ML24242A056
| ML24242A056 | |
| Person / Time | |
|---|---|
| Site: | 07201025 |
| Issue date: | 02/10/2025 |
| From: | Storage and Transportation Licensing Branch |
| To: | |
| Shared Package | |
| ML24242A043 | List: |
| References | |
| CAC 001028, EPID L-2023-LLA-0149, EPID L-2023-LLA-0148 | |
| Download: ML24242A056 (9) | |
Text
PRELIMINARY SAFETY EVALUATION REPORT DOCKET NO. 72-1025 NAC INTERNATIONAL NAC MULTI-PURPOSE CANISTER SYSTEM RENEWED CERTIFICATE OF COMPLIANCE NO. 1025 AMENDMENT NO. 9, AND REVISION TO AMENDMENT NOS. 6, 7, AND 8
SUMMARY
This preliminary safety evaluation report (SER) documents the U.S. Nuclear Regulatory Commission (NRC) staffs review and evaluation of NAC Internationals (NAC, the applicant) request to amend renewed Certificate of Compliance (CoC) No. 1025, and to revise renewed CoC No. 1025, Amendment Nos. 6, 7, and 8, for the NAC Multi-Purpose Canister (NAC-MPC)
System.
By letter dated September 7, 2023 (Agencywide Documents Access and Management System
[ADAMS] Accession No. ML23250A056), as supplemented on February 14, 2024 (ML24040A027), May 14, 2024 (ML24135A322), July 10, 2024 (ML24193A110), and August 27, 2024 (ML24240A133), NAC submitted an application to revise the description of the vertical concrete cask (VCC, concrete cask, cask) in CoC No. 1025 and its technical specifications (TS) for the NAC-MPC System to make a distinction between the VCC body and the VCC lid, in terms of applicability of the American Concrete Institute (ACI) Specifications ACI 349, Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary, and ACI 318, Building Code Requirements for Structural Concrete and Commentary. NAC proposed a new definition for the VCC lid so that the VCC body and the VCC lid can be treated as separate components. This is to allow for the specification of construction requirements for the VCC lid that are separate from the construction requirements for the VCC body.
NAC requested that these changes be made in a new Amendment No. 9 to the CoC, and that these changes also be included in the existing Amendment Nos. 6, 7, and 8 through a revision to these amendments. The amended CoC, when codified through rulemaking, will be denoted as Renewed Amendment No. 9 to CoC No. 1025. The revised CoCs, when codified through rulemaking, will be denoted as Renewed Amendment No. 6, Revision 1; Renewed Amendment No. 7, Revision 1; and Renewed Amendment No. 8, Revision 1 to CoC No. 1025.
The NRC staff's evaluation is based on a review of NACs application, as supplemented, and whether it meets the applicable requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste, for dry storage of spent nuclear fuel. The NRC staff reviewed the amendment and revision request using guidance in NUREG-2215, Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities, dated April 2020 (ML20121A190). The NRC staff also used the guidance in NUREG-2214, Managing Aging Processes In Storage (MAPS) Report, dated July 2019 (ML19214A111), as the NAC-MPC System is a renewed storage system.
2 The NRC staff determined that not all disciplines described in NUREG-2215 are affected by the requested amendment and revision. This amendment and revision review only requires evaluations on Structural, Materials, and Shielding, as described in this SER.
GENERAL DESCRIPTION The objective of this section is to review the changes requested to renewed CoC No. 1025 for the NAC-MPC System to ensure that the applicant provided an adequate description of the pertinent features of the storage system and the changes requested in the application. The specific changes requested by the applicant are described and evaluated in the following sections of this SER.
Proposed Changes The applicant proposed to revise the NAC-MPC system description in the CoC to distinguish between the VCC body and VCC lid to clearly indicate that only the VCC body is a reinforced concrete structure. The applicant also proposes to revise the definitions in the CoC appendix A, Technical Specifications, section A1.1, by revising the definition of VCC body (identified as VERTICAL CONCRETE CASK (CONCRETE CASK) in the technical specification definitions) and adding a new definition for VCC lid (identified as VERTICAL CONCRETE CASK LID in the proposed technical specification definitions).
