ML25029A237
| ML25029A237 | |
| Person / Time | |
|---|---|
| Site: | 07201015 |
| Issue date: | 11/18/2025 |
| From: | Garcia-Santos N Division of Fuel Management, Storage and Transportation Licensing Branch |
| To: | Baldner H NAC International |
| References | |
| CAC 001028 | |
| Download: ML25029A237 (14) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 OFFICIAL USE ONLY - PROPRIETARY INFORMATION PRELIMINARY SAFETY EVALUATION REPORT NAC International, Inc.
Model No. NAC-UMS Universal Storage System Docket No. 72-1015 Renewed Amendment Nos. 5 To 9, Revision No. 1 Renewed Amendment 10 OFFICIAL USE ONLY - PROPRIETARY INFORMATION
OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Table of Contents Page
SUMMARY
1 1.0 GENERAL INFORMATION...2 2.0 STRUCTURAL EVALUATION..3 2.1 Evaluation of Fuel Rods for the Tip-Over Structural Analysis.4 2.2 Evaluation of the PWR Fuel Assembly in the NAC-UMS under End Drop Condition.7 2.3 Evaluation Findings...9 3.0 AGING MANAGEMENT.9 4.0 CONDITIONS..9 5.0 TECHNICAL SPECIFICATIONS..10
6.0 REFERENCES
....10 CONCLUSION..12 Tables Table 2.1.1 Calculated Maximum Stress and Margin of Safety under Tip-Over Accident Condition...................................................................................................................................... 5 Table 2.1.2 Calculated Maximum Stress and Margin of Safety for PWR Tip-Over Accident Condition...................................................................................................................................... 7 Table 2.2 Calculated Maximum Stress Intensity and Margin of Safety under the End Drop Accident Condition....................................................................................................................... 7
OFFICIAL USE ONLY - PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 OFFICIAL USE ONLY - PROPRIETARY INFORMATION PRELIMINARY SAFETY EVALUATION REPORT NAC INTERNATIONAL, INC.
NAC-UMS UNIVERSAL STORAGE SYSTEM DOCKET NO. 72-1015 AMENDMENT NOS. 5 to 9, REVISION NO. 1 AMENDMENT NO. 10
SUMMARY
By application dated October 10, 2023 (NAC, 2023b and NAC, 2023c), as supplemented on February 13, 2024 (NAC, 2024a), June 26, 2024 (NAC, 2024b), July 17, 2025 (NAC, 2025a),
and August 27, 2025 (NAC, 2025b), NAC International, Inc. (NAC or the applicant) submitted a request to the United States (U.S.) Nuclear Regulatory Commission (NRC) to revise the licensing basis of Amendments Nos. 5 through 9 and issue Amendment No. 10 of the Model No.
NAC-UMS spent fuel storage system (NAC-UMS thereafter) in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 72.244 to amend certificate of compliance (CoC) No.
1015.
The applicant requested the following changes to correct licensing deficiencies identified in the letters dated March 10, 2023 (NAC, 2023a) and July 17, 2025 (NAC, 2025a):
- 1)
Revise a parameter used in the computation of bending stress in the finite element model used to structurally evaluate a fuel rod under the non-mechanistic tip-over accident condition.
- 2)
Update the pressurized water reactor (PWR) end drop evaluation using the methodology already approved for the boiling water reactor (BWR) end drop evaluation.
- 3)
Change the address in the CoC to reflect the new address of the applicants headquarters offices.
The applicant submitted page changes labeled as Revision 23A, 24A, 25A, and 25B of the safety analysis report (SAR) for the proposed changes to Renewed Amendment Nos. 5 through 9 of the Model No. NAC-UMS. The applicant noted in its application that these changes did not result in changes to aging management related to the renewed CoC No. 1015 for the Amendment Nos. 5 to 9 of the NAC-UMS. Staff evaluated the proposed Renewed Amendment Nos. 5 through 9, Revision 1, and Renewed Amendment No. 10 and agree with applicants conclusion that these changes do not require modification of the existing aging management activities or change the staffs prior conclusions for the renewal of the NAC-UMS CoC No.
