ML25063A145
| ML25063A145 | |
| Person / Time | |
|---|---|
| Site: | 07109235 |
| Issue date: | 04/03/2025 |
| From: | Yoira Diaz-Sanabria Storage and Transportation Licensing Branch |
| To: | Baldner H NAC International |
| Shared Package | |
| ML25063A144 | List: |
| References | |
| EPID L-2024-LLA-0110 | |
| Download: ML25063A145 (1) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Heath Baldner, Licensing Manager NAC International 2 Sun Court, Suite 220 Norcross, GA 30092
SUBJECT:
REVISION 26 OF CERTIFICATE OF COMPLIANCE NO. 9235 FOR THE MODEL NO. NAC-STC PACKAGE
Dear Heath Baldner:
By application dated July 31, 2024 (Agencywide Documents Access and Management System
[ADAMS] Accession No. ML24214A017), enclosed is Certificate of Compliance No. 9235, Revision No. 26, for the Model No. NAC-STC package. Changes made to the enclosed certificate are indicated by vertical lines in the margin. The staffs safety evaluation report is also enclosed.
The approval constitutes authority to use the package for shipment of radioactive material and for the package to be shipped in accordance with the provisions of Title 10 of the Code of Federal Regulations (10 CFR) Section 71.17, General License: NRC-Approved Package and 49 CFR 173.471.
In accordance with 10 CFR 2.390, Public inspections, exemptions, requests for withholding, a copy of this letter will be available electronically for public inspection in the NRC Public Document Room (PDR) or from the Publicly Available Records component of the NRCs ADAMS. ADAMS is accessible from the NRC website at http://www.nrc.gov/reading-rm/adams.html. The PDR is open by appointment. To make an appointment to visit the PDR, please send an email to PDR.Resource@nrc.gov or call 1-800-397-4209 or 301-415-4737, between 8 a.m. and 4 p.m. eastern time (ET), Monday through Friday, except Federal holidays.
April 3, 2025
H. Baldner If you have any questions regarding this certificate, please contact Martin Ortiz Gonzalez at 301-415-3637.
Sincerely, Yoira Diaz-Sanabria, Chief Storage and Transportation Licensing Branch Division of Fuel Management Office of Nuclear Material Safety and Safeguards Docket No. 71-9235 EPID L-2024-LLA-0110
Enclosures:
1.
CoC No. 9235, Rev. No. 26 2.
Safety Evaluation Report cc w/encls. 1& 2: R. Boyle, U.S. Department of Transportation J. Shenk, U.S. Department of Energy c\\o L. F. Gelder Signed by Diaz-Sanabria, Yoira on 04/03/25
H. Baldner
SUBJECT:
REVISION 26 OF CERTIFICATE OF COMPLIANCE NO. 9235 FOR THE MODEL NO. NAC-STC PACKAGE DATED: April 3, 2025 Closes EPID No. L-2024-LLA-0110 DISTRIBUTION:
DSFM r/f NMSS r/f KJamerson, NMSS ADimitriadis, RI BDesai, RII DHills, RIII GWarnick, RIV 71sf9235all@listmgr.nrc.gov 71sf9235coc@listmgr.nrc.gov ADAMS Accession Package No.: ML25063A144(pkg), ML25063A145(ltr & encl 2),
ML25063A146(encl 1)
OFFICE:
NMSS/DFM NMSS/DFM NMSS/DFM NMSS/DFM NAME:
MOrtizGonzalez SFigueroa ASotomayor JSolis DATE:
3/12/2025 3/12/2025 3/12/2025 3/13/2025 OFFICE:
NMSS/DFM NMSS/DFM NMSS/DFM NMSS/DFM NAME RPatel DDunn CStanko DJohnson DATE 3/14/2025 3/17/2025 3/17/2025 3/21/2025 OFFICE:
NMSS/DFM NMSS/DFM NMSS/DFM NMSS/DFM NAME LRegner GGeorge TGovan YDiaz-Sanabria DATE 3/21/2025 3/24/2025 3/24/2025 4/3/2025 OFFICIAL RECORD COPY SAFETY EVALUATION REPORT Docket No. 71-9235 Model No. NAC-STC Certificate of Compliance No. 71-9235 Revision 26
SUMMARY
By letter dated July 31, 2024 (Agencywide Documents Access and Management System Accession No. ML24214A017), as supplemented on August 22, 2024 (ML24235A515), August 29, 2024 (ML24242A111), September 18, 2024 (ML24263A097), October 23, 2024 (ML24298A155), and February 12, 2025 (ML25043A295), NAC International (NAC) submitted an application to revise Certificate of Compliance (CoC) No. 9235 for the Model No. NAC-STC package. The applicant requested to make changes to the containment boundary leakage testing requirements as well as to the safety analysis report (SAR) section 8.1.5, for the tests associated with the cask fabrication neutron shielding integrity test. The U.S. Nuclear Regulatory Commission (NRC) staff reviewed the application, including its supplement, using the guidance in NUREG-2216, Standard Review Plan for Spent Fuel and Transportation Packages for Radioactive Material. Based on the statements and representations in the application, as supplemented, and the conditions listed in this safety evaluation report (SER),
the staff concludes that the packages meet the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 71, Packaging and Transportation of Radioactive Material.
