.0 CONCLUSION

S The GGNS fuel pool and upper containment pool with neutron absorbing inserts have been analyzed for the storage of GE14, GNF2, and GNF3 fuel using the MCNP-05P Monte Carlo neutron transport program and the in-core k criterion methodology. A maximum SCCG, uncontrolled peak in-core eigenvalue (k) of 1.29 as defined by TGBLA06 is specified as the rack design limit for GE14, GNF2, and GNF3 fuel in the fuel pool and upper containment pool with neutron absorber rack inserts installed. The analyses resulted in a storage rack maximum k-effective (kmax(95/95)) less than the 10 CFR 50.68 limit of 0.95 for normal and credible abnormal

NEDO-34125 Revision 0 Non-Proprietary Information 44 operation with tolerances and computational uncertainties taken into account. Justification for the continued storage of all legacy GGNS fuel is found in Appendix B.

8.0 REFERENCES

1. Letter from Stuart A. Richards (NRC) to Glen A. Watford (GE), Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, GESTAR II - Implementing Improved GE Steady-State Methods (TAC No. MA6481), MFN-035-99, November 10, 1999.

2. United States (US) NRC, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling, Revision 3, March 2007. (NRC ADAMS Accession Number ML070570006).

3. NEI 12-16 Revision 4, Guidance for Performing Criticality Analyses of Fuel Storage at Light-Water Reactor Power Plants, September 2019. (NRC ADAMS Accession Number ML19269E069).

4. US NRC, Regulatory Guide 1.240, Fresh and Spent Fuel Pool Criticality Analyses, Revision 0, March 2021. (NRC ADAMS Accession Number ML20356A127).

5. Los Alamos National Laboratory, LA-UR-03-1987, MCNP - A General Monte Carlo N-Particle Transport Code, Version 5, April 24, 2003 (Revised February 1, 2008).

6. US NRC, NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, January 2001. (NRC ADAMS Accession Number ML050250061).

7. J.R. Taylor, An Introduction to Error Analysis, page 268-271, 2nd Edition, University Science Books, 1997.

NEDO-34125 Revision 0 Non-Proprietary Information 45 APPENDIX A - MCNP-05P CODE VALIDATION Table 23 presents the results of the benchmark calculations described in Section 3.4. Note that it is necessary to make an adjustment to the calculated keff value if the critical experiment being modeled was not at a critical state. This adjustment is done by normalizing the kcalc values to the experimental values, which is valid for small differences in keff. This normalization is reported as knorm and is determined using Equation A-1. The combined uncertainty from the measurement and the calculation (t) is also determined using Equation A-2.

knorm=kcalc/kexp (A-1) t= 2 +2 (A-2) calc exp Table 23 - MCNP-05P Results for the Benchmark Calculations Experiment Expt.

Benchmark Eigenvalue (kexp)

Experimental Uncertainty (exp)

MCNP-05P Result (kcalc)

MCNP-05P Uncertainty (calc)

Norm.

Result (knorm)

Combined Uncertainty (t)

((

))

NEDO-34125 Revision 0 Non-Proprietary Information 46 Table 23 - MCNP-05P Results for the Benchmark Calculations Experiment Expt.

Benchmark Eigenvalue (kexp)

Experimental Uncertainty (exp)

MCNP-05P Result (kcalc)

MCNP-05P Uncertainty (calc)

Norm.

Result (knorm)

Combined Uncertainty (t)

((

))

NEDO-34125 Revision 0 Non-Proprietary Information 47 Table 23 - MCNP-05P Results for the Benchmark Calculations Experiment Expt.

Benchmark Eigenvalue (kexp)

Experimental Uncertainty (exp)

MCNP-05P Result (kcalc)

MCNP-05P Uncertainty (calc)

Norm.

Result (knorm)

Combined Uncertainty (t)

((

))

NEDO-34125 Revision 0 Non-Proprietary Information 48 Table 23 - MCNP-05P Results for the Benchmark Calculations Experiment Expt.

Benchmark Eigenvalue (kexp)

Experimental Uncertainty (exp)

MCNP-05P Result (kcalc)

MCNP-05P Uncertainty (calc)

Norm.

Result (knorm)

Combined Uncertainty (t)

((

))

NEDO-34125 Revision 0 Non-Proprietary Information 49 Table 23 - MCNP-05P Results for the Benchmark Calculations Experiment Expt.