1015 (NRC 2024b). In section 3.0 of this SER, the staff evaluated whether the proposed CoC amendment and revisions would change the staffs prior conclusions for the renewal of CoC No.
1015 (NRC, 2024b and NRC, 2024c). The staff found that the changes do not affect the ability of the NAC-UMS to meet the requirements of Part 72.
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION 1.0 GENERAL INFORMATION NAC requested that the changes mentioned in this section of the SER be made in a new Amendment No. 10 to the CoC, and that these changes also be included in the existing Amendment Nos. 5 through 9 via a revision to these amendments. The amended CoC, when codified through rulemaking, will be denoted as Renewed Amendment No. 10 to CoC No. 1015.
The revised CoCs, when codified through rulemaking, will be denoted as Renewed Amendment No. 5, Revision 1; Renewed Amendment No. 6, Revision 1; Renewed Amendment No. 7, Revision 1; Renewed Amendment No. 8, Revision 1; and Renewed Amendment No. 9, Revision 1 to CoC No. 1015.
This SER documents the staffs review of proposed Renewed Amendment Nos. 5 through 9, Revision 1, and Amendment No. 10, to CoC No. 1015 for the NAC-UMS, with the exception of the change in address because it is an administrative change that does not affect the performance of the NAC-UMS. The NRC staff (the staff thereafter) determined that the following areas of review are not affected by this amendment and, therefore, not addressed in this SER:
- 1) general description,
- 2) principal design criteria,
- 3) confinement evaluation,
- 4) shielding evaluation,
- 5) criticality safety evaluation,
- 6) materials evaluation,
- 7) operating procedures,
- 8) acceptance tests and maintenance program,
- 9) radiation protection, and
- 10) accident analyses.
The staff reviewed the amendment request using guidance in NUREG-2215, Standard Review Plan for Dry Cask Storage Systems and Facilities - Final Report, Revision 0 (NRC, 2020). The NRC staff also used the guidance in NUREG-2214, Managing Aging Processes in Storage (MAPS) Report, dated July 2019 (NRC, 2019), as the NAC-UMS System is a renewed storage system.
For the reasons stated below and based on the review of the statements and representations in the application, and the conditions specified in the CoC and technical specifications, the staff concludes that the requested changes meet the requirements of 10 CFR Part 72.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 2.0 STRUCTURAL EVALUATION The purpose of the structural evaluation is to ensure that the structural integrity of structures, systems, and components (SSCs) of the NAC-UMS dry storage system (DSS) maintains their safety functions under all credible loads and their combinations for normal, off-normal, and accident conditions and natural phenomena effects. This section describes the staffs evaluation of the proposed changes described in the Summary section of this SER and the staffs rationale for reaching reasonable assurance that the NAC-UMS DSS will maintain its intended function.
The applicant indicated that a parameter used in the structural analysis to evaluate a fuel rod under the non-mechanistic tip-over accident condition was incorrectly specified in the analyses that constitutes the licensing basis for Amendment Nos. 5 through 9 of the NAC-UMS resulting in the non-conservative calculation of stress values. Furthermore, the applicant indicated that a cladding material property adjustment was necessary for the PWR cask tip-over evaluation. The applicant corrected the errors and revised the structural evaluations for the PWR and BWR fuel rods in section 11.2.16, Fuel Rods Structural Evaluation for Burnup to 62,500 MWd/MTU,1 of the NAC-UMS SAR (NAC, 2023b and NAC, 2023c). Additionally, the applicant submitted an updated PWR fuel rod end drop evaluation, which also incorporated the cladding material property adjustment (NAC, 2025a), using the structural methodology approved for the end drop evaluation of the BWR fuel authorized as content for the NAC-UMS.