The applicant also proposes to revise the CoC appendix B, Approved Contents and Design Features, section 3.3, Codes and Standards, to indicate that ACI 349 and ACI 318 govern the design and construction only of the VCC body. NAC also proposes new fabrication requirements for the concrete in the VCC lid, that are different from the construction requirements for the VCC body, to include:
the minimum concrete density, the allowable methods for measuring the concrete density, and requirements for the concrete mix, placement, and curing methods used in the construction of the concrete cask lid.
The proposed TS changes are to ensure that the concrete in the VCC lid has the physical properties (e.g., density, dimensions, geometry, internal structure) needed to perform its radiation shielding safety function as intended, while eliminating the requirement that the concrete in the VCC lid be in compliance with ACI 349 and ACI 318. These ACI standards will remain in the TS for the design and construction of the VCC body.
During its review of the application, the NRC staff identified conflicting information in the NAC-MPC final safety analysis report (FSAR) regarding minimum concrete density (ML23360A160).
The applicant revised its shielding analysis and submitted revised FSAR page changes to correct the conflicting information (ML24040A027). Therefore, the staffs review was expanded to review these changes, and the staffs evaluation of these changes is addressed in the Shielding Evaluation section of this SER.
Description of the VCC Lids The applicant noted that the NAC-MPC system has two types of lids that can be installed. The lids used on the Yankee Multi-Purpose Canister (MPC) and Connecticut Yankee MPC VCCs are a carbon steel closure, while the lid for the MPC-LACBWR [La Crosse Boiling Water Reactor]
3 VCC is made up of carbon steel with encased concrete material. The MPC-LACBWR VCC concrete and carbon steel lid design was introduced in CoC Amendment No. 6. This subcomponent is a thick concrete and carbon steel closure for the concrete cask. The licensing basis design function for the lid is to reduce skyshine radiation and to protect the transportable storage canister from the environment and postulated tornado missiles. The proposed changes for the concrete in the VCC lid in this application are only applicable to the MPC-LACBWR VCC lid.
Concrete Cask Lid Construction The NRC staff reviewed the proposed changes to TS for the VCC lid and noted that the proposed TS include the following controls for the concrete used in the construction of the concrete cask lid:
The concrete shall, at minimum, be a commercial grade ready-mix type that can develop a density of 140 pounds per cubic foot (pcf).
The concrete mix and batching should meet the purchasers requirement for density and any additional purchaser-indicated attributes, such as air content, as allowed by American Society for Testing and Materials (ASTM) International Standard ASTM C94, Standard Specification for Ready-Mixed Concrete.
The density of the concrete can be verified by either test method ASTM C138, Standard Test Method for Density (Unit Weight), Yield, and Air Content (Gravimetric) of Concrete, or an approved shop fabrication procedure by following the equation for density, where density is equal to weight divided by volume. The shop procedure shall include steps to weigh the lid before and after concrete placement and in calculating the actual volume of the cavity to be filled with a record of the weight of concrete placed into the cavity.
The concrete placement shall be in a dry and clean cavity or form with procedures and equipment that ensure the concrete placed is thoroughly consolidated and worked around any reinforcement and/or embedded fixtures and into the corners of the cavity or form.
The concrete shall be protected from the environment during curing to minimize development of cracks by one or more of various methods such as moist cure or liquid membrane forming chemicals. Type II Portland cement may be substituted by an alternate cement type for the concrete if the above density requirement can be met.
The NRC staff noted that the current NAC-MPC system TS (i.e., prior to the submittal of this application) do not differentiate between the VCC body and the VCC lid, and they require that the entirety of ACI 349 and ACI 318 govern the design and construction of the entire VCC (including the lid). These ACI standards provide comprehensive and detailed requirements for the design and construction of structural concrete. As discussed in more detail below, because the concrete used in the construction of the VCC lid has a radiation shielding safety function but no structural strength requirements, the staff considered whether the more limited set of requirements proposed in the TS for the concrete in the cask lid, in lieu of the requirements of ACI 349 and ACI 318, are suitable to ensure that the lid can adequately perform its radiation shielding safety function.
4 STRUCTURAL EVALUATION The NRC staff reviewed the MPC-LACBWR concrete cask lid drawings and descriptions in the FSAR, sections 1.7 and 1.A.7, for the three different cask designs, and confirmed that only the MPC-LACBWR concrete cask lid is made up of steel with encased concrete material. Therefore, the staff focused its review to determine whether the proposed changes to the TS altered previous safety findings, and to assess the impact of the changes on the structural performance of the MPC-LACBWR VCC. The staff reviewed the TS changes against the safety functions and design limits of the affected components as specified in the FSAR as summarized below.