Benchmark Eigenvalue (kexp)

Experimental Uncertainty (exp)

MCNP-05P Result (kcalc)

MCNP-05P Uncertainty (calc)

Norm.

Result (knorm)

Combined Uncertainty (t)

((

))

A.1 - Trend Analysis To determine if any trend is evident in this pool of experiments, the parameters listed in Table 24 were considered as independent variables.

Table 24 - Trending Parameters Energy of the Average Lethargy Causing Fission (EALF)

Uranium Enrichment (wt.% 235U)

Plutonium Content (wt.% 239Pu)

Atom Ratio of Hydrogen to Fissile Material Each parameter was plotted against the knorm results independently for each case that was analyzed.

These plots are provided in Figure 15 through Figure 18. This scatterplot of data was first analyzed by visual inspection to determine if any trends were readily apparent in the data. During this inspection, the axes of the graphs were modified to different scales to allow for a more thorough review. No clear evidence of a trend, linear or otherwise, was observed from this inspection.

NEDO-34125 Revision 0 Non-Proprietary Information 50

((

))

Figure 15 - Scatterplot of EALF versus knorm

NEDO-34125 Revision 0 Non-Proprietary Information 51

((

))

Figure 16 - Scatterplot of wt.% 235U versus knorm

NEDO-34125 Revision 0 Non-Proprietary Information 52

((

))

Figure 17 - Scatterplot of wt.% 239Pu versus knorm

NEDO-34125 Revision 0 Non-Proprietary Information 53

((

))

Figure 18 - Scatterplot of H/X versus knorm To further check for trends in the data, a linear regression was performed. The linear regression fitted equation is in the form y(x)= a +bx, where y is the dependent variable (kcalc) and x is any of the predictor variables from Table 24. Unweighted kcalc values were used in this evaluation, though it is noted that, due to the very similar calc values reported in Table 23, using weighted values would produce very similar results. This regression was performed using the built-in regression analysis tool in Excel. The fitted lines are included in Figure 15 through Figure 18. Again, it is noted through visual inspection that the trends do not appear to exhibit a strong correlation to the data. A useful tool to validate this claim is the linear correlation coefficient. This is a quantitative measure of the degree to which a linear relation exists between two variables. It is often expressed as the square term, r2, and can be calculated directly using built in functions in Excel. The closer r2 gets to the value of one, the better the fit of data is expected to be to the linear equation. Results from this linear regression evaluation are summarized in Table 25.

A final method to test for goodness of fit is the chi squared test (2). This method is explained in detail in (Reference 7). In general, it can be stated that 2 is an indicator of the agreement between the observed (calculated) and expected (fitted) values for some variable. For linear goodness of

NEDO-34125 Revision 0 Non-Proprietary Information 54 fit testing using this method, Equation A-3 is utilized, where the expected value of f(xi) corresponds to the linear fitted equation for the trending parameter, xi.

(A-3) 1 J

A more convenient way to report this result is the reduced chi squared value, which is denoted as

- and is defined by Equation A-4, where d is the degrees of freedom for the evaluation.

- /

(A-4)

If a value of order one or less is obtained for this equation, then there is no reason to doubt the expected (fitted) distribution is reasonable; however, if the value is much larger than one, the expected distribution is unlikely to be a good fit. Results for each trending parameter are summarized in Table 25.

Table 25 - Trending Results Summary Trend Parameter Intercept Slope r2

~

Valid Trend H/X

((

No 235U wt.%

No EALF No 239Pu wt.%

))

No The results in Table 25 clearly demonstrate that there are no statistically significant or valid trends of knorm with any of the trending parameters.

A.2 - Bias and Bias Uncertainty Calculation - Single Sided Tolerance Limit As no trends are apparent in the critical experiment results, a weighted single-sided tolerance limit methodology is utilized to establish the bias and bias uncertainty for this AOA and code package combination. Use of this method requires the critical experiment results to have a normal statistical distribution. This was verified using the Anderson-Darling normality test. A graphical image of the results for this normality test, including the p-value for the distribution, is provided in Figure 19. Because the reported p-value is greater than 0.05, it is confirmed that the data fits a normal distribution, and the single sided tolerance limit methodology is confirmed to be applicable.