The staff reviewed the amendment application and issued a request for additional information (RAI) (NRC, 2024a). The RAI focused on:
- 1) the parameter that was incorrectly specified in the computation of bending stress in the ANSYS finite element (FE) model for a fuel rod under the tip-over accident condition, and
The staff evaluated the following to verify that the structural performance of the PWR and BWR fuel rods contained in the NAC-UMS meet the requirements of 10 CFR Part 72:
- 1) proposed changes (NAC, 2023b),
- 2) the structural calculation packages (NAC, 2023c), and
The following sections include the staffs evaluation of the proposed changes with respect to structural safety of the NAC-UMS.
1 Megawatts per day per metric ton of uranium.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 4
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 2.1 Evaluation of Fuel Rods for the Tip-Over Structural Analysis 2.1.1 Evaluation of the Corrected ANSYS Parameter for the Tip-Over Structural Analysis for PWR and BWR Fuel Rods By letter dated March 10, 2023 (NAC, 2023a), the applicant reported that it identified a licensing basis deficiency for the NAC-UMS (NAC, 2023b). The applicant identified that a parameter used in the computation of bending stress in the ANSYS FE model to evaluate a fuel rod under the tip-over accident condition was incorrect. As a result, the tip-over structural analysis resulted in a non-conservative calculation of stress values.
The staff reviewed the responses to the RAIs, in which the applicant stated that the parameter that was incorrectly specified was the height of the cross-section (h) of the 2-D elastic beam element used to compute the bending stress in the ANSYS model. This element parameter is set using the R command in ANSYS, which defines the element Real Constants (NAC, 2024b).
The ANSYS FE computer program (ANSYS thereafter) calculates a bending stress using the following equation:
= x
= x
x, where b = bending stress, M = moment causing the bending, c = distance from neutral axis to outer fiber (radius of the fuel rod) = h/2, h = full height of beam cross-section (outer diameter of the fuel rod), and I = moment of inertia.
The applicant noted that the value of c in the ANSYS R command was incorrectly entered in the ANSYS input. The height real constant (h), input using the R command for the 2-D elastic beam element used in the ANSYS FE model, is defined to be the full height of the beam, or in this case, the outer diameter of the fuel rod (i.e., h), which is then used by ANSYS for the computation of the bending stress of the element. However, the applicant incorrectly set the value of h for the R command as h/2, or the outer radius of the fuel rod, which is one half of the correct value. As a result, the applicant incorrectly calculated the bending stress as one half of the correct stress.
The applicant further noted that, except for this input error of the R command all other parameters (e.g., the fuel rod diameter, clad thickness, etc.) and material properties (e.g., density, modulus of elasticity) in the ANSYS input file were correctly defined. The applicant correctly determined and specified element cross-sectional area (A) and moment of inertia () in the same ANSYS R command (R, A, I, h). The incorrect input has no effect on the mass or stiffness of the element. The applicant calculated the ANSYS FE solution for the element forces and moments correctly.
The applicant corrected the value of the R command in the ANSYS FE model and performed new analyses to assess the potential for fuel rod yielding in a non-mechanistic tip-over event for BWR fuel assemblies and PWR fuel assemblies with partially damaged fuel grids. The new analyses, presented in NAC Calculations EA790-2520, Rev. 1, BWR Fuel Assembly Structural Evaluation, and EA790-2519, Rev. 0, PWR Fuel Rod Evaluation for the UMS Storage Tip-
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION Over Accident, (NAC, 2023b), represented the tip-over loading as a combination of the peak acceleration from the response spectrum corresponding to the acceleration time history (ATH) at the top of the fuel basket from an LS-DYNA tip-over analysis and applying this, conservatively, to the entire length of the fuel rod using a dynamic load factor (DLF) based on the modal analysis of the fuel rod.