The FSAR table 1.A-1 defines the MPC-LACBWR concrete cask lid as a thick carbon steel bolted closure with encased concrete for shielding for the storage cask. The lid precludes access to the canister and provides radiation shielding to reduce skyshine. The FSAR section 1.A.2.1.2 describes that the top of the storage cask is closed by a lid with integral radiation shield. The radiation shield is approximately 8-inch-thick concrete encased in a carbon steel shell extending into the cask cavity from the bottom surface of the 1.5-inch-thick carbon steel lid.
The specification summary for the encased concrete in the FSAR table 1.A.2-6 requires concrete mix to be a standard weight concrete density of 140 pcf (minimum), commercial grade concrete from a commercial grade supplier and there is no strength requirement. The staff reviewed the FSAR section 1.A.7 Drawing 630045-863, Rev. 2, which shows a cross-section of the cask lid assembly. Accordingly, the 1.5-inch-thick carbon steel cover plate spans the entire opening and is bolted onto the top of the concrete cask, while the 8-inch-thick encased concrete extends into the cask cavity from the bottom of the plate.
Per the FSAR section 3.A.4.3.1, the applicant designed the MPC-LACBWR concrete cask to be lifted vertically from its bottom end using four hydraulic jacks to allow insertion of an air pad system under the bottom end of the concrete cask. Also, the handling of the concrete cask lid is done by engaging the 1.5 thick steel closure plate, which in turn carries the weight of the encased concrete. In addition, per the FSAR section 11.A.2.13.2.2, the MPC-LACBWR concrete cask is designed to withstand the effects of impacts from the design basis tornado-generated missiles described in the FSAR section 2.2.1.3. The applicant evaluated that the 1.5-inch-thick steel lid of the MPC-LACBWR concrete cask is greater than the plate thickness required to prevent perforation by tornado-generated missiles.
Based on the above review, the NRC staff confirmed that there is no structural strength demand for the encased concrete in the cask lid. Thus, the only relevant licensing basis design requirement for the concrete in the lid is for radiation shielding. Further, the staff considered whether the limited requirements proposed in the TS B3.3 for the concrete in the cask lid, in lieu of the structural concrete requirements of ACI 349 and ACI 318, are suitable to ensure that the lid can adequately perform its required shielding safety function as discussed in detail below in the Materials Evaluation and Shielding Evaluation sections of this SER.
Based on the above review, the NRC staff concludes that the proposed changes in this amendment and accompanying revisions have been adequately described and evaluated; and that the NAC-MPC system continues to maintain structural integrity and meets the requirements of 10 CFR Part 72.
5 MATERIALS EVALUATION The NRC staffs materials review of the application addressed the concrete cask lid construction criteria and concrete degradation mechanisms applicable to the concrete in the cask lid.
The staff also verified that the proposed TS requirements for determining concrete density will ensure the measurement of concrete weight and volume are correctly performed and that the density is correctly calculated based on the measured weight divided by the measured volume.
The staff noted that these methods are sufficient to ensure that the density of the concrete in the lid meets the TS requirement of 140 pcf. Therefore, the staff determined that the proposed TS is acceptable for ensuring that the concrete in the lid will have the density needed to adequately perform its radiation shielding safety function.
For the commercial grade concrete in the lid to maintain physical characteristics needed to adequately perform the radiation shielding function, the NRC staff identified that the finished concrete should not undergo unacceptable shrinkage, and it should remain free of significant defects (such as voids and cracks) that could cause unacceptable radiation streaming through the concrete in the lid. Therefore, in addition to density, the staff considered whether the proposed TS requirements for construction of the VCC lid are adequate to ensure that the concrete can maintain the physical properties (i.e., no unacceptable shrinkage and no significant voids or cracks) needed to perform its radiation shielding safety function during the operating life of cask. The NRC staffs evaluation of potential concrete shrinkage that may result in a loss of radiation shielding performance is provided in the SER section below.
Concrete Shrinkage Concrete shrinkage is a reduction in the dimensions of a formed concrete component that occurs when hardened concrete dries from a saturated condition, as discussed in NUREG-2214. For concrete components of certain dimensions that are relied upon to provide radiation shielding (i.e., to reduce external dose rates to acceptable levels), concrete shrinkage may have the potential to cause a reduction in the dimensions of the component by an amount that results in unacceptable radiation streaming and unacceptable external dose rates. Concrete shrinkage occurs initially during curing and can be controlled through concrete formulation.