NEDO-34125 Revision 0 Non-Proprietary Information 55 t

t

((

))

Figure 19 - Normality Test of knorm Results When using this method, the weighted bias and bias uncertainty are calculated using the following equations:

1 (A-5)

(A-6) knorm

i1

knormi 2

t 1

(A-7)

S P 2

i1 2

n

(A-8)

(A-9) 1 i1 2

n n

n s 2 2

.J

NEDO-34125 Revision 0 Non-Proprietary Information 56

t t

1 n 1 s 2 n 1 i 1 2 1

n 1 (A-10)

Where:

knorm = Average weighted knorm n i 1 2 SP = Pooled standard deviation s2 = Variance about the mean 2 = Average total variance U = one-sided tolerance factor for n data points at (95/95 confidence/probability level) n = number of data points (= ((

)))

Table 26 summarizes the results of these calculations.

Table 26 - Bias and Bias Uncertainty for MCNP-05P with ENDF/B-VII Bias (weighted)

((

Bias Uncertainty (95/95 level)

Variance About the Mean Average Total Variance Pooled Standard Deviation (1)

One-Sided Tolerance Factor

))

Using the average weighted bias and pooled standard deviation; the upper one-sided 95/95-tolerance limit (bias uncertainty) was calculated for use in criticality calculations, in accordance with NUREG/CR-6698 guidance (Reference 6). As seen in Figure 19, ((

)) As shown in Table 26, the MCNP-05P bias uncertainty (95/95) was ((

)) Table 27 summarizes the recommended bias and bias uncertainty to be used in criticality calculations.

Table 27 - Recommended Bias and Bias Uncertainty in Criticality Analyses for MCNP-05P with ENDF/B-VII Bias

((

Bias Uncertainty (95/95)

))

knorm i knorm 2

NEDO-34125 Revision 0 Non-Proprietary Information 57 APPENDIX B - LEGACY FUEL STORAGE JUSTIFICATION Exposure dependent, maximum, uncontrolled in-core k results have been calculated for each fuel assembly in the GGNS fuel storage racks and are confirmed to be less than 1.29. The limit for the highest in-core k value for the bundles currently in use at the GGNS is 1.26. The in-core k values have been calculated using the process for validating that specific assembly designs are acceptable for storage in the GGNS fuel storage racks, as outlined in Section 3.5. The legacy lattice with the highest in-core reactivity value is presented in Table 28. This information demonstrates that all fuel assemblies currently in the GGNS fuel storage racks have considerable margin to the reactivity of the GNF2 design basis lattice used in this analysis.

The GNF2 design basis lattice with an in-core k value of 1.29 has been shown to be below the 10 CFR 50.68 0.95 in-rack limit when analyzed in the storage racks. Because of this, and the fact that the legacy fuel types are sufficiently less reactive than this design basis lattice (see Table 28),

it is confirmed that all legacy fuel bundles are safe for storage in the GGNS fuel storage racks with rack inserts installed.

Table 28 - Limiting SCCG In-Core Eigenvalue of all Legacy GGNS Bundles Bundle Name In-core k

((

))

GNRO2024-00033 NEI 12-16 Appendix C: Criticality Analysis Checklist (8 pages below)

C-1 APPENDIX C: CRITICALITY ANALYSIS CHECKLIST The criticality analysis checklist is completed by the applicant prior to submittal to the NRC. It provides a useful guide to the applicant to ensure that all the applicable subject areas are addressed in the application, or to provide justification/identification of alternative approaches.

The checklist also assists the NRC reviewer in identifying areas of the analysis that conform or do not conform to the guidance in NEI 12-16. Subsequently, the NRC review can then be more efficiently focused on those areas that deviate from NEI 12-16 and the justification for those deviations.

Subject Included Notes / Explanation 1.0 Introduction and Overview Purpose of submittal YES Changes requested YES Summary of physical changes YES Summary of Tech Spec changes NO Not included in the criticality analysis. To be included in a separate license amendment.

Summary of analytical scope YES 2.0 Acceptance Criteria and Regulatory Guidance Summary of requirements and guidance YES Requirements documents referenced YES Guidance documents referenced YES Acceptance criteria described YES 3.0 Reactor and Fuel Design Description Describe reactor operating parameters NO See Section 5.5.1 for discussion.

