ML24101A213
| ML24101A213 | |
| Person / Time | |
|---|---|
| Site: | 07109401 |
| Issue date: | 04/23/2024 |
| From: | Roman C Division of Fuel Management |
| To: | Corcoran S Holtec |
| References | |
| EPID L-2023-NEW-0005 | |
| Download: ML24101A213 (1) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Sean Corcoran, Licensing Engineer Holtec International 1 Holtec Blvd Camden, NJ 08104
SUBJECT:
NON-ACCEPTANCE OF APPLICATION FOR CERTIFICATE OF COMPLIANCE NO. 9401 FOR THE MODEL NO. HI-STAR 370 PACKAGE (ENTERPRISE PROJECT IDENTIFIER L-2023-NEW-0005)
Dear Sean Corcoran:
By letter dated January 30, 2024 (Agencywide Documents Access and Management System Package Accession No. ML24030A975), Holtec International (Holtec) submitted an application for approval of the Model No. HI-STAR 370 transport package. On June 13, 2023, and July 18, 2023, we conducted pre-application meetings during which we expressed initial concerns on your proprietary technical approach for this submittal. Based on these discussions, we considered the significant technical issues raised by the staff would take time to address. Going forward we would encourage continued engagements prior to a submittal.
This letter is to inform you that the U.S. Nuclear Regulatory Commission (NRC) staff is not accepting your application for a detailed review because Holtec has failed to provide appropriate and adequate information about the package design and its analysis. The NRC staff identified major deficiencies in your application and significant missing technical information. As such, the NRC staff determined that it could not proceed with an adequate safety review for reasonable assurance of adequate protection.
- 1. The application does not provide a comprehensive description and design details (e.g.,
thicknesses, diameter, center of gravity location) of each reactor vessel to be transported along with its internals and other contents including inactivation cover and blind flanges, which is essential for the development of models used in the safety analyses. The application failed to explain how defueled and irradiated reactor vessels, as initial contents, along with irradiated reactor vessel assemblies (non-fuel),
also referred to cask contents, as other contents, and water, some of which is immobilized with absorbent material impact the static and dynamic analyses of the package in order to evaluate the validity of the package response under normal conditions of transport (NCT) and hypothetical accidental conditions (HAC) drop tests.
The review of the presented analysis of transport scenarios under NCT and HAC cannot be undertaken without the missing information on the contents and its modelling in the drop simulations provided.
- 2. The applications statement that LS-DYNA has been benchmarked through the quarter-scale model 9-meter drop experiments in support of HI-STAR 100 Part 71 certification and the 1/4 scale model 9-meter drop test in support of HI-STAR ATB 1T does not constitute a validation of the LS-DYNA model for the HI-STAR 370 April 23, 2024 S. Corcoran 2
package. The applicant has not provided a comparison of the similarities between the HI-STAR 370 and the scale models before any test drop response can be compared with the simulated drop results. Without such a validation, the results of the different scenario simulations cannot be considered as acceptable for the demonstration that the regulatory requirements are satisfied.
- 3. The LS-DYNA impact analysis material model uses true stress values derived from engineering stress strain curves and the applicant states that the material strains are to be derived from material testing. However, the applicant did not provide (i) any information on the source of the data used in the material models, (ii) a validation that the material model, as used in the simulation, can indeed simulate the entire range of the physical testing conducted to gather the material data. Thus, the staff cannot evaluate the LS-DYNA model simulation capabilities for the HI-STAR 370 drop scenarios.
- 4. The application of the American Society of Mechanical Engineers (ASME) BPVC section VIII, Division 2 elastic-plastic analyses has not been justified by the applicant, nor has the applicant advised which specific clauses of the ASME code are being applied to the analysis. The applicant has not provided a thorough explanation of the process by which results are extracted from the model and has not provided a justification as to why the triaxiality factor used to determine the limiting strain value is set to a constant value.
- 5. The application fails to demonstrate that HAC loading events are being applied sequentially and in a manner to produce maximum damage. For example, the Puncture test is performed at mid-height of the package, when it has already been established in the Free Drop test that a breach of the package corner is likely. Finally, the Fire test is performed for an unbreached package, when the Free Drop test has already established that a breach of the package corner is likely.
- 6. The application failed to include a damage model that is used in the LS-DYNA simulation. The damage model allows the simulation to process incremental damages which may lead to the erosion of some elements and allow the loss of confinement to occur. It is essential for the incorporation of such a simulation model with some erosion criteria associated with the limiting strain for the staff to evaluate the extent of loss of confinement in a specific drop scenario. Lacking this information, the staff is unable to evaluate compliance with the release limits in a loss of confinement.
