ML20155K745

Certificate of Compliance No. 1032 Amendment No. 4 Final Safety Evaluation Report

ML20155K745
Person / Time
Site:	07201032
Issue date:	06/15/2020
From:	John Mckirgan Office of Nuclear Material Safety and Safeguards
To:	Holtec
YJChen NMSS/DFM/STL 415.1018
Shared Package
ML20155K740	List: ML20155K741 ML20155K742 ML20155K743 ML20155K744 ML20155K745
References
CAC 001028, EPID L-2017-LLA-0030
Download: ML20155K745 (34)

Text

FINAL SAFETY EVALUATION REPORT DOCKET NO. 72-1032 HOLTEC INTERNATIONAL HI-STORM FLOOD/WIND MULTIPURPOSE CANISTER STORAGE SYSTEM CERTIFICATE OF COMPLIANCE NO. 1032 AMENDMENT NO. 4

SUMMARY

This safety evaluation report (SER) documents the U.S. Nuclear Regulatory Commission (NRC) staffs (staff) review and evaluation of the amendment request to amend Certificate of Compliance (CoC) No. 1032 for the HI-STORM Flood/Wind (FW) Multipurpose Canister (MPC)

Storage System submitted by Holtec International (Holtec) by letter dated March 11, 2016 (Holtec, 2016a), and supplemented on September 16, 2016 (Holtec, 2016b), January 31, 2017 (Holtec, 2017a), April 27, 2018 (Holtec, 2018a), July 27, 2018 (Holtec, 2018b), April 12, 2019 (Holtec, 2019a), June 11, 2019 (Holtec, 2019b), and July 5, 2019 (Holtec, 2019c). Holtec proposed the following changes:

1. Add MPC-32ML for storage in HI-STORM FW system and allow fuel assembly class 16x16D as content for MPC-32ML.

2. Add fuel assembly class 16x16E as content for MPC-37.

3. Separate the design pressure for the short-term operation from the off-normal condition to provide clarity in final safety analysis report (FSAR) Table 2.2.1.

4. Add a caution note in FSAR Section 9.2.1 that states fuel cladding is not exposed to air during loading operations.

5. Update the definition of undamaged fuel assembly in FSAR Glossary to be aligned with the definition in Appendix A and FSAR Table 2.1.3 (Note 14).

6. Replace Charpy test program with fracture toughness test program from the revised Metamic-HT Sourcebook (Holtec, 2017b) in FSAR Sections 1.2.1.4.1 and 3.4.

7. Add a caution note in FSAR Section 9.2.3 that states low-enriched fuel must be shown to be without known or suspected grossly breached rods.

The staff did not evaluate the original request to allow for gadolinium credit for certain boiling water reactor (BWR) fuel assemblies and the addition of MPC-31C for storage of VVER fuel type. Both proposed changes were removed from Amendment No. 4 when responding to NRCs request for additional information (RAI) (Holtec, 2018a).

This revised CoC, when codified through rulemaking, will be denoted as Amendment No. 4 to CoC No. 1032.

Enclosure 4

This SER documents the staffs review and evaluation of the proposed amendment. The staff followed the guidance in NUREG-1536, Revision 1, Standard Review Plan for Dry Cask Storage Systems at a General License Facility, July 2010 (NRC, 2010). The staffs evaluation is based on a review of Holtecs application and supplemental information to determine whether it meets the applicable requirements of Title 10 of 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 staffs evaluation focused only on modifications requested in the proposed amendment and did not reassess previous revisions of the FSAR nor previous amendments to the CoC.

1.0 GENERAL INFORMATION EVALUATION The purpose of the review is to ensure that the applicant has provided in its documentation for the spent fuel storage system a non-proprietary description, or overview, that is adequate to familiarize reviewers and other interested parties with the pertinent features of the system. The following proposed changes are applicable to the general information evaluation:

• Proposed Change #1: Add MPC-32ML for storage in HI-STORM FW system.

• Proposed Change #5: Update the definition of undamaged fuel assembly in FSAR Glossary to be aligned with the definition in Appendix A and FSAR Table 2.1.3.

