ML19182A060
| ML19182A060 | |
| Person / Time | |
|---|---|
| Site: | Holtec |
| Issue date: | 07/09/2019 |
| From: | Yen-Ju Chen Spent Fuel Licensing Branch |
| To: | Tomlinson J Holtec |
| Chen Y | |
| References | |
| CAC 001208, EPID: L-2019-LLA-0059 | |
| Download: ML19182A060 (1) | |
Text
July 9, 2019 Joyce Tomlinson Adjunct Licensing Manager Holtec International Holtec Technology Campus 1 Holtec Blvd.
Camden, NJ 08104
SUBJECT:
ACCEPTANCE REVIEW OF REQUEST FOR AMENDMENT NO. 15 TO CERTIFICATE OF COMPLIANCE NO. 1014 FOR THE HI-STORM 100 MULTIPURPOSE CANISTER STORAGE SYSTEM (DOCKET NO. 72-1014, CAC NO. 001028, EPID: L-2019-LLA-0059) - REQUEST FOR SUPPLEMENTAL INFORMATION
Dear Ms. Tomlinson:
By letter dated March 20, 2019 [Agencywide Document Access and Management System (ADAMS) Accession No. ML19092A192], Holtec International (Holtec) submitted to the U.S.
Nuclear Regulatory Commission a request to amend the Certificate of Compliance No. 1014 for HI-STORM 100 Multipurpose Canister Storage System.
The staff has performed an acceptance review of your application to determine if the application contains sufficient technical information to begin a detailed technical review. The staff has determined that the application does not provide sufficient technical information to begin a detailed review and that supplemental information is needed. The information needed to continue our review is described in the enclosed request for supplemental information (RSI).
In order to schedule our technical review, responses to the enclosed RSI should be provided within 30 days from the date of this letter. If Holtec is unable to meet this response date, please notify us, at least one week prior to the due date, of your new submittal date and the reasons for the delay. If Holtec is not able to respond within this timeframe or the RSI responses do not provide sufficient information, the application may not be accepted for review.
J. Tomlinson Please reference Docket No. 72-1014, CAC No. 001028, and EPID No. L-2019-LLA-0059 in future correspondence related to this licensing action. If you have any questions, please contact me at 301-415-1018.
Sincerely,
/RA/
Yen-Ju Chen, Sr. Project Manager Spent Fuel Licensing Branch Division of Spent Fuel Management Office of Nuclear Material Safety and Safeguards Docket No.: 72-1014 CAC No.: 001208 EPID: L-2019-LLA-0059
Enclosure:
RSI
J. Tomlinson
SUBJECT:
ACCEPTANCE REVIEW OF REQUEST FOR AMENDMENT NO. 15 TO CERTIFICATE OF COMPLIANCE NO. 1014 FOR THE HI-STORM 100 MULTIPURPOSE CANISTER STORAGE SYSTEM (DOCKET NO. 72-1014, CAC NO. 001028, EPID: L-2019-LLA-0059) - REQUEST FOR SUPPLEMENTAL INFORMATION, DOCUMENT DATE: July 9, 2019 File Location: G:\SFST\HI-STORM 100\Amendment 15\RSI\HI-STORM_A15_RSI_2019-0627.docx ADAMS No.: ML19182A060 *concur via email OFFICE: DSFM DSFM DSFM DSFM DSFM NAME: YChen WWheatley* JSolis* JIreland* VWilson*
DATE: 6/26/2019 6/27/2019 6/20/2019 6/25/2019 5/24/2019 OFFICE: DSFM DSFM DSFM DSFM YDiaz-NAME: ABarto* TTate* JMcKirgan Sanabria*
DATE: 5/7/2019 6/26/2019 6/6/2019 7/9/2019 OFFICIAL RECORD COPY
Request for Supplemental Information Docket No. 72-1014 Holtec International HI-STORM 100 Multipurpose Canister Storage System Certificate of Compliance No. 1014 Amendment No. 15 By letter dated March 20, 2019 [Agencywide Document Access and Management System (ADAMS) Accession No. ML19092A192], Holtec International (Holtec) submitted to the U.S.
Nuclear Regulatory Commission (NRC) a request to amend the Certificate of Compliance (CoC)
No. 1014 for HI-STORM 100 Multipurpose Canister Storage System.
The staff has performed an acceptance review of the application and determined that the application does not provide sufficient technical information to begin a detailed review. The information needed to continue staffs review is provided in the request for supplemental information (RSI) below.
Thermal RSI 4-1 Provide detailed examples of sizing calculations for the Dry Ice Jacket and the HI-DRIP auxiliary cooling systems.