Describe all fuel in pool YES Section 4.0 and Appendix B Geometric dimensions (Nominal and Tolerances)

NO Section 4.0 for GE14, GNF2, and GNF3 designs only.

Geometric data not provided for legacy fuel.

Schematic of guide tube patterns YES Water rod locations described in Section 4.1 for GE14, GNF2, and GNF3 designs only. Water rod locations not provided for legacy fuel.

Guide tube patterns not applicable for BWR fuel.

Material compositions YES Section 4.0 for GE14, GNF2, and GNF3 designs only.

Material compositions not provided for legacy fuel.

C-2 Subject Included Notes / Explanation Describe future fuel to be covered YES Section 4.0 Geometric dimensions (Nominal and Tolerances)

YES Section 4.0 Schematic of guide tube patterns YES Water rod locations described in Section 4.2. Guide tube patterns not applicable for BWR fuel.

Material compositions YES Section 4.0 Describe all fuel inserts NO There are no fuel inserts in this analysis.

Geometric Dimensions (Nominal and Tolerances)

Schematic (axial/cross-section)

Material compositions Describe non-standard fuel YES Section 4.0 Geometric dimensions Describe non-fuel items in fuel cells YES Section 4.0 Nominal and tolerance dimensions NO Not applicable 4.0 Spent Fuel Pool/Storage Rack Description New fuel vault & Storage rack description NO The proposed change does not include the new fuel storage racks.

Nominal and tolerance dimensions Schematic (axial/cross-section)

Material compositions Spent fuel pool, Storage rack description YES Sections 5.1-5.2 Nominal and tolerance dimensions Schematic (axial/cross-section)

Material compositions Other Reactivity Control Devices (Inserts)

YES Sections 5.1-5.2 Nominal and tolerance dimensions Schematic (axial/cross-section)

Material compositions 5.0 Overview of the Method of Analysis New fuel rack analysis description NO The proposed change does not include the new fuel storage racks.

Storage geometries Bounding assembly design(s)

Integral absorber credit Accident analysis Spent fuel storage rack analysis description YES Section 5.0 and Sections 3.5-3.7 Storage geometries YES Section 5.2 Bounding assembly design(s)

YES Section 5.3 Soluble boron credit NO Not applicable. No soluble boron is used at GGNS.

Boron dilution analysis

C-3 Subject Included Notes / Explanation Burnup credit NO No burnup credit in BWR peak reactivity analysis. Fuel is evaluated at peak reactivity.

Decay/Cooling time credit NO No decay/cooling time credit.

Integral absorber credit YES Section 5.3 Other credit NO No other credit.

Fixed neutron absorbers YES Credit for neutron absorbing inserts.

Aging management program YES Accident analysis YES Section 5.5.3 Temperature increase YES Section 5.5.3 and Section 5.4.1 Assembly drop YES Section 5.5.3 Single assembly misload NO Uniform pool with peak reactivity fuel, so no opportunity for misload.

Multiple misload NO Boron dilution NO Not applicable. No soluble boron is used at GGNS.

Other YES Section 5.5.3 Fuel out of rack analysis YES Section 5.5.2 Handling Movement Inspection 6.0 Computer Codes, Cross Sections and Validation Overview Code/Modules Used for Calculation of keff YES Described in Section 3.0.

Cross section library YES Section 3.1 Description of nuclides used YES Section 4.4 Convergence checks YES Section 3.3 Code/Module Used for Depletion Calculation YES Described in Section 3.0 Cross section library YES Section 3.1 Description of nuclides used YES Sections 3.7 and 4.4 Convergence checks YES Section 3.3 Validation of Code and Library YES Section 3.4 and Appendix A Major Actinides and Structural Materials YES Section 3.4 Minor Actinides and Fission Products YES Section 3.4 Absorbers Credited YES Section 3.4 7.0 Criticality Safety Analysis of the New Fuel Rack Rack model NO Boundary conditions Source distribution Geometry restrictions

C-4 Subject Included Notes / Explanation Limiting fuel design Fuel density Burnable Poisons Fuel dimensions Axial blankets Limiting rack model Storage vault dimensions and materials Temperature Multiple regions/configurations Flooded Low density moderator Eccentric fuel placement Tolerances Fuel geometry Fuel pin pitch Fuel pellet OD Fuel clad OD Fuel content Enrichment Density Integral absorber Rack geometry Rack pitch Cell wall thickness Storage vault dimensions/materials Code uncertainty Biases Temperature Code bias Moderator Conditions Fully flooded and optimum density moderator 8.0 Depletion Analysis for Spent Fuel Depletion Model Considerations YES Described in Section 3.3, Section 3.7, and Section 4.4.