• Proposed Change #6: Replace Charpy test program with fracture toughness test program from the revised Metamic-HT Sourcebook in FSAR Sections 1.2.1.4.1 and 3.4.

The applicant proposed to update the definition of undamaged fuel assembly in FSAR Glossary to be aligned with the definition in the CoC Appendix A and FSAR Table 2.1.3; add MPC-32ML related information in the FSAR; and replace Charpy test program with fracture toughness test program from Metamic-HT Sourcebook in the FSAR. The staff determined that the description in the proposed changes to FSAR Chapter 1, General Description, is adequate to allow the staffs detailed evaluation as documented in the following sections of this SER.

2.0 PRINCIPAL DESIGN CRITERIA EVALUATION The objective of evaluating the principal design criteria related to structures, systems, and components (SSCs) important to safety is to ensure that the principal design criteria comply with the relevant general criteria established in the requirements in 10 CFR Part 72. The following proposed changes are applicable to the principal design criteria evaluation:

• Proposed Change #1: Add MPC-32ML for storage in HI-STORM FW system and allow fuel assembly class 16x16D as content for MPC-32ML.

• Proposed Change #2: Add fuel assembly class 16x16E as content for MPC-37.

• Proposed Change #3: Separate the design pressure for the short-term operation from the off-normal condition to provide clarity in FSAR Table 2.2.1.

The applicant revised Chapter 2, Principal Design Criteria, of the FSAR to add Quality Assurance Safety Category for MPC-32ML enclosure vessel and fuel basket in Tables 2.0.9 and 2.0.10, respectively. Since MPC-37 is the bounding cask, the staff compared the safety category of MPC-32ML components with that of MPC-37 enclosure vessel (Table 2.0.2) and fuel basket (Table 2.0.3) and noted that the safety categories are consistent. Therefore, the 2

staff concludes the quality assurance safety categories for MPC-32ML in Tables 2.0.9 and 2.0.10 are acceptable.

The applicant also revised Chapter 2 of the FSAR to add descriptions related to MPC-32ML, including radiological parameters for spent fuel and non-fuel hardware, additional fuel type 16x16D, burnup, and cooling time. The staff determines the principal design criteria comply with the relevant general criteria established in 10 CFR Part 72 as documented in Sections 3 through 8 of this SER.

The applicant included fuel characteristics of fuel assembly class 16x16E as the content for MPC-37 in FSAR Chapter 2. The staff determines the principal design criteria comply with the relevant general criteria established in 10 CFR Part 72 as documented in Sections 3, 6, and 8 of this SER.

Proposed Change #3 modifies the design pressures in FSAR Table 2.2.1 to separate the short-term operation design pressure from the off-normal condition. The staff reviewed the proposed change and determined that this change clarifies the design pressure to be used in normal, short-term, off-normal, and accident conditions. In addition, the short-term operation design pressure has been lowered, which is more conservative. Therefore, the staff concludes this change is acceptable.

3.0 STRUCTURAL EVALUATION The staff reviewed the proposed changes to verify that the application has performed adequate structural evaluation to demonstrate that the system, as proposed, would be acceptable under normal and off-normal operations, accident conditions, and natural phenomena events. In conducting this evaluation, the staff seeks reasonable assurance that the cask system will maintain confinement, subcriticality, radiation shielding, and retrievability or recovery of the fuel, as applicable, under all credible loads for normal, off-normal conditions, accidents, and natural phenomenon events.

In addition to the application and the supplemental information, the staff also reviewed Holtecs proprietary position paper DS-307, Construction of True-Stress-True-Strain Curve for LS-DYNA Simulations, Revision 2. The design criteria utilized for the structural evaluation are identical to the criteria described in the original HI-STORM FW FSAR.

The following proposed changes are applicable to the structural evaluation:

• Proposed Change #1: Add MPC-32ML for storage in HI-STORM FW system.

• Proposed Change # 2: Add fuel assembly class 16x16E as content for MPC-37.

