ML23046A110

Enclosure 1 - Castor geo69 Transport Package - Non-Proprietary

ML23046A110
Person / Time
Site:	07109383
Issue date:	03/16/2023
From:	Storage and Transportation Licensing Branch
To:	Gesellschaft fur Nuklear-Service mbH
Shared Package
ML23046A109	List: ML23046A110 ML23046A112
References
EPID L-2021-NEW-0002, CoC No. 9383
Download: ML23046A110 (18)

Text

Non-Proprietary Request for Additional Information Docket No. 71-9383 Certificate of Compliance No. 9383 Model No. CASTOR geo69

By letter dated July 10, 2020 (Agencywide Documents Access and Management System

[ADAMS]Accession No. (ML20198M431), Gesellschaft für Nuklear-Service mbH (GNS) requested approval of its quality assurance program. By letter dated January 14, 2021 (ML21033A353), as supplemented on July 2, 2021 (ML21196A374), August 25, 2021 (ML21245A234) and March 2, 2022 (ML22073A009), GNS submitted an application for a certificate of compliance for the Model No. CASTOR geo69 spent fuel transportation package.

This request for additional information identifies information needed by the U.S. Nuclear Regulatory Commission (NRC) staff in connection with its review of the application. The requested information is listed by chapter number and title in the applicants safety analysis report (SAR). The NRC staff used NUREG-2216, Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive MaterialFinal Report, in its review of the application. Each question describes information needed by the staff for it to complete its review of the application and to determine whether the applicant has demonstrated compliance with regulatory requirements.

1.0 General information

1-1 Justify categorizing the aluminum shells of the impact limiters, and their respective welds as not important to safety on the parts list for Drawing No. 1014-DPL-38772.

The NRCs guidance in NUREG/CR-6407, Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety, section 5.5.1, Category A Items, states The shell is a metal skin that covers the energy-absorbing material and protects that material from minor damage and weather.

In some cases, the shell may provide structural support to the impact limiter assembly.

For the steel shells and their welds to be not important to safety (i.e., not a part of the package relied on to meet the requirements in Title 10 of the Code of Federal Regulations (10 CFR) Part 71, the shells would not have to provide structural support to the foam for the impact limiters. In addition, the steel would not be relied on to ensure that the impact limiters remained attached to the package body during the tests for normal conditions of transport and hypothetical accident conditions.

This information is needed to demonstrate compliance with 10 CFR 71.47, 10 CFR 71.51(a), 10 CFR71.55(d) and 10 CFR 71.55(e), 10 CFR 71.71, and 10 CFR 71.73.

2.0 Structural Evaluation

2-1 Explain the methodology used for the development of the deceleration time-history (T-H) including selection of the node for recording the T-H. Using T-H plots show how much of the drop energy converted to internal energy and what was dissipated in other forms during contact. In addition, provide the displacement T-H of the point on the impact limiter closest to the rigid surface at impact along with the development of the contact area with respect to time.

Enclosure 1 This information is needed by the staff to evaluate whether the simulation model of the package reasonably replicates the response of an actual drop under the conditions required by 10 CFR 71.71 and 71.73.

This information is needed to demonstrate compliance with 10 CFR 71.71 and 10 CFR 71.73.

2-2 Provide a discussion of the anticipated non-linear behavior of the package and the different elements of the LS-DYNA used to capture the non-linearities in the evaluation for the 30-foot drop analysis under hypothetical accident conditions.

The applicant used LS-DYNA non-linear elements in its numerical simulation in lieu of an actual test to evaluate the 30 foot drop test for hypothetical accident conditions. To evaluate whether the model has the capability to adequately simulate the impact, the staff requires a discussion of the anticipated non-linear behavior of the package and the different elements of the LS-DYNA used to capture these non-linearities. This information is required to assure the staff that the model simulation can capture the non-linear behavior that would occur in a drop test as required by 10 CFR 71.73.

This information is needed to demonstrate compliance with 10 CFR 71.73(c)(1).

2-3 Either provide an English version of reference 5 in section 2.1, Safety Standards of the Nuclear Safety Standards Commission (KTA) 3201.2 Components of the Reactor Coolant Pressure Boundary of Light Water Reactors Part 2: Design and Analysis 2017-11 or revise the application to use an equivalent U.S. code or standard.

The document is required to determination whether the preload on bolts and the length of engagement is sufficient.

This information is required to determine compliance with 10 CFR 71.31(c),

10 CFR 71.71, 10 CFR 71.73, and 10 CFR 71.45.

2.1 Material Evaluation

2-4 Provide additional information on the cast iron package coatings that demonstrate coating durability and support the emissivity values credited in the thermal analysis.

SAR table 3.2-7, Applied emissivity of the relevant components, provides emissivity values for the coated inner and outer cask body surfaces. The outer cask surface is described as having an epoxy paint, while the inner surface is described as having a Thermaline coating. Neither of these descriptors is considered to provide a sufficient basis for the emissivity values, as they describe broad varieties of coatings. Provide the manufacturer data sheets or other bases to support the emissivity values in the thermal analysis.

In addition, for the Thermaline coating on the inner package surface, provide materials qualification data, manufacturer data sheets, or other bases that demonstrate the performance of the coating under the elevated heat and radiation exposures of the internal package environment.