1.0 GENERAL INFORMATION 1.1 Requested Changes On July 31, 2024, the applicant submitted Revision 24A, as supplemented by submittals 24B, 24C, 24D, 25A, and submittal 423-4001, Revision 0 for the Model No. NAC-STC package to request changes to the containment boundary leakage testing requirements as well as the cask fabrication neutron shielding integrity test. As a part of this proposed revision, in order to confirm the maximum containment boundary leakage, the applicant performed an evaluation of the internal pressure for the Model No. NAC-STC, for both High Burn-Up (HBU) Fuel and Low Burn-Up (LBU) Fuel. In previous revisions of the application the NRC staff required that the NAC-STC transportation package shall be leak-tight for HBU, because of the lack of availability of HBU release rate fraction information. The release of NUREG-2224 Dry Storage and Transportation of High Burnup Spent Nuclear Fuel allows the applicant to perform containment calculations for both LBU and HBU release fractions. The application also includes an administrative editorial change to restore section 5.7.6, figures 5.7.6-1 and 5.7.6-2 to the chapter; they were unintentionally deleted in Revision 21 of the SAR.
1.2 Evaluation Findings
Based on a review of the statements and representations provided by the applicant, the staff concludes that the contents have been adequately described to meet the requirements of 10 CFR Part 71.
1.3 Drawings NAC revised 3 drawings:
Drawing 423-800, Rev. 22P Cask Assembly - NAC-STC Cask Drawing 423-811, Rev. 14 Detail, NAC-STC Cask Drawing 423-900, Rev. 10 Package Assembly Transportation, NAC-STC Cask 2.0 STRUCTURAL EVALUATION The staff reviewed the proposed changes and determined that they did not impact previous SER findings regarding the packages structural design. Therefore, the staff finds that a new structural evaluation is not needed.
3.0 THERMAL EVALUATION The applicant, NAC, applied for the NRC review of a proposed revision to CoC No. 9235, (Revision 26) to request changes to the containment boundary leakage testing requirements as well as the cask fabrication neutron shielding integrity test. As a part of this proposed revision, in order to confirm the maximum containment boundary leakage, the applicant performed evaluation of the internal pressure for the Model No. NAC-STC, for both HBU and LBU. These calculations were performed for both Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) as described in 10 CFR Part 71.
3.1 Staff SAR Review The staff conducted the review using the general guidance provided in NUREG-2216 section 3, "Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material. The staff confirmed that the thermal performance of the Model No. NAC-STC, containing directly loaded HBU or LBU, was adequately evaluated for the tests specified under both NCT and HAC.
The staff reviewed the applicants proposed changes to SAR chapter 3 and determined that the changes made to the SAR thermal section were appropriate, with no changes to performance in the package which would decrease reasonable assurance of adequate protection. The staff also confirmed that the applicant performed the calculations required to determine the internal pressure in the NAC-STC spent fuel transportation system for the requested change in leakage testing requirements. The staff reviewed the temperatures reported by the applicant in this amendment request; the staff also reviewed the materials and calculations shown in NAC calculation 423-4001 and found the results to be bounded by those in the most recently approved version of the applicant SAR.
The staff concluded that none of the proposed changes in the SAR would impact the ability of the NAC-STC design to meet the thermal performance requirements in 10 CFR Part 71.
Based on this conclusion, the staff finds that the NAC-STC package design continues to meet the thermal requirements of 10 CFR Part 71.
3.2 Evaluation Findings
Based on a review of the statements and representations in the application, the NRC staff concludes that the thermal design has been adequately described and evaluated, and that the thermal performance of the package meets the thermal requirements of 10 CFR Part 71.