3.1 HI-STORM FW Overpack The staff reviewed the amendment application and supplemental documents to ensure that the applicant has verified the adequacy of the MPC-32ML enclosure vessel with fuel assemblies planned to be stored as the payload. The staff noted that the fuel weights (as shown in the FSAR Tables 3.2.3 and 3.2.4) to allow storage of MPC-32ML (carrying pressurized-water reactor [PWR] fuel) in the HI-STORM FW cask, are bounded by weights that were previously approved by the NRC for storing MPC-37 and MPC-89 in HI-STORM FW cask. The staff also noted that the total cask weight including the MPC-32ML enclosure vessel is less than the 3

bounding weight used for the lifting lug analysis previously approved by the staff for storing MPC-37 and MPC-89.

As the payload is bounded by previously approved analysis and satisfies the acceptance criteria for the lifting, the staff concludes that the fuel assemblies payload will maintain its structural integrity for the accident conditions during the storage period and will continue to comply with the structural adequacy requirements of the applicable regulations.

3.2 Fuel Rod Cladding during MPC Reflooding The staff reviewed the applicants furnished finite element analysis using the ANSYS code for the lower portion of the fuel cladding, when subjected to quenching due to MPC-32ML reflood event.

The staff noted that the applicant performed the analysis using several conservative assumptions, such as ignoring the effects of slowing down the thermal transient by the vapor blanketing the immersed portion of the fuel rod and using a conservative lower bound value of failure strain of 1.7% as the permissible limit.

The applicant reported that the maximum induced stress in MPC-32ML, the governing MPC, was well below the yield limit. Based on this, the staff concludes that the fuel cladding will remain elastic. In addition, the applicants analysis indicated that the maximum strain was less than the failure strain limit, rendering a safety factor of 6.0. Therefore, the staff determines that the reflood event will not cause a breach of the fuel rod cladding in MPC-32ML.

3.3 Non-mechanistic Tip-Over Analysis The staff reviewed Holtecs proprietary report HI-2166998, Analysis of the Non-Mechanistic Tipover Event of the HI-STORM FW Storage Cask Loaded with MPC-32ML and MPC-31C, Revision 0. The results of the LS-DYNA finite element analysis of the bounding tip-over events for the HI-STORM FW storage cask using conservative assumptions indicated that the fuel basket, the MPC enclosure vessel, and the cask overpack experienced only local plastic deformation. Moreover, the localized permanent deformation was limited to the periphery storage walls, and the overall shielding capacity of the cask was not compromised. As the cask lid bolts were found structurally safe, the cask lid will not dislodge after the tip-over event.

Based on the review, the staff determines that the HI-STORM FW cask satisfies the design criteria for normal, off-normal, and the accident conditions, and is in compliance with the storage requirements of 10 CFR Part 72.

3.4 Addition of Fuel Assembly Class 16x16E as Content for MPC-37 The staff reviewed the structural designs of the fuel assemblies (16x16A, 16x16B, 16x16C, 16x16D, and 16x16E) and found that their designs are almost identical. The total weight of each fuel assembly is bounded by the weights that were previously accepted by the NRC for the analysis of the HI-STORM FW cask. Therefore, the addition of fuel assembly class 16x16E as content for MPC-37 is acceptable.

3.5 Evaluation Findings

F3.1 The FSAR adequately describes SSCs that are important to safety, providing drawings and text in sufficient detail to allow evaluation of the structural effectiveness.

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F3.2 The applicant has met the requirements of 10 CFR 72.236(b). The SSCs important to safety are designed to accommodate the combined loads of normal or off-normal operating conditions and accidents or natural phenomena events with an adequate margin of safety and are found to be within limits of applicable codes, standards, and specifications. The staff has reasonable assurance that the addition of MPC-32ML for storage in HI-STORM FW system is acceptable.

F3.3 The staff determines the addition of fuel assembly class 16x16E as a content for MPC-37 is acceptable because it is bounded by previous analysis accepted by the NRC for the analysis of the HI-STORM FW cask.

F3.4 The applicant has met the requirements of 10 CFR 72.236(l) that the design analysis and bases used for evaluation demonstrate that the cask and other systems important to safety will reasonably maintain confinement of radioactive material under normal, off-normal, and credible accident conditions.