According to ACI 209R, Prediction of Creep, Shrinkage, and Temperature Effects in Concrete Structures, over 90 percent of shrinkage occurs during the first year, reaching 98 percent by the end of the first five years. Thus, concrete shrinkage is the most significant degradation mechanism that may impact radiation shielding performance of a concrete component during the initial years of storage following concrete fabrication, when the radioactivity of the spent fuel in dry storage is the highest.
In its May 14, 2024, RAI response (ML24135A322), the applicant included an evaluation of concrete shrinkage for the cask lid. The applicant calculated the expected radial gap around the edge of the concrete in the cask lid due to concrete shrinkage and determined that the gap will not affect the shielding performance of the concrete in the lid. The NRC staff compared the applicants calculation of the expected radial gap around the edge of the concrete in the cask lid (due to shrinkage) to the generic evaluation of concrete shrinkage in NUREG-2214 and found the licensees calculation of the expected radial gap to be acceptable for the 60-year approved storage term. The NRC staff confirmed that the expected radial gap around the edge of the concrete in the lid due to shrinkage will not have any impact on the radiation shielding function of the concrete in the cask lid, as discussed in the Shielding Evaluation section of this SER.
Therefore, the staff determined that the proposed TS controls for the concrete used in the
6 construction of the cask lid are sufficient to ensure that the expected concrete shrinkage in the cask lid will not adversely affect radiation shielding performance of the concrete in the lid during the 60-year approved storage term for the NAC-MPC System.
Other Concrete Degradation Mechanisms Over time, the concrete in the VCC lid may be prone to other degradation mechanisms, in addition to shrinkage, that could potentially have adverse effects on its ability to perform its radiation shielding function. Since there is no structural strength demand for the concrete used in the lid, other degradation mechanisms of potential concern are those that could cause the concrete in the lid to develop flaws, such as voids or cracks, that could potentially cause an increase in radiation dose rates through the concrete in the lid.
Section 3.5.1, Concrete, of NUREG-2214 provides a generic evaluation of potential aging degradation mechanisms and associated aging effects for concrete used in storage overpacks.
The staff considered the information on concrete degradation mechanisms to assess whether the proposed changes to the TS for the concrete in the cask lid could potentially result in increased susceptibility to aging deterioration that could have an adverse effect on the ability of the lid to perform its radiation shielding safety function during the approved 60-year storage term.
The staff noted that, for the specific NAC-MPC cask lid design, the concrete in the cask lid is completely encased in carbon steel. The staff identified that no significant intrusion of water, moisture, and dissolved compounds into the concrete, due to exposure of the outer surface of the carbon steel lid encasement to weather and precipitation, is expected, provided that the integrity of the carbon steel encasing the concrete is adequately maintained. The staff identified that since the carbon steel that encases the concrete in the cask lid is subject to aging management during the 60-year approved storage term in accordance with the aging management program for monitoring of metallic components, there is assurance that the integrity of the carbon steel for the cask lid will be adequately maintained to prevent the intrusion of water into the concrete. Therefore, the only aging degradation mechanisms that are potentially credible for non-structural encased concrete used for radiation shielding are shrinkage (addressed in the SER section above), dehydration at high temperature, and delayed ettringite formation (DEF).
Dehydration at high temperature could potentially contribute to cracking and may further exacerbate concrete shrinkage at sufficiently high temperatures if the concrete is not adequately fabricated. Considering the limit on the maximum bulk concrete temperature for normal operation specified in the FSAR, and the fact that fuel temperature decreases over time, the NRC staff confirmed that the proposed TS criteria for the concrete are sufficient to ensure that the concrete in the cask lid will not be prone to significant degradation due to dehydration at high temperatures.
DEF is a degradation mechanism characterized by the early-stage conversion of the mineral ettringite to monosulfoaluminate during curing at sufficiently high temperatures (greater than about 158 °F), and subsequent reconversion back to ettringite after the concrete hardens. This degradation mechanism may lead to concrete volume expansion and increased internal residual stresses, which could result in concrete cracking and spalling. As addressed in NUREG-2214, DEF of concrete is not considered credible for dry storage casks in outdoor, sheltered, below-grade, and fully encased environments, provided that adequate concrete placement and curing standards, such as those in ACI 349 and ACI 318, are followed. While the proposed TS change
7 removes these ACI standards, the staff confirmed that the specification of ASTM C94 for ready-mixed concrete and the additional specification that concrete shall be protected from the environment during curing are sufficient to ensure that DEF will not cause degradation that results in an unacceptable decrease in radiation shielding performance during the operating life of the cask.