This information is needed to demonstrate compliance with 10 CFR 71.33(a)(5) and 10 CFR 71.43(d).

2-5 Justify the maximum allowable temperatures for the containment gaskets during hypothetical accident conditions.

SAR table 3.2-9, Temperature limits of components, provides the maximum allowable temperatures for the metallic gaskets during hypothetical accident conditions. These limits exceed the manufacturers operating temperature limit, as described in SAR appendix 2-9, Material Qualification, Metal Gaskets. Provide the basis for allowing the gaskets to exceed the manufacturers operating limit.

This information is needed to demonstrate compliance with 10 CFR 71.41(a), and 10 CFR 71.51(a)(2).

2-6 Justify that the {proprietary information removed} fuel basket plates will have adequate fracture performance in a drop accident to maintain the assumed fuel configuration in the criticality analyses.

The SAR does not include toughness testing requirements to verify that brittle fracture will not affect the structural integrity of the basket in a drop accident. The staff notes that the criticality analysis relies on the maintenance of configuration of the neutron absorber plates and fuel assemblies.

Although nonferrous materials are generally excluded from fracture acceptance testing in consensus standards (e.g., the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel (B&PV) Code, Division 3, WD-2300), the proprietary

{proprietary information removed} metal matrix composite is a non-code material that contains boron carbide ceramic particles that may diminish fracture performance relative to the conventional aluminum materials that are considered in the ASME B&PV Code.

Therefore, justify that the current acceptance testing is capable of verifying adequate impact performance or otherwise propose additional test methods and acceptance criteria, see item 8-1, below.

This information is needed to demonstrate compliance with 10 CFR 71.55(e) and 10 CFR 71.59(a)(2).

2-7 Provide the basis for the conclusion that irradiation will not affect the shielding performance of the polyethylene moderator material.

SAR section 2.2.3, Effects of Radiation on Materials, states that the irradiation of the ultrahigh molecular weight polyethylene shielding is approximately {proprietary information removed} (per SAR appendix 5-3, table 1) and concludes that this level of radiation is insignificant with respect to the loss of shielding performance.

The staff notes that the stated radiation level has been found to change the properties of polyethylene moderators. For example, gamma radiation levels above 2x105 Gy were considered to lead to oxidation, crosslinking and hydrogen release of ultrahigh molecular weight polyethylene [von der Ehe, 2015]. Also, a study to evaluate the use of high density polyethylene in nuclear waste storage found that radiation exposures up to 100 Mrad (1x106 Gy) would be expected to lead to embrittlement [Dougherty, 1983].

Given the observations of the above (and similar) studies, provide the basis for concluding that changes in polyethylene under irradiation will have no effect on shielding performance.

This information is needed to demonstrate compliance with 10 CFR 71.43(d).

References

von der Ehe, K; Kmmling, A.; and Wolff D., Neutron Radiation Shielding Material Polyethylene: Consequences of Gamma Irradiation, WM2015 Conference, March 15-19, 2015, Phoenix, Arizona.

Dougherty, D. and Adams, J., Radiation Resistance Testing of High-Density Polyethylene, Brookhaven National Laboratory, Report No. BNL-NUREG-33641, January 1983.

2-8 Clarify the definition of allowable undamaged fuel contents with respect to the function of the cladding in the safety analyses.

SAR section 1.2.2, Contents, states that only undamaged fuel assemblies may be loaded, and SAR section 0.5, Glossary, defines undamaged fuel as fuel that can meet all fuel-specific and system-related functions.

It is unclear what fuel-specific and system-related functions the fuel must meet, as the fuel cladding does not appear to be credited with maintaining the configuration of the fuel pellets. As stated in the July 2, 2021, response to supplemental information 2-6

[Enclosure 2 to letter T1213-CO-00010], in the safety analyses the applicant assumes hypothetical reconfiguration of the high burnup spent fuel contents into justified geometric forms. If the fuel cladding does not need to maintain the configuration of the fuel pellets, then any degree of cladding damage would seem to be allowed.

Clarify the criteria that general licensees are to use to classify fuel as undamaged.

This information is needed to demonstrate compliance with 10 CFR 71.33(b)(3), 10 CFR 71.55(d)(2), 10 CFR 71.55(e)(1) and 10 CFR 71.59(a)(2).

2-9 Justify the package and canister dryness criteria with respect to the potential for corrosion of the package internals and contents and the potential to generate flammable gas (hydrogen).

SAR section 7.1.2 states that the package and canister cavities are vacuum dried to a pressure of approximately {proprietary information removed} Provide the basis for that threshold, including how much residual water is considered to remain within the cavities and the potential for that residual water to lead to corrosion and hydrogen generation.

This information is needed to demonstrate compliance with 10 CFR 71.43(d).

2-10 Provide additional information to demonstrate that the package drying criteria are adequate to prevent an unacceptable loss of cladding toughness due to hydrogen reorientation.

As described in question 2-8 above, it is unclear to the staff if the fuel cladding is credited with performing a fuel-specific or system-related function. If so, the following is requested:

SAR section 3.6, Thermal Evaluation for Short-term Operations, includes a maximum allowable cladding temperature during short-term operations (e.g., drying) of 400 °C, citing the guidance in NRC Interim Staff Guidance (ISG)-11, Revision 3 Cladding Considerations for the Transportation and Storage of Spent Fuel. The staff notes that this limit was recommended, in part, to inhibit the formation of radial hydrides in the cladding, which may reduce the claddings ductility and degrade its performance in a drop accident.