The results of this new analysis demonstrated that the stress in the BWR and PWR fuel rod cladding remains below the rod material allowable yield stress in a non-mechanistic tip-over event and supports the existing criteria of less than or equal to 60 inches of unsupported rod length (i.e., loss of grid strap support) for PWR, allowing PWR fuel assemblies with grid damage to not require placement into damaged fuel cans.
The staff reviewed the documents submitted and summarized the stresses as well as the margin of safety (MS) in Table 2.1.1 of this SER, where the MS is defined as the factor of safety (FS) minus one (MS = FS -1.0). The MS shown in the table is for the bounding PWR and BWR fuel assembly (the fuel assembly that has the minimum MS under the tip-over accident condition), and both have values greater than zero.
The results of the analysis provided by the applicant showed that the cladding materials have adequate allowable design margins in strength under the tip-over accident condition.
Based on the reviews of the statements, analyses and evaluations provided in the amendment application and the responses to the RAI, the staff determined that (i) the revision of the incorrectly specified ANSYS FE parameter to compute the bending stress of the fuel rod is acceptable, and (ii) the fuel rod will not rupture under the non-mechanistic tip-over accident condition.
Table 2.1.1 Calculated Maximum Stress and Margin of Safety under Tip-Over Accident Condition.
[Proprietary Information, Table Withheld per 10 CFR 2.390.]
2.1.2 Evaluation of the PWR Fuel Rods with Updated Material Property for the Tip-Over Event The applicant submitted a revised calculation (EA790-2519, Revision 1) for the PWR cask tip-over event (NAC, 2025a) to update a material property employed in the evaluation. The applicant performed a conservative evaluation to represent the effects of this tip-over accident event on 26 PWR (11 fuel types) high burnup fuel assemblies, at a bounding fuel temperature of 400 °C (752 °F), the bounding maximum fuel temperature for normal storage conditions.
For the PWR rod analyses, the applicant considered high burnup conditions and cladding alloys that are typical for PWR fuel. The fuel rod and fuel pellet material properties of elastic modulus, density and yield strength, as applicable, are taken from various references. The staff reviewed
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 6
OFFICIAL USE ONLY - PROPRIETARY INFORMATION the applicants selection of material properties for the cladding alloys to support the analyses. In addition, the staff reviewed information on cladding material properties not cited by the applicant including:
Shimskey, R., et al. FY2014 PNNL Zr Cladding Testing Status Report, PNNL-23594, August 30, 2014.
Wells, B.E., et al. Evaluation of Increased Peak Temperatures for Spent Fuel Cladding Performance during Dry Storage, PNNL-30430, Rev. 1, September 2020.
The staff determined that the material properties for most of the PWR fuel cladding alloys used by the applicant were obtained by measurements on the irradiated cladding samples at elevated temperatures and were therefore acceptable.
The staff noted that the data used by the applicant for one of the PWR cladding alloys was obtained using tensile tests under non-quasi-static testing conditions. Zirconium-based fuel cladding alloys are known to strain harden as a function of strain rate which, in turn, increases the measured yield strength: higher strain rate shows higher yield strength. Based on the information provided in Shimskey et al. (2014) and Wells et al., (2020), the staff determined that the yield strength of the cladding alloy would be increased by approximately 6 percent under the testing conditions in the reference cited by the applicant. The staff also determined that publicly available data for irradiated cladding material properties at elevated temperatures is limited and alternative references for properties for the alloy under quasi-static testing conditions are not available. After reviewing the available data and the applicants analysis, the staff determined that the cited material properties for the cladding alloy were acceptable because: (1) the measurements were conducted using irradiated materials over a range of temperatures that include the applicants analyzed maximum temperature for the cladding alloy, (2) the strain rates under tip-over accident conditions would be greater than quasi-static strain rates typically used to determine material properties, and (3) use of yield strength as an acceptance criteria is conservative because all zirconium-based cladding alloys strain harden above the yield stress and retain measurable ductility after irradiation.