Section 1.2.1.7 of the final safety analysis report (FSAR) states that the dimensions of the chiller and the amount of dry ice will depend on the rate of heat extraction required and the duration of the short-term operation and therefore, must be custom sized for each application.
Regarding the HI-DRIP auxiliary cooling system, Section 4.II.5.3 of the FSAR states that sizing of HI-DRIP for a typical scenario is archived in the Holtec Calculation Package (Holtec International Thermal Calculation Package HI-2043317, Latest Revision). The staff reviewed Holtec Calculation Package HI-2043317 and did not find any information related to sizing calculations of this auxiliary cooling system.
Detailed sizing calculations should be provided for design basis conditions. The staff needs detailed sizing calculations of these two cooling systems to determine adequate cooling is provided for the period of time that the systems are used and thus verify no thermal limits are exceeded.
This information is necessary to verify the requirements of 10 CFR 72.236(f).
Confinement RSI 5-1 Provide the following confinement information and update the application as described in the responses to the HI-STORM 100 Amendment No. 13 requests for supplemental information (RSI) that were provided as part of the HI-STORM 100 Amendment No. 15:
Enclosure
Verify that the statement, Condition D and SR 3.1.1.3 are not applicable to casks that were loaded to Amendment 2 through 7, does not appear, or is removed from page 3.1.1-1 of HI-STORM 100 proposed Technical Specification Appendices A and C. This is requested because Condition D and SR 3.1.1.1 are applicable to Amendment Nos. 2 through 7, and the HI-STORM 100S Version E Cask is not applicable to Amendment Nos. 2 through 7.
Also remove the statement, MPCs that were loaded under CoC Amendment No. 7 and prior amendments are subject to the requirements of those amendments, which may differ, from page 2-179 of the application, because the statement should not be applicable to vent and drain port cover plate welds, or proposed change No. 10.
This information is needed to determine compliance with 10 CFR 72.236(d), and (f), and 72.244.
Shielding RSI 6-1 Update the information that justifies the source term(s) assumed within the shielding evaluation are appropriate to represent the equation that defines burnup and cooling time for the MPC-32M.
Burnup and cooling time specifications for the MPC-32M contain an approach that has been presented to the NRC under review of Amendment No. 4 of the HI-STORM FW (Docket No. 72-1032). This is a new approach that uses an enrichment that bounds 99% of the fuel population from a database and an equation that calculates cooling time as a function of burnup. The staff discussed this approach with Holtec during a teleconference on March 7, 2019 (Conversation Record: Discuss Holtecs proposed approach in response to NRCs second round of request for additional information (RAI) for HI-STORM Flood/Wind Amendment No. 4, March 7, 2019, ADAMS Accession No. ML19072A166), and at a public meeting on March 19, 2019 (Memorandum from Y.
Chen to C. Reagan, Summary of March 19, 2019 Meeting with Holtec International to Discuss Proposed Response to the Second Round of Request for Additional Information for Certificate of Compliance No. 1032 for HI-STORM Flood/Wind, Amendment No. 4, April 4, 2019, ADAMS Accession No. ML19093A048). During these meetings with Holtec, the staff shared its request for information to justify the methodology. Holtec stated that it would provide the information that the staff requested and has updated the HI-STORM FW Amendments No. 4 and 5 applications with additional information; however, the application for HI-STORM 100 Amendment No. 15 does not include the requested information. The staff requests that the applicant update the information related to the burnup and cooling time equation method and provide the following information:
- a. The applicant needs to provide the reference for the database used to generate Figure 5.II.2-2 from the safety analysis report (SAR) and the minimum enrichment Table 5.II.2.4 from the SAR. The applicant should justify that the database is appropriate and it is using the most recent available data.
- b. The applicant needs to provide a discussion of how it determined that the enrichments in Table 5.II.2.4 are bounding for 99% of discharged PWR spent nuclear fuel.
- c. For higher burnup assemblies (> 55 GWd/MTU), there is less data available for establishing an enrichment value that bounds 99% of the discharged fuel population. These are also the assemblies whose source term (and ultimately dose/dose rate) is most sensitive to enrichment (see Figures 12 and 13 of NUREG/CR-6716, Recommendations on Fuel Parameters for Standard Technical Specifications for Spent Fuel Storage Casks). The applicant needs to justify that the selected enrichments for these higher burnup assemblies are bounding for 99% of the discharged PWR spent nuclear fuel population.
- d. The applicant needs to provide dose and dose rate results for the cooling time/burnup points along the curves specified in Table 2.1-4 of CoC Appendix D (also in Table 2.II.1.6 of the SAR) to demonstrate that it found the bounding cooling time/burnup combination along the curve for use in these evaluations.