However, ISG-11 (which has since been incorporated NUREG-2216, section 7.4.14.2) includes a second drying recommendation to prevent hydride reorientation: limiting the thermal cycling of the cladding to fewer than 10 cycles, where the cladding temperature variations during each cycle do not exceed 65°C (117°F). The CASTOR geo69 SAR does not include this criterion in the short-term operation limits.

Provide the basis for allowing fuel drying operations without limits on thermal cycling, justifying that the Zircaloy-2 cladding will not undergo an unacceptable reduction in ductility due to hydride reorientation.

This information is needed to demonstrate compliance with 10 CFR 71.55(d)(2).

2-11 Resolve the following incorrect or missing information in the SAR:

a. Several parts of the SAR state that the impact limiter housing is constructed of steel and other portions state it is constructed of aluminum:

SAR section 1.1 states that the impact limiters basically consist of a steel casing filled with polyurethane foam. SAR section 5.1.1 states that both impact limiters include steel as housing and inner lying structure sheets. Section 1.2.1.6, Impact limiters states The lid impact limiter (Item 90) and the bottom impact limiter (Item 95) consist of an aluminum housing with a PU-foam.

b. In SAR table 2.1-2, the Sm and Su stress value limits for SA-965 Grade FXM-19 at 120°C appear to be incorrect.

The stress value limits appear to be consistent with the 100°C data from the ASME B&PV Code, rather than with values that would be expected for 120°C.

c. Footnote 7 in SAR table 8.2-1 is missing

This information is needed to demonstrate compliance with10 CFR 71.31 and 10 CFR 71.33.

3.0 Thermal Evaluation

3-1 Describe and justify the modeling of gaps in the CASTOR geo69 package (packaging and content) thermal models for normal conditions of transport, hypothetical accident

conditions, and short-term operations, recognizing that gaps can have a large impact on temperatures and thermal performance.

SAR section 3.3.1.7 mentions the gap between the canister and the package body and performs a gap sensitivity analysis. However, the entire system is composed of components with corresponding tolerances, including the basket, moderator rods and plates, and spacing between solid aluminum shielding elements with the basket and canister shell, all of which have corresponding gaps associated with fabrication. For example, the SAR and basket (drawing no. 1014-DD-30984) appear to indicate that the basket is not welded but is fabricated from stacked fitted parts. The lack of continuous welded connections between the baskets steel and aluminum alloy components (e.g.,

structure sheets, outer sheets, shielding sheets) indicates the presence of gaps that cause thermal resistance in conducting heat transfer from the fuels decay heat axially and radially outward through the basket.

A detailed discussion, including a sensitivity analysis, should be provided that justifies the choice of gap sizes and how the gaps are modeled in a bounding manner, recognizing there is uncertainty in tolerances and changes in fitting, including due to distortion during steady-state and transient operations (e.g., storage, vacuum drying, transfer, fire hypothetical accident conditions). In addition, discussion should include:

a. details of whether the above mentioned basket design (which does not have continuous welded connections) has been satisfactorily used in other CASTOR spent fuel systems and

b. details of whether the thermal performance of that basket and packages similar to the CASTOR geo69 package design have been experimentally validated as part of CASTORs experimental validation program mentioned in SAR section 3.3.1.1.

This information is needed to determine compliance with 10 CFR 71.35.

3-2 Discuss the ability of the fins to resist deformation during short-term operations and loading onto the transport vehicle and their ability to retain effectiveness over time due to buildup of dirt and debris.

The CASTOR geo69s performance is based on a finned transport package design.

However, there was no discussion of the robustness of the fins to resist deformation and damage during short-term operations, loading, and normal conditions of transport. In addition, it is not clear from the SAR whether the fins are a load-bearing component (i.e.,

either due to the tiedown system, the bearing surfaces of the transport frame, the locking mechanism or the lashing belts) when the package is in the transfer frame. In addition, there was no sensitivity analysis of thermal performance due to damaged fins and impacts of dirt or debris buildup between fins (e.g., thermal resistance, change in emissivity and absorptivity) and no discussion whether there is a need for periodic maintenance to remove dirt and debris.

This information is needed to determine compliance with 10 CFR 71.35.

3-3 Perform a surface energy balance of the packages outer surface to calculate the package surface temperature during the hypothetical accident conditions fire condition

so that a review of the thermal models boundary conditions and thermal inputs (e.g.,

decay heat flux, fire input) can be performed.

SAR Figure 3.4-7 shows a package surface temperature much less than the fire temperature even though the surface experiences the thermal convection input and thermal radiation heat transfer input from the 800 °C fire. Note that the effective surface emissivity equation described in SAR section 3.4.2.1 should consider the engulfing nature of the fire (e.g., top and bottom package surfaces). The derivation of a surface energy balance will aid in understanding the interaction of the thermal inputs and boundary conditions.

This information is needed to determine compliance with 10 CFR 71.35.