Time step verification Convergence verification Simplifications Non-uniform enrichments Post Depletion Nuclide Adjustment Cooling Time Depletion Parameters Burnable Absorbers Integral Absorbers Soluble Boron

C-5 Subject Included Notes / Explanation Fuel and Moderator Temperature Power Control rod insertion Atypical Cycle Operating History 9.0 Criticality Safety Analysis of Spent Fuel Pool Storage Racks Rack model YES Section 5.2 Boundary conditions Source distribution Geometry restrictions Design Basis Fuel Description YES Section 5.3 Fuel density YES Section 4.0 Burnable Poisons YES Section 5.3 Fuel assembly inserts NO No fuel assembly inserts in this analysis.

Fuel dimensions YES Section 4.1, Section 4.2, and Section 4.3 Axial blankets NO Section 3.7 Configurations considered YES Single configuration, uniform pool. See Section 6.0.

Borated NO Not applicable for this analysis.

Unborated YES Multiple rack designs NO Not applicable. One rack design with inserts in every location.

Alternate storage geometry NO Not applicable for this analysis.

Reactivity Control Devices YES Fuel Assembly Inserts NO No fuel assembly inserts in this analysis.

Storage Cell Inserts YES Section 5.1 Storage Cell Blocking Devices NO No cells are required to be empty, so no blocking devices are considered in this analysis.

Axial burnup shapes NO Section 3.7 Uniform/Distributed YES Nodalization NO Blankets modeled NO Tolerances/Uncertainties YES Section 5.6 Fuel geometry Fuel rod pin pitch Fuel pellet OD Cladding OD

C-6 Subject Included Notes / Explanation Axial fuel position NO Section 3.7 Fuel content YES Section 5.6 Enrichment Density Assembly insert dimensions and materials NO No fuel assembly inserts in this analysis.

Rack geometry YES Section 5.6 Flux-trap size (width)

NO Not applicable Rack cell pitch YES Section 5.6.1 Rack wall thickness NO Section 5.6.1 Neutron Absorber Dimensions NO Not applicable because Boraflex is modeled as water.

See Section 5.6.1 Rack insert dimensions and materials YES Section 5.6.1 Code validation uncertainty YES Described in Section 3.4 and Section 5.6.2.

Criticality case uncertainty YES Section 5.6.2 Depletion Uncertainty YES Described in Section 3.4 and Section 5.6.2.

Burnup Uncertainty NO Not applicable for BWR peak reactivity analysis.

Biases YES Section 5.0 Design Basis Fuel design YES Section 5.3 Code bias YES Section 3.4 and Section 5.5.4 Temperature YES Section 5.4 and Section 5.5.4 Eccentric fuel placement YES Not applicable. See Section 5.4.1.

Incore thimble depletion effect NO Not applicable for this analysis.

NRC administrative margin NO Not applicable for this analysis.

Modeling simplifications YES Sections 3.7 and 4.4 Identified and described 10.0 Interface Analysis Interface configurations analyzed NO Not applicable because the pool is uniform with rack inserts in every cell. See Section 6.0.

Between dissimilar racks NO Between storage configurations within a rack NO Interface restrictions NO None

C-7 Subject Included Notes / Explanation 11.0 Normal Conditions Fuel handling equipment YES Section 5.5.2 Administrative controls YES Defective fuel storage locations are procedurally not allowed to have fuel stored in the edge tubes. Fuel can only be stored in the interior tubes.

Fuel inspection equipment or processes YES Section 5.5.2 Fuel reconstitution YES Replaced rods are covered, but storage of assemblies with missing pins is not allowed.

See Section 4.0.

12.0 Accident Analysis Boron dilution NO Not applicable. No soluble boron used at GGNS.

Normal conditions Accident conditions Single assembly misload NO Uniform pool with peak reactivity fuel, so no opportunity for misload.

Fuel assembly misplacement YES Section 5.5.3 Neutron Absorber Insert Misload YES Section 5.5.2 Multiple fuel misload NO Uniform pool with peak reactivity fuel, so no opportunity for misload.