4.0 CONTAINMENT EVALUATION The objective of the review is to verify that the containment performance of the Model No. NAC-STC transportation package has been adequately evaluated for the tests specified under both NCT and HAC of transport and that the package design satisfies the containment requirements of 10 CFR Part 71.
NAC-STC Revision 26 makes changes to the containment boundary leakage testing requirements. In previous revisions of the application NRC staff required that the NAC-STC transportation package shall be leak-tight for HBU, because of the lack of availability of HBU release rate fraction information. The release of NUREG-2224 Dry Storage and Transportation of High Burnup Spent Nuclear Fuel allows the applicant to perform containment calculations for both LBU and HBU release fractions.
4.2 Description of Containment System The primary containment vessel for the NAC-STC consists of an inner shell, two transition sections, bottom inner forging, and a top forging. The containment vessel components, except for the transition sections, are fabricated from American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Type 304 stainless steel nuclear pressure vessel material. The two transition sections are ASME Boiler and Pressure Vessel Code, Type XM-19 stainless steel nuclear pressure vessel material. The O-rings of the inner lid, the vent port cover plate and the drain port cover plate are the seals that provide primary containment.
4.3 Containment under Normal Conditions of Transport The entire containment vessel, including each penetration is designed to meet the leak-tight criteria (per American National Standards Institute [ANSI] N14.5-1997) when metallic and nonmetallic O-rings seals are used. For LBU or HBU maximum leakage values are specified in the application, but the values dont meet the leak-tight criteria. For normal conditions of transport permissible leakage rate are calculated and presented in table 4.2-3 of the application.
Table 4.2-3 also provides test sensitivity values. The structural and thermal analyses presented in chapters 2 and 3 of the application, demonstrate that the cask remains leak-tight under any of the normal conditions of transport for metallic and nonmetallic O-ring seals, which ensures there will be no release of radioactive material or ingress of water during transportation. For nonmetallic O-ring seals (for both LBU and HBU), the containment analysis provided in the application ensures that leakage will not exceed allowable limits during transportation.
4.4 Containment under Hypothetical Accident Conditions The entire containment vessel, including each penetration is designed to meet the leak-tight criteria (per ANSI N14.5-1997) when metallic and nonmetallic O-rings seals are used.
For hypothetical accident conditions of transport permissible leakage rate are calculated and presented in table 4.3-1 of the application. The structural and thermal analyses presented in chapter 2 and 3 of the application, demonstrate that the cask remains leak-tight under any of the hypothetical conditions of transport for metallic and nonmetallic O-ring seals, which ensures there will be no release of radioactive material or ingress of water during transportation. For non-metallic O-ring seals (for both LBU and HBU), the containment analysis provided in the application ensures that leakage will not exceed allowable limits during transportation. The results of the structural and thermal analyses presented in chapters 2 and 3, respectively, demonstrate that the package will remain leak-tight and, thus, meet the leakage criteria of 10 CFR 71.51 and prevent ingress of water for all the HAC. For nonmetallic O-ring seals (for both LBU and HBU), the containment analysis provided in the application ensures that leakage will not exceed allowable limits during transportation.
4.5 Leakage Rate Tests The application states that the NAC-STC leakage testing is performed in accordance with the requirements of ANSI N14.5-1997, which is acceptable to the staff. The acceptance criterion for the fabrication, periodic, maintenance, and pre-shipment leak testing is specified in the application.
4.6 Evaluation Findings
The staff has reviewed the description and evaluation of the NAC-STC containment system and concludes that: (1) the application identifies established codes and standards for the containment system; (2) the package includes a containment system securely closed by a positive fastening device that cannot be opened unintentionally or by a pressure that may arise within the package; (3) the package is made of materials and construction that assure that there will be no significant chemical, galvanic, or other reaction.
The staff has reviewed the evaluation of the containment system under NCT and concludes that the package is designed, constructed, and prepared for shipment so that under the tests specified in 10 CFR 71.71 (NCT) the package satisfies the containment requirements of 10 CFR 71.43(f) and 10 CFR 71.51(a)(1) for NCT with no dependence on filters or a mechanical cooling system.
The staff has reviewed the evaluation of the containment system under HAC and concludes that the package satisfies the containment requirements of 10 CFR 71.51(a)(2) for HAC, with no dependence on filters or a mechanical cooling system.
Based on review of the statements and representations in the application, the NRC staff concludes that the NAC-STC package has been adequately described and evaluated to demonstrate that it satisfies the containment requirements of 10 CFR Part 71, and that the package meets the containment criteria of ANSI N14.5-1997.