3-4 Clarify and update the various parameters used in the pressure calculations and provide pressure information associated with low burnup fuel, which has a higher fission gas release fraction.

a. The fission gas release fraction for normal conditions of transport used in the pressure calculations (e.g., SAR tables 4.2-1, 4.2-2) used a 15 percent value for high burnup fuel. However, there was no indication that pressures based on high burnup fuel would bound a low burnup fuel, which is assumed to have a 30 percent fission gas release fraction.

b. The canister and package pressure at hypothetical accident conditions impact reported in SAR table 4.3-3 and table 4.3-4 was based on a sum of gas released from the content into the canister cavity of {proprietary information removed}.

However, the sum of the mobilized fission gas in the canister and mobilized fuel rod filling gas helium is {proprietary information removed} (according to the values in table 4.3-3). In addition, the larger number of moles would also affect the sum of gas released from the content into the canister and cask cavity reported in SAR table 4.3-4 and the resulting package pressure, which may require updated containment analyses.

This information is needed to determine compliance with 10 CFR 71.35.

3-5 Clarify in the SAR whether there are any time limits associated with transferring the spent fuel content between the vertical orientation (e.g., transfer cask) to the CASTOR geo69 transportation package (horizontal orientation).

Section 3.6 of the CASTOR geo69 transportation SAR refers to the section 4.7 in the CASTOR geo69 storage SAR for thermal performance and time limits associated with short-term operations. There was no discussion associated with operations associated with the transfer between horizontal orientations (e.g., transfer package) to the CASTOR geo69 transportation package. The operations discussed in section 4.7 dealt with vertical canister orientations, but there was no discussion concerning horizontal orientations, which would tend to have different thermal correlations and boundary conditions. Likewise, there was no analysis or discussion relating to the content and package temperatures when the transfer cask is positioned on top of the CASTOR geo69 package with the bottom lid closed, as described in SAR table 7.1.2 (e.g., steps G1.5 and G1.6).

This information is needed to determine compliance with 10 CFR 71.35.

3-6 Provide additional discussion of the UHMW-PE moderators allowable temperature and the effect of the material being at temperatures greater than its allowable temperature.

SAR table 3.2-9 does not provide the maximum allowable temperature of the

{proprietary information removed} moderator material. SAR table 3.4-2 indicates that temperatures can reach 310°C during the hypothetical accident conditions fire. Although SAR appendix 4-2 indicated that hydrogen generation from radiolysis would result in a small amount of hydrogen (less than {proprietary information removed} per year), the SAR should clarify that the UHMW-PE material does not thermally degrade during the fire such that combustible gases may form and react.

This information is needed to determine compliance with 10 CFR 71.43(d).

3-7 Provide in the SAR the minimum and maximum allowable temperatures of the VMQ (vinyl-methyl-silicon rubber) and FKM (fluorocarbon rubber) elastomeric seals during normal conditions of transport.

SAR section 2 mentioned that VMQ and FKM seals are used during package operations.

The seal allowable temperatures are needed to confirm that the seals would be operable during and after normal conditions of transport, including at cold conditions.

This information is needed to determine compliance with 10 CFR 71.35 and 10 CFR 71.51(a)(1).

3-8 Provide additional discussion and justification for the water convection heat transfer parameter described in the Part 72 CASTOR geo69 SAR section 4.7 and the water flow rate mentioned in SAR table 7.1-1.

a. Section 3.6 of the CASTOR geo69 transportation SAR refers to Section 4.7 in the CASTOR geo69 storage SAR with regards to short-term operations. Section 4.7 of the CASTOR geo69 storage SAR indicated a {proprietary information removed}

water convection heat transfer coefficient for short-term operations within the pool.

However, the description in Section 4.7 appears to indicate relatively small temperature differences between components in the water pool. Buoyant heat transfer correlations between parallel walls with small temperature differences would indicate convection heat transfer coefficients less than the assumed

{proprietary information removed} value in the SAR. The sensitivity of temperatures (and resulting relevant time constraints) for different heat transfer coefficient values was not considered in the SARs analysis.

b. It appears from SAR Table 7.1-1 (step D2.6) that flushing water in the annulus between the canister and transfer cask is necessary for cooling purposes, but a flow rate and supporting calculation were not provided.

This information is needed to determine compliance with 10 CFR 71.35.

3-9 Clarify that the CASTOR geo69 package thermal model design and generation, analysis, and review falls under the GNS CASTOR Part 71 quality assurance program.

a. Although SAR section 3.3.1.1 mentioned that a particular ANSYS code is verified and validated (e.g., SAR appendix 2-5 to document number 1014-SR-00001 described a verification of the ANSYS finite element program for mechanical calculations), there was no discussion that the thermal modeling design and review process is part of the GNS CASTOR Part 71 quality assurance program. In addition, Staff also notes that although SAR chapter 3 indicates that ANSYS Release 17.2 is used for thermal calculations, SAR appendix 2-5 to document number 1014-SR-00001, which described the verification of the ANSYS finite element program for mechanical calculations, addressed ANSYS 15.0.7 and ANSYS Version 2019 R1; therefore, clarify that the ANSYS version used in the thermal model (Release 17.2) is part of the quality assurance program.

b. SAR section 3.3.1.1 stated that the calculation methods for the thermal analyses of several other GNS transport and storage package designs have been validated, including comparison of calculations with experimental results. The validations and experimental result comparisons of the thermal ANSYS models of GNS transport packages should be provided, including discussion regarding their relevance with the GNS CASTOR geo69 transportation package.