The applicant determined the peak deceleration at the basket top support disk in the storage cask during the tip-over event using the LS-DYNA finite element program. The applicant performed a modal analysis of each fuel rod, following the general principles presented in section 2.3.5.2 of NUREG-2224, to determine a dynamic load factor versus the tip-over peak deceleration. The applicant conservatively applied the resulting acceleration value of 24.5 g uniformly along the entire length of the damaged PWR fuel rods using the ANSYS program to determine the maximum fuel rod stresses from the tip-over accident.
For the PWR rod analyses, the margin of safety for all cladding stresses are greater than a value of zero for all material types, indicating that the fuel rod cladding material meets the acceptance criteria during the tip-over accident event. Based on these results, staff finds the fuel rods to be structurally adequate for this accident event per by 10 CFR 72.236(l).
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION Table 2.1.2 Calculated Maximum Stress and Margin of Safety for PWR Tip-Over Accident Condition.
[Proprietary Information, Table Withheld per 10 CFR 2.390.]
2.2 Evaluation of the PWR Fuel Assembly in the NAC-UMS under End Drop Condition The applicant also reanalyzed and evaluated the PWR fuel assemblies approved as contents of the NAC-UMS cask under the 24-inch end drop condition for the corrected ANSYS parameter and the updated material property and updated SAR subsection 11.2.16.1, PWR Fuel Rod Evaluation. The purposes of this structural analysis are the following:
- a.
to demonstrate that the maximum stress in high burnup PWR fuel remains below the yield strength in the design basis end drop accident condition, and
- b.
confirm that the PWR fuel rods will return to their original configuration following the end drop event.
The applicant considered four different cases as shown in Table 2.2 below and performed the structural analysis using the ANSYS and LS-DYNA FE computer programs for the 24-inch cask end drop condition at a bounding fuel temperature of 350 °C (662 °F). These cases envelop the range of the cross-sectional moments for the PWR fuel rods and the grid spacing at the bottom of the fuel assembly.
Table 2.2 Calculated Maximum Stress Intensity and Margin of Safety under the End Drop Accident Condition.
[Proprietary Information, Table Withheld per 10 CFR 2.390.]
Based on the result of the analysis as shown in Table 2.2, the applicant concluded that
- a.
the bounding case uses the 60-inch spacing in conjunction with the minimal cross-section (Case 3),
- b.
all stresses are shown to be less than the yield strength, and
- c.
the results of the analysis confirm that high burnup PWR fuel with a maximum distance of 60 inches from the bottom to the first grid will remain structurally adequate for the storage design basis cask end drop condition.
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION The staff reviewed the results of the PWR end drop analysis presented in Calculation No.
71160-2026 (NAC, 2023c, 2025a) and compared them with the results of the previous structural analysis documented in the SAR, Revision 15. The staff found that three grid spacings (25-inch, 33-inch, and 60-inch) were considered and the 33-inch grid spacing was identified as the bounding case in the previous analysis, while two grid spacings (29.6-inch and 60-inch) are considered and the 60-inch grid spacing is identified as the bounding case in the current analysis.
The staff issued an RAI and requested the applicant to provide detailed information regarding the differences in the analyses (i.e., critical bounding grid spacing of 33-inch vs. 60-inch, different stress results with MS, etc.) (NRC, 2024a). The applicant explained in their response that the differences are due to the modeling methodology. The methodology for the PWR fuel rod end-drop evaluation in the NAC Calculation No. 71160-2026 of the SAR, Rev. 23A, in this application significantly changed from the methodology used in the Final SAR (FSAR), Rev. 15.
The applicant used the latest acceptable and consistent methodology for both the PWR and BWR fuel rod end drop evaluations in subsection 11.2.16.1, PWR Fuel Rod Evaluation, and 11.2.16.2, BWR Fuel Rod Evaluation. In the FSAR, Rev. 15, the applicant modelled the pellet and clad independently and applied the maximum bow possible for the 14x14 of 1.23 inches to the FE model for 17x17 model. In the NAC Calculation No. 71160-2026 of the FSAR, Rev. 23A, the applicant added the pellet mass to the clad without any stiffening effect of the pellet.