- e. For the regionalized loading patterns that contain a unique cooling time and burnup correlation for each region corresponding to each decay heat, the applicant needs to provide a detailed discussion of the procedure it used to determine the maximum dose and dose rates.
This information is needed for the staff to evaluate the appropriateness of the technical specification dose rate limits, as well as to be able to evaluate if the shielding features ensure that the DSS design and operations facilitate licensee compliance with the requirements prescribed in 10 CFR Part 20, Subpart C, Occupational Dose Limits, and 10 CFR 72.126(a) and the capability of the cask system to meet the requirements of 10 CFR 72.104 and 72.106, including the dose limits, as required by 10 CFR 72.236(d).
6-2 Provide the minimum lead and water jacket thickness for the HI-TRAC MS, and provide normal, off-normal, and accident condition dose evaluations for the HI-TRAC MS with these minimum thicknesses.
Based on the discussion in Section 1.II.2.3 of the SAR and Drawing No. 11381, the HI-TRAC MS has variable lead and water jacket thicknesses. Supplement 5.II of the SAR does not include information on the lead thickness or water jacket thickness used to perform the dose rate evaluations, however it does include statements (such as the following) that lead the staff to conclude that it did not use the minimum thickness to evaluate the dose/dose rates:
Section 5.II.4 of the SAR states: the shielding analysis of Version MS under the accident condition is not necessary for the reference Version MS and the expected results are bounded by the analysis of reference HI-TRAC 100. Nonetheless, the additional site-specific shielding evaluations shall be performed to confirm the shielding performance of Version MS, if the lead thickness of the customized Version MS cask is less than the lead thickness of the reference 100-ton HI-TRAC, analyzed in the main body of Chapter 5.
Although the staff may have accepted nominal dimensions used in Chapter 5 for previous amendments, the nominal in those cases are with respect to a tolerance on the order of a few millimeters rather than a variable lead thickness. The difference between minimum and nominal lead for the HI-TRAC MS with a variable lead shield could be significant. In Section 1.II.2.3 of the SAR, the applicant compares the HI-TRAC MS to the HI-TRAC VW (approved for use in the HI-STORM FW system, Docket No. 72-1032) by stating that the HI-TRAC MS is a smaller diameter counterpart of HI-TRAC VW Version V in HI-STORM FW FSAR. The staff notes that the difference in nominal versus minimum lead in the HI-TRAC VW is greater than 1 inch. The amount of lead needed to attenuate 1 MeV gammas by half is roughly 9 mm (0.35 inches). Thus, a reduction in lead of more than 1 inch would have a significant effect on the gamma dose rates.
Although site specific evaluations are needed and used to demonstrate compliance with 10 CFR 72.104 (per 10 CFR 72.212(b)(5)) at the general licensees site, 10 CFR 72.236(d) requires that the applicant for a CoC ensure that radiation shielding features of the spent fuel storage cask system are sufficient for compliance with 10 CFR 72.104 and 72.106.
This information is needed for the staff to be able to review compliance with 10 CFR 72.236(d).
6-3 Provide an evaluation demonstrating that the MPC-32M with stainless steel clad fuel assemblies meets the regulatory dose requirements.
CoC Appendix D Table 2.1-1 Section V.A.2.a allows the storage of stainless steel clad assemblies within the MPC-32M; however, there is no corresponding analysis of stainless steel clad assemblies within Chapter 5.II of the SAR demonstrating that storage of stainless steel clad assemblies in the MPC-32M meets regulatory dose limits.
Due to the activation of the Co-59 impurity within stainless steel cladding, assemblies with stainless steel clad can have a significant increase in dose/dose rate over that of zircaloy clad fuel. Section 3.4.2.6 of NUREG/CR-6716, Recommendations on Fuel Parameters for Standard Technical Specifications for Spent Fuel Storage Casks, states:
The steel clad fuel potentially increases the cask dose rate by more than an order of magnitude over that from conventional Zircaloy clad fuel.
This information is needed for the staff to be able to review compliance with 10 CFR 72.236(d).
Criticality RSI 7-1 Revise the application to provide Reference 6.A.19, HI-2033039, CRITICAL EXPERIMENT BENCHMARK, Revision 5.
Appendix 6.A of the SAR, Benchmark Calculations, refers to Reference 6.A.19, HI-2033039, CRITICAL EXPERIMENT BENCHMARK, Revision 5, for detailed discussion of the MCNP5-1.51 benchmarking analysis. This document supports the benchmarking
analysis for the previously approved HI-STORM 100 MPCs. This document has been updated to include experiments relevant to the partial gadolinium credit requested in this amendment for BWR fuel in the MPC-68M canister. This information is needed for the staff to confirm that the MCNP5-1.51 code used for the partial gadolinium credit calculations is properly validated for this purpose.