5Property "ANSI code" (as page type) with input value "ANSI N14.5-1997.</br></br>5" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..0 SHIELDING EVALUATION The objective of this shielding evaluation is to verify that the proposed design changes to the NAC-STC, as it pertains to shielding, provides adequate protection to immediate area workers and members of the public against direct radiation that is above the regulatory limits stated in 10 CFR Part 71 for NCT and HAC.
The applicant is requesting a revision to the certificate for the NAC-STC package (CoC No. 71-9235) to use for canister contents, high burn-up 20 assembly fuel contents, and low burnup 26 assembly fuel contents with reduced hydrogen in the neutron shield.
5.1 Package Contents The package is designed to transport canistered PWR and BWR fuel assemblies as well as GTCC waste. Additionally, the NAC-STC can transport up to 20 directly loaded HBU PWR assemblies with varying decay heats with thermal shunts in empty basket cells or 26 directly loaded LBU PWR assemblies with varying decay heats. Directly loaded contents may use a shield ring to provide additional radiation shielding.
5.2 Design Criteria The NAC-STC design change reduces the NS-4-FR hydrogen content from its original licensing basis. The cask configuration that exhibits the highest dose rate is the 20 PWR Zircaloy Clad, High Burn-Up Assembly, and it is considered the bounding configuration for this analysis. This analysis calculated fuel neutron and n gamma dose rates for radial components in each configuration. Radial shields show the maximum effect of reduced hydrogen versus the axial neutron shield configuration.
In the 26 PWR LBU configuration, the dose rates provided in the SAR were calculated in a configuration without a shield ring which houses the hydrogen to determine the impact of reduced hydrogen in the NS-4-FR neutron shield.
The Canister Contents configuration contains additional shielding from the canister itself and a 60-day cooling time. For this configuration, the applicant did not calculate input for Zr alloy material for conservatism.
There is no change to the source term from the previous amendment.
5.3 Model Specification The applicants evaluation used Monte Carlo N-Particle (MCNP) 6.2 code modeled after their previous evaluation using MCBEND, a Monte Carlo uncertainty method for their analysis.
MCNP provided an updated analysis of dose rates of the prior revision. The calculations for this shielding model provided by the applicant reduced the hydrogen content and removed the shield ring for LBU assemblies from its previously approved licensing amendment.
5.4 Shielding Evaluation Neutron and N-gamma dose rates were calculated using the results from an MCNP analysis that included minimum cooling times in the dose rate comparison table in the SAR. In the PWR High burn-up configuration, Zircaloy Clad fuel assembly dose rates were used as its source term. Zircaloy exhibits higher neutron dose rates than stainless steel and lead. Additionally, the geometry in this analysis calculates from a radial direction versus an axial direction with the former being inherently more conservative. Axial neutron shields are thin compared to radial neutron shield thickness; therefore, the radial shields will represent the maximum effect of the change in hydrogen content. This shielding model provided by the applicant in this SAR displays minimal difference from the dose rates of the previously approved package. The dose rates calculated for the 26 PWR LBU assemblies and the Canister Contents are lower than those of the HBU configuration. The staff finds these approaches are conservative and therefore, acceptable.
5.5 Flux to Dose Rate Conversion Factors The SAR uses the ANSI/American Nuclear Society Standard 6.1.1-1977 flux-to-dose rate conversion factors to calculate dose rates, which are acceptable.
5.6 Dose Rates The dose rates provided in the SAR were calculated to determine the impact of reduced hydrogen in the NS-4-FR neutron shield. For directly loaded, HBU, the radial surface maximum dose rate changed from 375.9 mrem/hr. to 379.1 mrem/hr. with reduced hydrogen. Directly loaded LBU dose rates changed from 65.5 mrem/hr. to 65.9 mrem/hr. for the radial surface maximum neutron and n-gamma dose rates with reduced hydrogen. The maximum dose rate for canister items is 52.3 mrem/hr. Each configuration meets the 10 CFR Part 71 dose rate requirements. For normal conditions, the dose rate limits specified in 10 CFR 71.47 for packages are: 1,000 mrem/hr. on the surface of the enclosed package, 200 mrem/hr. on the outer surfaces of the transport vehicle, and 10 mrem/hr. at 2 meters from the vertical planes represented by the outer lateral surfaces of the transport vehicle. The vehicle surface is defined as the personnel barrier on the same plane as the outer radial surface of the impact limiters.
The personnel barrier will attach to the edge of the vehicle between the impact limiters.
For this analysis, the NAC-STC met the regulatory dose criteria under 10 CFR 71.47.