Considering the potential aging degradation mechanisms, the staff determined that the proposed TS for the VCC lid are sufficient to ensure that the concrete in the lid will maintain the physical properties needed to adequately perform its radiation shielding safety function. The staff also determined that the applicant demonstrated that the properties of the concrete cask lid support the safe storage of spent nuclear fuel, and the applicant has met the requirements in 10 CFR 72.236(g).
SHIELDING EVALUATION The objective of the shielding review of the proposed changes to the NAC-MPC System is to ensure that the design features relied on for shielding provide adequate protection against direct radiation.
Concrete Density The staff identified that two different values for minimum concrete density for the MPC-LACBWR VCC lid were referenced in the FSAR design bases, and the staff requested clarification from the applicant in the January 10, 2024, Observation (ML23360A160). NAC responded on February 14, 2024 (ML24040A027), noting that the concrete cask shielding analysis in the FSAR incorrectly used a lid concrete density of 145 pcf, rather than the design minimum density of 140 pcf. NAC revised the shielding analysis using a concrete density of 140 pcf and submitted a revised shielding calculation (63004500-5013, Rev. 0) to evaluate the change to dose rates using the design minimum density of 140 pcf and also evaluate the impact on site boundary dose rates, compared to the existing shielding calculation 63004500-5011 in the FSAR that used a 145 pcf density. NAC also submitted the corresponding page changes to the FSAR Shielding chapter 5.A.
For the dose rate analysis, the applicant used existing inputs for the concrete cask imported from 63004500-5011, NAC International, LACBWR Storage & Transfer Cask Shielding Analysis, Calculation 63004500-5011 Rev. 1. The computer code used for the dose rate analysis was MCNP6.2 version. The applicant shows the top surface dose rate profile for the baseline analysis (MCNP5 at 145 pcf lid concrete) and the reduced density (MCNP6 at 140 pcf lid concrete) in figure 5.A.4-32 of the FSAR page changes in revision 24A submitted on February 14, 2024 (ML24040A027). Based on the applicant dose rate calculations for undamaged fuel, the staff found that the dose rates increase over the center of cask (axial) and over the annulus between the canister outside diameter and the concrete cask inside diameter.
The increase in maximum top axial dose rates, using the design minimum concrete density of 140 pcf, is within the statistical uncertainty (1) of the results. The increase in average dose rates over the top of the cask, using the design minimum concrete density of 140 pcf, exceeds the 1 threshold. Given that top axial dose rates are a maximum of 2.2% of the downfield dose rates, the 8.6% increase in average dose rates equates to an overall increase in downfield dose rates of less than 0.2%. Therefore, the increase in downfield dose rates is negligible, and thus, the increase in site boundary dose rates and ISFSI occupational dose rates is negligible. In terms of damaged fuel dose rates results, a recalculation of storage cask top dose rates with 140 pcf lid concrete is not required because the dose rates for damaged fuel are bounded by
8 the dose rates for undamaged fuel. The staff finds that the change in dose rates using the design minimum density of 140 pcf is acceptable because the top axial dose rates are within the statistical uncertainty (1) of the results, and the increase in site boundary dose rates and ISFSI occupational dose rates is negligible.
Further, the staff considered whether the requirements proposed in TS B3.3 for the concrete in the cask lid, in lieu of the structural concrete requirements of ACI 349 and ACI 318, are suitable to ensure that the lid can adequately perform its required shielding safety function. ACI 349, section R3.3.1 states that shielding requirements for concrete components are dependent on the density of the concrete. Section R1.4 of ACI 349 cites American National Standard Institute and American Nuclear Society (ANSI/ANS) 6.4, Nuclear Analysis and Design of Concrete Radiation Shielding for Nuclear Power Plants, as specific guidance for evaluating the radiation shielding effectiveness of concrete components. The staff verified that the minimum required 140 pcf concrete density in the proposed TS B3.3 is the minimum acceptable density per table 1 of ANSI/ANS-6.4, for ordinary concrete that performs a radiation shielding function. The staff verified that the proposed TS requirements for determining concrete density per ASTM C138, or equivalent procedure will ensure the density is correctly calculated based on the measured data. The staff also verified that the proposed TS requirements for placement and curing of concrete will ensure that cast-in place concrete is free of unacceptable cracks and voids.