Dropped assembly YES Section 5.5.3 Temperature YES Section 5.5.3 Seismic event/other natural phenomena YES Section 5.5.3 13.0 Analysis Results and Conclusions Summary of results YES Sections 5.7 and 7.0 Burnup curve(s)

NO Not applicable for BWR peak reactivity analyses.

Intermediate Decay time treatment NO Not applicable for BWR peak reactivity analyses. See Section 4.4.

New administrative controls YES Fuel with missing fuel rods shall not be loaded into a spent fuel rack cell. If not already present, this administrative control needs to be added in the site fuel movement procedure(s).

Technical Specification markups YES

C-8 Subject Included Notes / Explanation 14.0 References YES Section 8.0.

Appendix A: Computer Code Validation:

Appendix A.

Code validation methodology and bases YES Appendix A New Fuel Depleted Fuel MOX HTC Convergence Trends Bias and uncertainty Range of applicability YES Described in Section 3.4.

Analysis of Area of Applicability coverage YES Described in Section 3.4.

GNRO2024-00033 Global Nuclear Fuels - Americas Proprietary Information Affidavits (4 pages below)

NEDC-34125P Revision 0 GNF Proprietary Information - Non-Public PROPRIETARY INFORMATION NOTICE This document contains proprietary information of Global Nuclear Fuel - Americas, LLC (GNF).

and is furnished in confidence solely for the purpose(s) stated below in the notice regarding the contents of this report. No other use, direct or indirect, of the document or the information it contains is authorized. The recipient shall not publish or otherwise disclose this document or the information therein to others without the prior written consent of GNF and shall return the document at the request of GNF.

The header of each page in this document carries the notation, GNF Proprietary Information -

Non-Public.

GNF proprietary information within the text and tables is identified by a dotted underline inside double square brackets. ((This sentence is an example.{3})) GNF proprietary information in figures and large objects is identified by double square brackets before and after the object. In all cases, the superscript notation {3} refers to Paragraph (3) of the enclosed affidavit that provides the basis for the proprietary determination.

Curtiss-Wright Flow Control Service, LLC (CW) information is identified by a solid underline inside double square brackets. ((This sentence is an example.{C})) CW proprietary information in figures and large objects is identified by double square brackets before and after the object. In all cases, the superscript notation {C} refers to the enclosed affidavit that provides the basis for the proprietary determination.

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The design, engineering, and other information contained in this document is furnished for the purpose of providing the results of the fuel storage rack criticality analysis for Grand Gulf Nuclear Station. The only undertakings of GNF with respect to information in this document are contained in the contracts between Entergy and GNF, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than Entergy, or for any purpose other than that for which it is furnished by GNF is not authorized; and with respect to any unauthorized use, GNF makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

ii

NEDC-34125P Revision 0 Affidavit Page 1 of Global Nuclear Fuel - Americas, LLC AFFIDAVIT I, Lisa K. Schichlein, state as follows:

(1)

I am a Senior Licensing Engineer, Regulatory Affairs, Global Nuclear Fuel -

Americas, LLC (GNF), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld and have been authorized to apply for its withholding.

(2)

The information sought to be withheld is contained in GNF proprietary report, NEDC-34125P, Grand Gulf Nuclear Station: Fuel Storage Criticality Safety Analysis of Spent Fuel Storage Racks with Rack Inserts, Revision 0, July 2024.

GNF proprietary information within the text and tables is identified by a dotted underline placed within double square brackets. ((This sentence is an example.{3}))

Figures and large objects containing GNF proprietary information are identified with double square brackets before and after the object. In all cases, the superscript notation {3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3)

In making this application for withholding of proprietary information of which it is the owner or licensee, GNF relies upon the exemption from disclosure set forth in the Freedom of Information Act (FOIA), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of trade secret, within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F2d 871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F2d 1280 (DC Cir. 1983).

(4)

Some examples of categories of information which fit into the definition of proprietary information are:

a.

Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GNF's competitors without license from GNF constitutes a competitive economic advantage over other companies;

b.

Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c.

Information which reveals aspects of past, present, or future GNF customer-funded development plans and programs, resulting in potential products to GNF;

d.

Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

NEDC-34125P Revision 0 Affidavit Page 2 of The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. above.