5.7 Confirmatory Calculations The NRC staff performed a confirmatory analysis using MicroShield and are consistent with that of the applicants. For this amendment, the staff finds the dose rate limits on the package in radial configurations satisfactorily meet the requirements of 10 CFR Part 71 for NCT and HAC regulatory limit criteria. The staff also verified the information provided in the applicants SAR is consistent with the applicable regulations.
5.8 Evaluation Findings
Based on the discussion above, the NRC staff concludes that the revised technical specification for this package, as it is described, sufficiently protects any real individual from dose, meet the requirements of 10 CFR Part 71, and the guidance on format and content in NUREG-2216.
6.0 CRITICALITY EVALUATION
The staff reviewed the proposed changes and determined that they did not impact previous SER findings regarding the package criticality design. Therefore, the staff finds that a new evaluation is not needed.
7.0 MATERIALS EVALUATION The staff reviewed the proposed changes and determined that they did not impact previous SER findings regarding the package materials design. Therefore, the staff finds that a new evaluation is not needed.
8.0 OPERATING PROCEDURES EVALUATION The applicant, NAC, applied for the NRC review of a proposed revision to the CoC to request changes to the operating procedure for package loading, and unloading and preparation of empty package for transport. As a part of this proposed revision, in order to confirm the maximum containment boundary leakage, the applicant performed evaluation of the internal pressure for the Model No. NAC-STC, for both HBU and LBU. These calculations were performed for both NCT and HAC as described in 10 CFR Part 71.
8.1 Staff SAR Review The staff conducted the review using the general guidance provided in section 8, Operating Procedures Evaluation, of NUREG-2216, "Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material.
The purpose of the package operating procedure evaluation is to verify that the proposed changes to the operating procedure of the transport package continue to meet the requirements of 10 CFR Part 71.
SAR chapter 7 provides the operational procedures for package loading, unloading, and preparation of the empty Model No. NAC-STC package for transport.
The staff reviewed the applicants proposed changes to SAR chapter 7 to verify that the package will be operated in a manner that is consistent with its design evaluation. The staff determined that that minor changes made to the SAR operating procedures section were appropriate, with no changes to performance in the package which would decrease reasonable assurance of adequate protection.
8.2.
Staff Findings Based on its evaluation, the staff concludes that the operating procedures provide adequate measures and reasonable assurance for safe operation of the NAC-STC cask in accordance with the requirements of 10 CFR Part 71. Further, the CoC is conditioned such that the package must be prepared for shipment and operated in accordance with the Operating Procedures specified in SAR chapter 7, as amended.
9.0 PACKAGE ACCEPTANCE TESTS AND MAINTENANCE PROGRAM 9.1 Staff SAR Review The staff conducted the review using the general guidance provided in section 9, Acceptance Tests and Maintenance Program Evaluation, of NUREG-2216.
The purpose of the package acceptance test evaluation and maintenance program is to verify that the proposed changes to the acceptance tests and maintenance program to be used for the NAC-STC transport package continue to meet the requirements of 10 CFR Part 71.
SAR chapter 8 identifies the acceptance tests and maintenance programs to be conducted on the empty Model No. NAC-STC package and verifies their compliance with the requirements of 10 CFR Part 71.
The staff reviewed the proposed changes to SAR chapter 8 to verify that the package will be tested and maintained in a manner that is consistent with its evaluation for approval. The staff determined that the proposed changes made to the SAR acceptance tests for the gamma and neutron shielding test were sufficient detail to ensure that the changes have no adverse impact on safety.
9.2 Staff Findings Based on the statements and representation in the application, the staff concludes that the revised acceptance tests for the NAC-STC packaging meet the requirements of 10 CFR Part 71, and that they are adequate to assure the package will be constructed in a manner consistent with its evaluation for approval. Further, the CoC is conditioned to specify that the package must be prepared for shipment and operated in accordance with the Acceptance Tests and Maintenance Program in SAR chapter 8, as amended.
CONDITIONS The CoC includes the following condition(s) of approval:
Condition No. 5(a)(3)(i) was revised to update the drawing revision numbers.
Condition No. 15 was revised to extend the validity of Revision 25 of the certificate to March 31,2026.
The references section has been updated to include this request.
Minor editorial corrections were made.
CONCLUSIONS Based on the statements and representations contained in the application, as supplemented, and the conditions listed above, the staff concludes that the design have been adequately described and evaluated, and the Model No. STC package meets the requirements of 10 CFR Part 71.
Issued with CoC No. 9235, Revision 26, on April 3, 2025.