Therefore, the staff determined that the proposed TS requirements are acceptable for ensuring that the concrete in the lid will perform its radiation shielding safety function. The staff found that the requested changes to the TS did not alter the previous safety findings and therefore, are acceptable.
Concrete Shrinkage The applicant used the previously approved CoC No. 1031 Amendment No. 12 for NACs MAGNASTOR storage system (ML23250A364) and its estimate of the effects on dose rate due to concrete shrinkage on the MAGNASTOR overpack lid as the basis to perform an evaluation to estimate the effects of concrete shrinkage on the NAC-MPC VCC lid. The MAGNASTOR evaluation focuses on the potential effects that any radial concrete shrinkage would have on the lids ability to perform its radiation shielding safety function. In NACs supplement for the MAGNASTOR application, dated March 18, 2022 (package ML22077A768), Enclosure 2, Potential Concrete Radial Shrinkage, the applicant stated that based on publicly available literature, for every 100 feet (ft) of concrete there is about 0.6 inches of shrinkage. Applying this approach to the MAGNASTOR overpack concrete top lid yields approximately 0.039 inches of shrinkage on the diameter [78.5 inches x (0.6 in./(100 ft x 12 inches)) = 0.039 inches], which is equivalent to about a 0.02-inch radial gap around the edge of the concrete cask lid. For the approved MAGNASTOR Amendment No. 12, the staff found that the expected radial gap around the edge of the concrete in the lid due to shrinkage did not impact either occupational doses or site boundary doses, as the dose rate results fall within the statistical uncertainty band of the original calculations. The staff concluded that the shielding and radiation protection design features of the MAGNASTOR system, including the changes to the requirements in TS A4.2, which specify the concrete used in the construction of the top lid concrete cask, are in compliance with 10 CFR Part 72, and that the applicable design and acceptance criteria continue to be satisfied.
The changes to MAGNASTOR TS A4.2 (regarding the fabrication requirements for the concrete in the lid) approved in Amendment No. 12 are identical to the proposed changes for this NAC-MPC application. In its supplements to this NAC-MPC application, the applicant noted that the NAC-MPC concrete lid geometry is documented in licensing drawing 630045-863
9 (ML24135A322 and ML24240A133). The staff confirmed the lid steel ring outer diameter is 78.0 inches with the ring material being 3/8 inch thick producing a concrete cavity diameter of 77.25 inches. This is similar to the 78.5-inch diameter evaluated for MAGNASTOR. Using the same concrete shrinkage approximation applied in the MAGNASTOR application yields approximately 0.039 inches of shrinkage on the diameter [77.25 inches x (0.6 in./(100 ft x 12 inches)) = 0.039 inches], which is an approximate 0.02-inch radial gap around the edge of the concrete cask lid. Within rounding this value is identical to the gap calculated for MAGNASTOR. Thus, the evaluation presented for the approved MAGNASTOR Amendment No. 12 is reasonably representative of the NAC-MPC system, and the staff considerations and findings from the MAGNASTOR Amendment No. 12 review are relevant for this NAC-MPC application. Based on the findings from the MAGNASTOR Amendment No. 12 review and as discussed above in the Materials Evaluation section of this SER, the staff finds that shrinkage will not have any impact on the radiation shielding function of the concrete in the cask lid since dose rates are shown to be unaffected by this small dimensional change. The staff concludes that the dose to any real individual beyond the controlled area and the occupational doses from the storage cask are bounded by previous analyses. Therefore, the proposed changes to the NAC-MPC system are acceptable.
CONCLUSION For the reasons stated in this SER and based on the statements and representations in NACs application, as supplemented, the NRC staff concludes that the proposed changes do not affect the ability of the storage cask system to meet the requirements of 10 CFR Part 72. Amendment No. 9, and Revision to Amendment Nos. 6, 7, and 8, of CoC No. 1025 for the NAC-MPC system should be approved.
Issued with Renewed Certificate of Compliance No. 1025, Amendment No. 9 and Revision 1 to Amendment Nos. 6, 7, and 8 On _____________.