Additionally, in the NAC Calculation No. 71160-2026 of the FSAR, Rev. 23A, the applicant applied a bow of 0.55 inch to the 17x17 as opposed to the overly conservative 1.23 inches in the FSAR, Rev. 15. In the NAC Calculation No. 71160-2026 (NAC, 2023c), the applicant provided its technical assumption of using the maximum bow of 0.55 inch. The maximum bow of 0.55 inch is based on the study in the reference (Williamson, 1994).
The staff reviewed the responses to the RAI and found them acceptable because the modeling methodology taken by the applicant for the PWR fuel rod end drop analysis is an acceptable methodology (i.e., adding pellet mass to the clad without any stiffening effect of the pellet) in engineering practice, which the staff has previously reviewed and accepted for the PWR and BWR fuel rod end drop analysis in NAC, 2023c. In addition, the results of the structural analysis with the assumption of the 0.55-inch bow for a PWR fuel rod end drop event are acceptable because of the following reasons:
- a.
the actual bow size of the fuel rod is likely less than 0.55-inch based on the study in the reference (Williamson, 1994), thereby, the actual stresses (e.g., buckling stress, bending stress) would likely be smaller than the stresses calculated in the analysis,
- b.
the computed maximum stress with the conservative value of the 0.55-inch bow used in the analysis is less than the yield strength of the cladding material, and
- c.
all calculated MS are positive and large indicating that the fuel rod has design strength under the 24-inch end drop condition.
For the PWR rod analyses, the applicant considered high burnup conditions and cladding alloys that are typical for PWR fuel. The fuel rod and fuel pellet material properties of elastic modulus, density and yield strength, as applicable, are taken from various references, as described in SER section 2.1.2, above.
After reviewing the available data and the applicants analysis, the staff determined that the cited material properties for the cladding alloy were acceptable because: (1) the measurements
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 9
OFFICIAL USE ONLY - PROPRIETARY INFORMATION were conducted using irradiated materials over a range of temperatures that include the applicants analyzed maximum temperature for the cladding alloy, (2) the strain rates under drop accident conditions would be greater than quasi-static strain rates typically used to determine material properties, and (3) use of yield strength as an acceptance criteria is conservative because all zirconium-based cladding alloys strain harden above the yield stress and retain measurable ductility after irradiation.
The fuel rod cladding bounding stresses resulting from the drop analyses were determined by applying an acceleration time history in LS-DYNA of a duration considered sufficient to capture the response of the fuel rods.
For the PWR rod analyses results presented in Table 2.2, the margin of safety for all fuel rod cladding stresses are greater than a value of zero for the bounding material type, indicating that all fuel rod cladding materials meet the acceptance criteria during the storage cask 24-inch end drop accident event for PWR fuel rods. Based on these results, staff finds the fuel rods to be structurally adequate for this accident event per 10 CFR 72.236(l).
Based on the reviews of the statements, analyses and evaluations provided in the amendment application and the responses to the RAI, the staff determined that the structural evaluation for the high burnup PWR fuel rods contained in the NAC-UMS cask adequately demonstrates that they will retain their structural integrity during the storage design basis cask end drop load condition.
2.3 Evaluation Findings
The staff reviewed and evaluated the applicants statements, representations in the amendment application, and responses to the RAI. Based on the reviews and evaluations, the staff concludes that the PWR and BWR fuel rods contained in the NAC-UMS cask are adequately analyzed and evaluated to demonstrate that their structural capability and integrity meet the regulatory requirements of 10 CFR Part 72.