This information is needed to ensure compliance with the criticality safety requirements in 10 CFR 72.124 and 72.236(c).
Acceptance Criteria and Maintenance Program RSI 10.1 Based on sizing calculations, include appropriate thermal tests that demonstrate the HI-DRIP and Dry Ice Jacket auxiliary cooling systems will function as designed.
Section 9.II of the FSAR does not include appropriate acceptance thermal test that demonstrate the performance of these systems. The test should demonstrate systems design criteria are met. Thermal performance tests are needed to determine the adequacy of these auxiliary systems to provide the necessary cooling to prevent exceeding any thermal limits.
This information is necessary to verify the requirements of 10 CFR 72.162, 72.234(a),
and 72.236(f).
Radiation Protection RSI 11-1 Update SAR Chapter 10 to include the HI-TRAC MS.
Chapter 10: Radiation Protection, from the SAR was not updated to account for the operations of the HI-TRAC MS. Throughout this chapter there are many considerations that warrant inclusion of the HI-TRAC MS especially considering variable shielding options for this transfer cask.
Section 10.II of the SAR states: additional auxiliary/temporary shielding, such as listed in Table 10.1.1, may be used to further reduce the dose rates around the HI-TRAC when performing short term operations. Table 10.1.2 of the SAR states requirements for using auxiliary and temporary shields for the other HI-TRAC designs but needs to be updated to include requirements for the HI-TRAC MS. The dose evaluations need to be performed without additional auxiliary/temporary shielding. If additional auxiliary/temporary shielding is assumed in the dose assessments, these should be listed as required.
Table 10.3.1b and Table 10.4.1 of the SAR need to be updated to consider HI-TRAC MS when using minimum lead and water jacket thicknesses.
This information is needed for the staff to evaluate the capability of the cask system to facilitate the control and limiting of occupational exposures consistent with the requirements in 10 CFR Part 20 and 10 CFR 72.126(a), including maintaining exposures
ALARA, and to evaluate the capability of the cask system to meet the requirements in 10 CFR 72.104 and 106, including the dose limits, as required by 10 CFR 72.236(d).
OBSERVATIONS O-6-1 Tables 5.II.4.1 and 5.II.4.4 are used to compare dose rate performance between the HI-STORM 100 overpacks and the HI-TRAC transfer casks, respectively. The applicant needs to justify why the source terms selected to perform these comparisons are representative of the variety of spent fuel contents that may be loaded and stored in the transfer casks and overpacks or provide additional comparisons with source terms that have different gamma/neutron contributions. The shielding performance for the different transfer casks and overpacks may change depending on the source terms contribution from gammas and neutrons, i.e., whether the source is neutron or gamma dominated. In other words, the shielding performance may be similar for a gamma dominated source term if both cask features are especially good at shielding gammas, however, for a neutron dominated source, depending on the cask features, the shielding performance may not be similar. In addition, the applicant needs to consider any features that would affect the positioning of the contents and consequently the shielding performance of the Version E overpack and the HI-TRAC MS as compared to the other overpacks and transfer casks.
O-6-2 The shielding performance for damaged fuel adopts the conclusions from Section 5.4.2.2 of the SAR. Additional justification is needed to show how these conclusions remain applicable for the HI-STORM 100S Version E and the HI-TRAC MS. The damaged fuel analysis for PWR fuel is based on the MPC-24 with 4 damaged fuel locations while the MPC-32M is allowed for 16 damaged fuel locations within CoC Appendix D, Table 2.1-1.V.B. The 16 locations where damaged fuel are allowed are all periphery locations and would cause a uniform increase in dose rates around the periphery versus localized increase of the 4 damaged assemblies within the MPC-24.
Also, due to its variable shield thicknesses, the HI-TRAC MS can have less shielding than the HI-TRAC 100, which is what the damaged fuel analysis in Section 5.4.2.2 of the SAR is based on. In addition, NUREG/CR-7203, A Quantitative Impact Assessment of Hypothetical Spent Fuel Reconfiguration in Spent Fuel Storage Casks and Transportation Packages, shows a significant increase in dose rates near storage cask air vents due to reconfiguration. This report also shows reconfiguration can have a significant impact on controlled area boundary doses.
O-10-1 In Supplement 9.II, Acceptance Criteria and Maintenance Program, of the application, the word, toto, should be revised to clarify the meaning of the sentence.