This information is needed to determine compliance with 10 CFR 71.35.

3-10 Provide results of thermal energy balances (e.g., numerical residuals), spatial grid generation sensitivity results for steady-state runs, and time step sensitivity results of transient runs for the CASTOR geo69 package ANSYS thermal model.

Although SAR chapter 3 provided normal conditions of transport and hypothetical accident conditions results of the three-dimensional half-symmetric CASTOR geo69 package thermal analyses, there was no discussion that confirmed grid and timestep parameters were appropriate. In addition, there was no discussion that the thermal analyses were appropriately converged. These numerical parameters help to understand the relevance of the numerical results described in the SAR.

This information is needed to determine compliance with 10 CFR 71.35.

3-11 Clarify and discuss how the increased area of the fins was considered when imposing insolation boundary conditions during normal conditions of transport and for the increased radiation heat transfer during the 30 minute fire hypothetical accident conditions.

SAR section 3.3.1.3 indicates that the packages radial fins were not explicitly modeled; rather, a surface enhancement factor was applied to the heat transfer coefficient at the models corresponding unfinned surface. However, there was no discussion regarding the increased thermal input to the fins additional area from insolation during normal conditions of transport and radiation heat transfer from an 800 °C fire.

This information is needed to determine compliance with 10 CFR 71.35.

3-12 Perform a thermal analysis that assumes cladding failure at normal conditions of transport in order to address the assumption of hypothetical reconfiguration of high burnup spent fuel described in the response to the request for supplemental information RSI 2-6 (ML21245A234).

The response to RSI 2-6 indicated a design approach of including normal conditions of transport and hypothetical accident conditions analyses that assumed hypothetical fuel reconfiguration. Although SAR section 3.5.4 performed a hypothetical accident conditions thermal analysis assuming redistribution of fuel rods, a normal conditions of transport thermal analysis and its results that considered redistribution of fuel rods were not described in the SAR.

This information is needed to determine compliance with 10 CFR 71.35.

3-13 Provide additional information regarding the 2D detailed model whose temperature results were used to calculate the fuel assembly effective thermal conductivity.

SAR section 3.3 described in broad terms the process of determining the effective thermal conductivity. Additional model information to compare with other models and make a finding include the following: emissivity and absorptivity of the cladding, fuel rod, water rod, fuel channel, the full 2D model wall (basket sheets); basket cell width; dimensions of the components and distances between the listed components in SAR Figure 3.3-4(b).

This information is needed to determine compliance with 10 CFR 71.35.

3-14 Provide complete thermal analysis discussion of the short-term and operation thermal analyses mentioned in SAR section 3.6.

The SAR thermal chapter provides thermal analyses of normal conditions of transport and hypothetical accident conditions. However, the two pages of SAR section 3.6 (Thermal Evaluation for Short Term Operations) only briefly refer to thermal analyses of short-term and operations provided in the docket for the storage cask review (Docket No. 72-1053) which is also under review. However, a complete, standalone application is needed to complete the review. The thermal analyses provided should consider the varied operations discussed in SAR chapter 7, Package Operations, that could affect thermal performance because of differences in content, package, and transfer positions, canister and cask environments, times for operations to achieve safe conditions, etc.

This information is needed to determine compliance with 10 CFR 71.31, 10 CFR 71.33, and 10 CFR 71.35.

3-15Provide further details about surface conditions and treatments to achieve the emissivity values listed in SAR Table 3.2.7.

SAR table 3.2.7 listed the emissivity values for different important to safety components.

Further information should be provided that describes the manner that these components rely on surface treatments and the measures in place that ensure an improved thermal performance can be maintained over time.

This information is needed to determine compliance with the regulatory requirements in 10 CFR 71.43(d), 10 CFR 71.51.

4.0 Containment Evaluation

4-1 Demonstrate that the CASTOR geo69 transportation system meets the leaktight criteria, as described in American National Standards Institute (ANSI) N14.5-2014, American National Standard for Radioactive MaterialsLeakage Tests on Packages for Shipment by providing for a leakage test of the packaging/canister bodies during fabrication.

Section 4.1.1 of the SAR states: that the monolithic cask body and the lids can be considered as leaktight, so the containment analysis can be reduced to the gasket sealing system. Further, in section 4.2.3, the application states that the design leakage rate of the considered containment is not greater than 10-7 ref-cm3/s (leaktight according to ANSI N14.5) This application does not clearly indicate that leakage testing of the entire package/canister bodies will be done; however, in order to meet the leaktight criteria in ANSI N14.5, the entire containment boundary must undergo fabrication leak testing and meet the leak rate acceptance criteria specified in ANSI N14.5, as there is no recognized standard that allows for the assumption of monolithic materials to be leaktight without being leak tested.

This information is needed to determine compliance with 10 CFR 71.51.

4-2 Provide the calculation used to determine the leakage rate and potential release from the CASTOR Geo69 transportation package, for both normal and accident conditions.

While the NRC does not require leak rate calculations for a package that is tested to leaktight in accordance with ANSI N14.5, the SAR includes the summary of an analysis to calculate the allowable release rates; however, the actual calculations were not included.

In the SAR, the containment analysis presented is described as having been completed using the method described in ANSI N14.5. The actual calculations have not been provided as part of the application and the staff needs to review the release calculations in order to confirm that they were done in accordance with ANSI N14.5. Additional guidance on leakage rate testing may be found in section 4.4.2.2 of NUREG 2216.