3.0 AGING MANAGEMENT As described in the renewed CoC No. 1015, Condition 11, AMENDMENTS AND REVISIONS FOR RENEWED CoC, (A)ll future amendments and revisions to this CoC shall include evaluations of the impacts to aging management activities (i.e., time-limited aging analyses and aging management programs) to ensure they remain adequate for any changes to SSCs within the scope of the CoC renewal. The applicant stated in their submission letter (NRC, 2023b) that the proposed amendments do not introduce any new components or materials that would need to be evaluated for aging management. The staff reviewed the proposed Renewed Amendment Nos. 5 through 9, Revision 1, and Renewed Amendment No. 10 and agree that there are no changes that would require modification of the existing aging management activities or would change the staffs prior conclusions for the renewal of the NAC-UMS CoC No. 1015 (NRC 2024b).
4.0 CONDITIONS The following conditions were revised in the renewed Amendment Nos. 5 through 9:
The revision No. was increased to 1.
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 10 OFFICIAL USE ONLY - PROPRIETARY INFORMATION Added condition No. 12 to each certificate of compliance for renewed Amendment Nos. 5 through 9, as follows:
CONTINUED USE OF PREVIOUS VERSION OF RENEWED AMENDMENT NOS. [5 through 9, Revision 0]:
A general licensee may continue to use the previous version of this certificate, renewed Amendment No. [applicable renewed amendment No. 5 through 9, Revision 0], dated July 15, 2024, until {insert date 6 months after effective date}. By {insert date 6 months after effective date}, general licensees using renewed Amendment No. [applicable renewed amendment No. 5 through 9, Revision 0], must have implemented the changes authorized by this revision and completed the evaluation described below.
The general licensee shall perform written evaluations before use and before applying the changes authorized by this revised certificate which establish that the cask, once loaded with spent fuel or once the changes authorized by this revised certificate have been applied, will conform to the terms, conditions, and specifications of this revised certificate. The results of this review shall be documented in accordance with 10 CFR 72.212(b)(5) no later than {insert date 6 months after effective date}.
General licensees using the specific CoC amendments that are being revised are required to meet the conditions of the revised CoCs. The NRC added a condition to the revised CoCs that requires the general licensees to implement the revised CoCs within six months and perform written evaluations in accordance with 10 CFR 72.212(b)(5), which establish that the cask will conform to the terms, conditions, and specifications of the revised CoCs. The six-month timeframe in the condition is considered a standard timeframe for implementation, consistent with the information in Regulatory Issue Summary 2017-05, Administration of 10 CFR Part 72 Certificate of Compliance Corrections and Revisions, (NRC, 2017) and the applicant did not request an alternate timeframe.
5.0 TECHNICAL SPECIFICATIONS There were no changes to the certificates technical specifications.
6.0 REFERENCES
(NAC, 2023a)
Letter from Heath Baldner (NAC International) to the U.S. Nuclear Regulatory Commission (NRC), 10 CFR 72.242 Reportable Licensing Basis Non-Mechanistic Tip-over Evaluation Deficiency for the NAC-UMS and MAGNASTOR Dry Cask Storage System, dated March 10, 2023 (ML23069A215).
(NAC, 2023b)
Letter from Heath Baldner (NAC International) to the U.S. Nuclear Regulatory Commission (NRC), Submission of an Amendment Request for the NAC International Universal Storage System, dated October 10, 2023 (ML23283A249).
(NAC, 2023c)
Letter from Heath Baldner (NAC International) to the U.S. Nuclear Regulatory Commission (NRC), Submission of Data Files to Support the Nuclear Regulatory Commission's (NRC) Review of NAC-UMS
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 11 OFFICIAL USE ONLY - PROPRIETARY INFORMATION Amendment No. IO (Submittal 23A), dated October 10, 2023 (ML23291A095).
(NAC, 2024a)
Letter from Heath Baldner (NAC International) to the U.S. Nuclear Regulatory Commission (NRC), Supplement to the Amendment Request for the NAC International Universal Storage System (UMS) Amendment No. 10, dated February 13, 2024 (ML24044A221).