The staff has also identified deficiencies and missing technical information, thus preventing an efficient and effective review of your application. In particular, the applicant did not justify statements such as The majority of the residual water is expected to be immobilized by the absorbent material based on using twice the amount of absorbent necessary to react with the expected residual water (footnote on Page 4.3-1) or While a breach in the containment boundary is possible during HAC, the loaded reactor vessel is expected to remain largely intact due to the presence of internal steel reinforcing structural components and grout, robust thicknesses of the reactor vessel shell and inactivation cover, and a tight circumferential fit-up with the top flange" (the application page 4.3-2).
The staff noted that (i) the material type, properties (e.g., physical, chemical, thermal, radiolysis-related), behavior, and effectiveness of the absorbent and getter were not described, (ii) the S. Corcoran 3
water/absorbent is not fixed, and (iii) it appears that some internal components may also be loose inside the reactor vessel (i.e., are not immobilized internally with grout). In addition, the application did not clearly explain if the stated loose contamination activity encompasses the inner vessel non-fixed contamination (including water, absorbent, internal CRUD) and the vessel outer surface non-fixed contamination.
The applicant also did not provide specific acceptance criteria for grout curing time prior to lid installation/package transport, did not explain how any free water and excess water in the grout is quantified and accounted for in the calculation of hydrogen concentration, and did not address if there was a time requirement between lid closure and transport completion to ensure internal pressures (e.g., radiolysis gases and water vapor pressure effects) and hydrogen concentration from the excess water were below acceptance criteria.
The basis for the expectation of reactor vessel integrity was not provided and staff cannot determine whether the integrity of the reactor vessel content is maintained or not. Staff did not find any justification for the choices of drop test orientations and subsequent puncture test locations being bounding, and staff also noted that the application figure 5.3.5 shows a potential effect of the puncture test but does not address the behavior of the grout after the drop test.
The applicant did not explain why a 1-meter puncture test at specific potential locations, e.g.,
thin portions of reactor vessel and welded plates in conjunction with small grout thicknesses, was not investigated as part of the HAC drop tests. In addition, maximum temperatures of the reactor vessel and cavity interior for the fire HAC were not provided and there were no radiolysis and pressure calculations to determine hydrogen gas concentration and internal pressure (e.g.,
water vapor pressure effects) within the reactor vessel during NCT and HAC. Note that staff guidance is to not rely on getters for radiolysis mitigation.
You may resubmit the application after addressing the deficiencies identified above. The NRC staff is available to discuss the methods for assuring that an application is adequate for the NRC review and includes all analyses and references, in accordance with 10 CFR 71.7, that are necessary to support the regulatory findings the staff must make. In that regard, please review the Division of Fuel Management Division Instruction LIC-FM-2, Acceptance Review Process (ML22161B042) for further guidance.
Please note that the items listed above, many of which were discussed during earlier pre-application communications, should not be considered all-inclusive and only reflect that the acceptance of your application for a detailed technical review would have impacted our ability to maintain an effective and timely licensing review schedule due to the time it would take for Holtec to respond to a request for supplemental information, in conjunction with the potential for multiple requests for additional information that would be needed for the NRC staff to adequately conduct its safety evaluation; hence, this would not be consistent with the NRCs Principles of Good Regulation.
S. Corcoran 4 If you have any questions regarding this matter, please contact Pierre Saverot of my staff.
Please reference Docket No. 71-9401 and Enterprise Project Identifier L-2023-NEW-0005 in future correspondence related to this matter.
Sincerely, Cinthya Román, Deputy Director Division of Fuel Management Office of Nuclear Material Safety and Safeguards Docket No. 71-9401 EPID L-2023-NEW-0005 cc: 71bw9401all@listmgr.nrc.gov Signed by Roman, Cinthya on 04/23/24
ML24101A213 OFFICE NMSS/DFM NMSS/DFM NMSS/DFM NMSS/DFM NAME PSaverot JBorowsky DMarcano JSmith DATE 4/10/2024 4/11/2024 4/11/2024 4/12/2024 OFFICE NMSS/DFM NMSS/DFM NMSS/DFM NMSS/DFM NAME TBoyce SFigueroa YDiaz-Sanabria CRomán DATE 4/12/2024 4/12/2024 4/15/2024 4/23/2024