This information is required to determine compliance with 10 CFR § 71.51.

4-3 Provide the torque specification for the lid bolts of the CASTOR geo69 package lid.

Include this specification in either the operating procedures in SAR or the drawings for the package. Further, provide a more detailed description of {proprietary information removed} (SAR Page 1.2-5) and explain how it is used in the tightening of the lid bolts on the CASTOR geo69 package.

The staff was not able to locate a torque specification for the lid bolts of the CASTOR geo69 package in either the SAR or any of the accompanying drawings. The term

{proprietary information removed} is not one that is familiar to the staff in context of its use in section 1.2.1.2 of the SAR. Further clarification of this term is requested.

This information is required to determine compliance with 10 CFR 71.51.

4-4 Provide a table in the application to summarize all leakage rate tests performed on the CASTOR geo69 package and clarify (in the table) whether ANSI N14.5 or another standard, is used in the leakage rate tests.

a. The applicant described the leakage rate tests in SAR Sections 4.4, 7.1.2 and 8.2.2.

The applicant should provide a table in SAR chapter 7 or chapter 8 to summarize the fabrication, pre-shipment, maintenance, and periodic leakage rate tests.

Information provided in the table should include, but not be limited to, leakage test criteria, test sensitivity, test methods, test frequency, and containment components that are subject to testing for each of the fabrication, pre-shipment, maintenance, and periodic leakage rate tests.

b. Clarify whether ANSI N14.5 (a consensus standard), or another standard, is used in the leakage rate tests. If ANSI N14.5 is used, clarify whether the leakage rate testing procedures are approved by personnel whose qualifications and certifications in the non-destructive method of leak testing include certification by a nationally recognized society at a level appropriate to the writing and/or review of leakage rate testing procedures (e.g., an American Society of Non-destructive Testing (ASNT) Level III in leak testing) as noted in section 8.8, Quality Assurance, of ANSI N14.5, or equivalent. An ASNT Level III in leak testing can be of great value in the design of a high reliability, economical leak testing program that includes selection of methods, equipment, and generation of procedures.

This information is needed to determine compliance with 10 CFR 71.51.

5.0 Shielding Evaluation

5-1 Clarify the SAR to consistently identify the materials used in the casing of impact limiters. See items 1-1 and 2-11, above.

The applicant states in the SAR that The impact limiters protect the casks during transportation and is made of steel casing filled with polyurethane foam. The drawings show an aluminum casing.

The staff needs this information to determine compliance with the requirements of 10 CFR 71.47, and 10 CFR 71.51.

5-2 Provide justification that the 3 percent fuel failure rate in normal conditions of transportation is adequate for high burnup spent fuel or spent fuel that has been in storage for longer than 20 years.

The SAR state: CASTOR geo69 will transport high burnup spent fuel, therefore, the impact of the 3-percent fuel failure under NCT [normal conditions of transport] is evaluated according to: [1] the source occurring due to fuel failure is relocated to the bottom and to the top regions of the canister; the regions with potentially lower shielding performance due to flattenings in the trunnion regions; and due to finite axial size of the moderator rods.

The 3% fuel reconfiguration assumes that the fuel is loaded as intact fuel. In response to staff RSIs (ML21245A234), the applicant stated that they are not crediting the cladding for high burnup fuel, therefore no credit is taken for cladding as the primary barrier to

radioactive materials releases. If no credit taken for cladding in shielding, therefore, the shielding analyses for high burnup fuel must consider all possible fuel relocation, packing factor, vertical position, horizontal position, redistribution of source terms under transportation of the package. Justify that the 3 percent fuel failure during normal conditions of transport is acceptable since the CASTOR geo69 does not take credit for cladding.

The staff needs this information to determine if the CASTOR geo69 system design meets the regulatory requirements of 10 CFR 71.47.[1] NUREG-2224, Dry Storage and Transportation of High Burnup Spent Fuel, Washington, DC, November 2020.

5-3 Provide explanation how positions are identified in the loading schematic of the package.

It states in SAR Figure 1.2-8 shows the general loading schematic of the cask viewed from the top, including basket position numbers 1 to 69 and position groups A to F.

Position groups are defined by positions which are treated equivalently in terms of shielding and decay heat limit.

Explain what is meant by are treated equivalently in terms of shielding and decay heat limit, at the bottom of page 1.2-22.

The staff needs this information to determine compliance with the requirements of 10 CFR 71.47, and 10 CFR 71.51.

5-4 Explain clearly how the tolerance evaluated for nominal thickness of the package.

In the SAR, section 5.1.1, stated that; The thicknesses of the materials relevant for the shielding analysis are set to their minimum. For example, the nominal thickness of the cask is {proprietary information removed}, the shielding model assumes {proprietary information removed}.

Cleary explain how the tolerances are considered and why the tolerance can be

{proprietary information removed} and why this approach acceptable to use.

The staff needs this information to determine compliance with the requirements of 10 CFR 71.47, and 10 CFR 71.51.

5-5 Clarify nominal or reduced dimension used in evaluation the dose rates.

It states in SAR that: The dose rates on the face sides of the package are much smaller than on the shell side, which is decisive for the design, therefore, {proprietary information removed}.