(NAC, 2024b)
Letter from Heath Baldner (NAC International) to the U.S. Nuclear Regulatory Commission (NRC), Submission of Responses to the Nuclear Regulatory Commission Request for Additional Information for NAC International Universal Storage System (UMS) Amendment No. 10, dated June 26, 2024 (ML24179A059).
(NAC, 2025a)
Letter from Heath Baldner (NAC International) to the U.S. Nuclear Regulatory Commission (NRC), Submission of a Supplement to the Nuclear Regulatory Commissions Request for Additional Information for NAC International Universal Storage System (UMS) Amendment No. 10, dated July 17, 2025 (ML25198A286).
(NAC, 2025b)
Letter from Heath Baldner (NAC International) to the U.S. Nuclear Regulatory Commission (NRC), Submission of a Replacement Page for the Supplement to the Nuclear Regulatory Commission's Request for Additional Information for NAC International Universal Storage System (UMS) Amendment No. 10, dated August 27, 2025 (ML25240A911).
(NRC, 2017)
U.S. Nuclear Regulatory Commission, Regulatory Issue Summary 2017-05, Administration of 10 CFR Part 72 Certificate of Compliance Corrections and Revisions, dated September 13, 2017 (ML17165A183).
(NRC, 2019)
U.S. Nuclear Regulatory Commission, NUREG-2214, Managing Aging Processes in Storage (MAPS) Report, dated July 2019 (ML19214A111).
(NRC, 2020)
U.S. Nuclear Regulatory Commission, NUREG-2215, Standard Review Plan for Dry Cask Storage Systems and Facilities - Final Report, Revision 0, dated April 2020 (ML20121A190).
(NRC, 2024a)
Letter from Christopher Markley (NRC) to Heath Baldner (NAC International), Request for Additional Information for the Technical Review of the Application for Amendment Nos. 5 to 9, Revision No. 1, and Amendment No. 10, Revision 0, to Certificate of Compliance No.
1015 (Cost Activity Code System/Enterprise Project Identifier Nos.
001028/L-2023-LLA-0154), dated May 20, 2024 (ML24141A116).
(NRC, 2024b)
U.S. Nuclear Regulatory Commission, Renewal Package for the NAC-UMS System, CoC 1015, dated June 11, 2024 (ML24151A008).
(NRC, 2024c)
U.S. Nuclear Regulatory Commission, List of Approved Spent Fuel Storage Casks: NAC International, Inc. NAC-UMS Universal Storage System Certificate of Compliance No. 1015, Renewal of Initial Certificate
OFFICIAL USE ONLY - PROPRIETARY INFORMATION 12 OFFICIAL USE ONLY - PROPRIETARY INFORMATION and Amendment Nos. 1 to 9, Federal Register, Volume No. 89, Issue No. 118 (June 18, 2024), pages 51400 - 5140.
(Williamson, 1994)
D. A. Williamson, Postirradiation Fuel Assembly Dimensions for Transportation and Storage Cask Dimensions, Radwaste Magazine, January 1994.
CONCLUSION The staff performed a detailed safety evaluation of the application for the Renewed Amendment Nos. 5 through 9, Revision 1, and Renewed Amendment No. 10 to CoC No. 1015 for the NAC-UMS. The staff performed the review in accordance with the guidance in NUREG-2215 (NRC, 2020) and NUREG-2214 (NRC, 2019). Based on the statements, analyses, and representations contained in the application, the response to the RAI, and the technical specifications, and for the reasons described in sections 2 and 3, the staff concludes that these changes do not affect the ability of the NAC-UMS to meet the requirements of 10 CFR Part 72.
Issued with CoC No. 1015, Renewed Amendment No. 5, Revision 1; Renewed Amendment No.
6, Revision 1; Renewed Amendment No. 7, Revision 1; Renewed Amendment No. 8, Revision 1; Renewed Amendment No. 9, Revision 1; and Renewed Amendment No. 10 On.