(5)

To address 10 CFR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GNF, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GNF, no public disclosure has been made, and it is not available in public sources.

All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6)

Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GNF.

(7)

The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GNF are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8)

The information identified in paragraph (2), above, is classified as proprietary because it contains details of GNFs fuel design and licensing methodology. The development of this methodology, along with the testing, development and approval was achieved at a significant cost to GNF or its licensor.

The development of the fuel design and licensing methodology along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GNF asset.

(9)

Public disclosure of the information sought to be withheld is likely to cause substantial harm to GNF's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GNF's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

NEDC-34125P Revision 0 Affidavit Page 3 of The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by GNF.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GNF's competitive advantage will be lost if its competitors are able to use the results of the GNF experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GNF would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GNF of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on this 9th day of July 2024.

Lisa K. Schichlein Senior Licensing Engineer Regulatory Affairs Global Nuclear Fuel - Americas, LLC 3901 Castle Hayne Road Wilmington, NC 28401 Lisa.Schichlein@ge.com

GNRO2024-00033 Curtiss-Wright Nuclear Division Proprietary Information Affidavits (2 pages below)

CURTISS-WRIGHT AFFIDAVIT PURSUANT TO 10 CFR 2.390 I, Karl Scot Leuenroth, depose and say that I am the Division Manager of Curtiss-Wrights Scientech Division, duly authorized to make this affidavit, and have reviewed or caused to have reviewed the information which is identified as proprietary and referenced in the paragraph immediately below.

I am submitting this affidavit in conformance with the provisions of 10 CFR 2.390 of the Commission's regulations for withholding Curtiss-Wrights information for which proprietary treatment is sought as contained in NEDC-34125P, "Grand Gulf Nuclear Station: Fuel Storage Criticality Safety Analysis with Rack Inserts," Revision 0, July 2024.

I have personal knowledge of the criteria and procedures utilized by Curtiss-Wright in designating information as a trade secret, privileged or as confidential commercial or financial information.

Pursuant to the provisions of paragraph (b) (4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure, included in the above referenced document, should be withheld.

1)

The information sought to be withheld from public disclosure is a list technical information related to the Snap-In Insert technology, which involve considerable research and development of intellectual property by Curtiss-Wright. Curtiss-Wright Flow Control Service, LLC (CW) information is identified by a solid underline inside double square brackets. ((This sentence is an example.{C})) CW proprietary information in figures and large objects is identified by double square brackets before and after the object.

2)

The information is of a type customarily held in confidence by Curtiss-Wright, and not customarily disclosed to the public. Curtiss-Wright has a rational basis for determining the types of information customarily held in confidence by it.

3)

The information is being transmitted to the Commission in confidence under the provisions of 10 CFR 2.390 with the understanding that it is to be received in confidence by the Commission.

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The information, to the best of my knowledge and belief, is not available in public sources, and any disclosure to third parties has been made pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.

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A similar product is manufactured and sold by competitors of Curtiss-Wright.

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Development of this information by Curtiss-Wright required expenditure of considerable resources. To the best of my knowledge and belief, a competitor would have to undergo similar expense in generating equivalent information.

c)

In order to acquire such information, a competitor would also require considerable time and inconvenience related to the development of a design and analysis of a similar neutron attenuation technology for use in a spent fuel pool.

d)

The availability of such information to competitors would enable them to modify their product to better compete with Curtiss-Wright, take marketing or other actions to improve their product's position or impair the position of Curtiss-Wrights product, and avoid developing similar data and analyses in support of their processes, methods or apparatus.

Leuenroth,Scot 2024.07.1714:52:22

04'00' Karl Scot Leuenroth

v t e Letter Sequence Request
EPID:L-2024-LLA-0158, Submittal of an Amendment Request for the NAC International Magnastor Cask System Amendment No. 14 (Open)
Initiation Request, Request, Request, Request Acceptance, Acceptance Supplement, Supplement Administration Withholding Request Acceptance Questions 7 May 2024 Answers Results Approval... Other:
MONTHYEAR2024-05-07 [Table View]

ML24330A234

Text

SUBJECT:

Enclosure:

SUMMARY

SUMMARY

Enclosure:

1.0 INTRODUCTION

7.0 CONCLUSION

8.0 REFERENCES

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