This statement seems to contradict with other parts of the SAR where dimensions are decreased in the shielding model (see above section 5.1.1 which states that dimensions are reduced in the modeling).

The staff needs this information to determine compliance with the requirements of 10 CFR 71.47, and 10 CFR 71.51.

5-6 Clearly explain the other conservative assumptions (see underline below) that were made. In addition, provide a more thorough explanation of how length growth affects the shielding performance.

It states in SAR: Possible negative impact of the bigger lid side central opening

{proprietary information removed} on the dose rates of the lid side of the cask should be more than compensated by other conservative assumptions made. Firstly, the plating of the polyurethane foam is not considered in the shielding model at all; its presence would provide a better shielding. And secondly, the length growth would not only provide more material but also shift the normally occupied space further from the cask.

It is not clear what the possible negative impact on shielding is referring to, and what are the other conservative assumptions were made to justify compensate the dose rates, and what is the length growth and how it effects the dose rates. The staff needs this information to determine compliance with the requirements of 10 CFR 71.47, and 10 CFR 71.51.

6.0 Criticality Evaluation

6-1 Clarify the distribution of the empty basket positions in the partial loading evaluations.

The applicant evaluated a series of single, empty basket locations as well as different numbers of multiple empty locations. In section 6.3.6.1 of the application, the applicant describes the single empty locations as radially distributed and shows the location identifiers in Figure 6.3-10 of the application. The applicant then states the results are given in table 6.3-5 and Figure 6.3-11; however, these numbers do not correspond to the basket location identifiers shown in the figure described in the previous sentence. It seems these numbers relate to a separate evaluation of multiple empty positions, but it is not clear from the applicants description. There are also no description or figures showing the distribution of multiple empty basket positions.

This information is required to determine compliance with 10 CFR 71.55(b).

7.0 Operating Procedures

7-1 Provide general guidance in the operating procedures for controlling the radiation level limits on unloading operations and procedures for addressing situations when surface contamination and radiation surveys are too high.

Steps H1 and H2 insection 7.1.3 of the operating procedures requires verification of the contamination levels and dose rates prior to shipment. However, there are no procedures associated with controlling radiation level limits and addressing situations when contamination and radiation levels exceed regulatory limits. Section 8.4.2.1, Receipt of Package from Carrier, in NUREG-2216, Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material, states that the staff should ensure that the package operating procedures include actions to be taken if contamination or radiation levels exceed those listed in 10 CFR Part 71. The NRC notes that it occasionally receives notifications from its licensees that, after package receipt at its destination, measurements of contamination or radiation levels have exceeded the requirements in 10 CFR 71.87(i) or 10 CFR 71.47, respectively, even though

contamination and dose rate measurements prior to shipment did not exceed the requirements.

Regulatory Guide 7.7, Administrative Guide for Verifying Compliance with Packaging Requirements for Shipping and Receiving of Radioactive Material section 4.0, Receiving and Opening a Package, contains an approach that the staff considers acceptable for meeting the requirements associated with receipt of radioactive material in 10 CFR Part 71 and 10 CFR Part 20. In addition, NUREG/CR-4775, Guide for Preparing Operating Procedures for Shipping Packages, may contain useful information.

This information is needed for a user to be able to comply with 10 CFR 71.89 and 10 CFR 20.1906.

7-2 Either revise the operating procedures to include the package pre-shipment leakage rate tests and criterion in tables 7.1-2 and 7.1-3 Operations for preparation for transport, or the NRC will make the criterion for the leak rate tests a condition of the certificate of compliance.

In section 4.1.1 of the SAR, the applicant indicates that for both the inner and outer containment, a maximum reference air leakage rate of 1x10-7 ref cm3/s (leak test criterion) is demonstrated by a leakage test after loading (see SAR chapter 7, table 7.1-2) and that this maximum reference air leakage rate is proven for the outer containment prior to transport; however, a pre-shipment leakage test is not listed as a step in the procedures listed in table 7.1-3, Operations for preparation for transport. In addition, the acceptance criterion for the test in section 7.1.2 is not provided.

This information is needed to determine compliance with 10 CFR 71.51.

7-3 Either ensure that the tilting studs are rendered inoperable for lifting and tying down the package or evaluate the titling studs for compliance with the requirements for lifting and tie-downs.

While the tilting studs are not designed to be a lifting attachment or for tying down the package, they are a structural component of the package that could be used to lift or tie down the package. It is not clear from the application whether the tilting studs are accessible for use as tie-downs or lifting during transport or removed from the package.

This information is needed to determine compliance with 10 CFR 71.45(a).

7-4 Revise section 7.2.2.2, Removal of Contents via CLU, Step M in table 7.2-3 which provides the operational steps describe for lowering the canister into the pool water and reflooding of the canister cavity with pool water to provide criteria to control the temperature differential between the fuel rod and the pool water to ensure that fuel damage does not occur. Ensure that the application contains sufficient justification for the criteria.

Canister reflooding prior to removing the spent fuel should ensure that the temperature differential would not exceed that which would cause significant change in the thermal response of the fuel cladding to cause degradation and cladding failure prior to unloading the fuel to ensure that doses to a licensee unloading the package meets the

dose rates and the as low as is reasonably achievable (ALARA) principal in 10 CFR Part 20, Standards for Protection Against Radiation. Note that this should include, as appropriate, material degradation due to aging and irradiation.

The information is needed to determine compliance with 10 CFR 20.1101 and 10 CFR 20.1201, 10 CFR 71.55 and 10 CFR 71.89.

8.0 Acceptance Tests and Maintenance Program

8-1 Provide additional information on the qualification testing or acceptance testing of the

{proprietary information removed} fuel basket plates that demonstrates adequate uniformity of the boron carbide distribution to support subcriticality.

{proprietary information removed}

The guidance in NUREG-2216, section 7A.1.5, states that methods other than neutron attenuation testing (such as chemical analysis) can be used for acceptance testing, provided that the alternative method has been previously qualified and benchmarked with direct attenuation measurements. In that guidance, applicants are encouraged to provide statistically significant data showing the correspondence between neutron attenuation testing and the alternative testing.

The staff requests the following information to demonstrate how the {proprietary information removed} can be used to adequately characterize the boron uniformity in the acceptance testing of the manufactured plates:

1. Provide available data that benchmarks the {proprietary information removed} with neutron attenuation measurements.

2. Provide the acceptance criteria (and a justified basis) for the {proprietary information removed} for boron uniformity.

This information is needed to demonstrate compliance with 10 CFR 71.33(a)(5), and 71.55(b), (d) and (e).

8-2 Provide a frequency for replacement of the elastomeric and metallic seals for the CASTOR geo69 package lid.

SAR section 8.2.2 indicates that elastomeric sealing rings and metallic gaskets are inspected and may be replaced in the course of leakage testing of the package. There does not appear to be a specific frequency or schedule provided for replacement of these seals. Section 9.4.2.2, Leakage tests, of NUREG-2216 provides NRC guidance on seal replacement frequency.

This information is required to determine compliance with 10 CFR 71.51.

9.0 Quality Assurance

The following questions are on the quality assurance program submitted on July 10, 2020 (ML20198M431).

9-1 Revise section 10.4.3-1 to provide additional information on the design measures or specific method captured in your Quality Assurance Manual (QAM) II to select and review for suitability of materials, parts, and equipment that are important to safety such as standard or commercial (off-the-shelf) material, parts, and equipment.

GNS-Quality Assurance Program Description (QAPD) chapter 3.3 states, in part, that the design methods, as well as applicable materials, parts, computer codes, equipment and processes that are essential to the functions of the components and systems being designed are specified in the QAM II. Provide a brief description of the design measures specified in the QAM II to select and review for suitability of application of materials, parts, equipment, and process that are important to safety.

This information is needed to demonstrate compliance with 10 CFR 71.107.

9-2 Revise section 10.4.7-1 to provide additional information on how GNS plans to implement the control of commercial-grade items and services.

GNS-QAPD Chapter 7.7 states, in part, that when commercial-grade items and services need to be procured, the control of such purchased items and services shall be performed as described in the QAM II. Added the be to the sentence for editorial and verbiage consideration. Provide a brief description on how GNS plans to implement the control of commercial-grade items and services.

This information is needed to demonstrate compliance with 10 CFR 71.115.

9-3 Revise section 10.4.12-1 to provide additional information on the GNS process for using subcontractors for calibration services.

GNS-QAPD Chapter 12.3 states, in part, that calibration may be subcontracted to approved suppliers on the Approved Suppliers List. Does GNS plan to use accredited laboratories in lieu of performing a vendor evaluation (survey) to satisfy calibration testing requirements? The NRC staff provided a few restrictions when using laboratory accreditation by Accreditation Bodies that are signatories to the International Laboratory Accreditation Cooperation (ILAC) Mutual Recognition Arrangement (MRA) in lieu of performing commercial-grade surveys as part of the commercial-grade dedication process for procurement of calibration and testing services.

This information is needed to demonstrate compliance with 10 CFR 71.115 and 10 CFR 71.125.

9-4 Revise section 10.4.17-1 to provide additional information to describe how GNS comply with the record retention requirements.

GNS-QAPD Chapter 17.5 states, in part, that all records shall be maintained for their retention periods according to the QAM II. Provide a brief description of the record retention requirements (i.e., duration, location, fire protection, and assigned responsibilities) as provided in your QAM II.

This information is needed to demonstrate compliance with 10 CFR 71.135.

9-5 Revise section 10.4.17-2 to provide additional information on how GNS minimize the risk of loss, damage, or destruction to quality records or how GNS provide provisions for electronic storage.

This information is needed to demonstrate compliance with 10 CFR 71.135.

9-6 Revise section 10.4.18-1 to provide additional information to describe measures to analyze and trend audit deficiency data as well as ensure that the resulting reports, indicating quality trends and the effectiveness of the quality assurance program, are given to management for review, assessment, corrective action, and follow up.

GNS-QAPD Chapter 18.6 states, in part, that the individual responsible for the organization or activity audited shall investigate audit findings, establish corrective actions, including measures to prevent recurrence, and notify the Lead Auditor of Corrective Action taken or planned. The Lead Auditor shall evaluate the proposed measures in respect of effectiveness, suitability, and implementation of corrective action.

Please clarify what is meant by effectiveness, suitability, and implementation to equate to measures for analyzing and trending audit results and management reviews of audit findings.

This information is needed to demonstrate compliance with 10 CFR 71.137.