F3.5 The applicant has met the specific requirements of 10 CFR 72.236(g) and (h) as they apply to the structural design for spent fuel storage cask approval. The cask system structural design acceptably provides for the following required provisions:

• Storage of the spent fuel for the minimum required years.

• Compatibility with wet or dry loading and unloading facilities.

Based on the review of the applicants description, proposed design criteria, appropriate use of material properties and adequate structural analysis of the relevant structures, systems and components, the staff concludes that the SSCs of the HI-STORM FW are in compliance with 10 CFR Part 72 required regulations.

4.0 THERMAL EVALUATION The staff reviewed the proposed changes to the HI-STORM FW Amendment No. 4 to ensure that the cask components and fuel material temperatures of the HI-STORM FW cask system will remain within the allowable values under normal, short-term off-normal, and accident conditions.

These objectives include confirmation that the fuel cladding temperature will be maintained below specified limits throughout the storage period to protect the cladding against degradation that could lead to gross ruptures. The staff also confirms that the cask thermal design has been evaluated using acceptable analytical techniques and/or testing methods.

The staff conducted the review against the appropriate regulations, as described in 10 CFR 72.236, which identify specific requirements for the approval and fabrication of spent fuel storage cask designs. The unique characteristics of the spent fuel to be stored are identified, as required by 10 CFR 72.236(a), so that the design basis and the design criteria that must be provided for the structures, systems, and components important to safety can be assessed under the requirements of 10 CFR 72.236(b). The staff also determines whether the HI-STORM FW design fulfills the acceptance criteria listed in Sections 2, 4, and 12 of NUREG-1536, Revision 1, as well as associated Spent Fuel Storage and Transportation (SFST) Interim Staff Guidance (ISG) documents.

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The following proposed change is applicable to the thermal evaluation:

• Proposed Change #1: Add MPC-32ML for storage in HI-STORM FW system and allow fuel assembly class 16x16D as content for MPC-32ML.

4.1 Spent Fuel Cladding FSAR Table 4.4.3 shows the maximum temperatures of MPC-37 and MPC-32ML under long-term normal storage. The predicted temperatures of MPC-37 is higher than that of MPC-32ML, so MPC-37 is still the limiting MPC. Therefore, the resulting spent fuel cladding from the proposed addition of MPC-32ML is bounded by previous off-normal and accident condition evaluations in Amendment Nos. 0 through 3. Therefore, the previous evaluation continues to be acceptable to the staff and an evaluation is not required.

4.2 Thermal Properties of Materials The proposed changes do not impact thermal properties of materials which was previously reviewed by the staff in Amendment No. 0, and materials have not changed since Amendment No. 0. Therefore, the previous evaluation continues to be acceptable to the staff and an evaluation is not required.

4.3 Specifications for Components The proposed changes do not impact specifications of the storage system components which was previously reviewed by the staff. Material temperature limits have not changed since Amendment No. 0. Therefore, the previous evaluation continues to be acceptable to the staff and an evaluation is not required.

4.4 HI-STORM FW System 4.4.1 General Description A general description of the HI-STORM FW storage system is provided in the FSAR. Holtec requested to add one additional canister, MPC-32ML, to the system. Table 1.2.1 of the application provides the storage capacity of the new canister and its allowable contents.

4.4.2 Design Criteria An evaluation of the general design criteria has been performed previously. The proposed changes do not affect the design criteria. Therefore, the previous evaluation continues to be acceptable to the staff, and an evaluation is not required.

4.4.3 Design Features An evaluation of the general design features of the HI-STORM FW system has been performed previously in the original application and an evaluation is not required. MPC-32ML is designed for a maximum heat load of 44.16 kW, which is bounded by the design basis of the HI-STORM FW system.

The staff reviewed the applicants general description, design criteria, and design features of the HI-STORM FW storage system. Based on the information provided in the application and the 6

fact that the maximum heat load of MPC-32ML is bounded by the system design basis, the staff concludes that the description of the decay heat removal system is acceptable as the description is consistent with NUREG-1536 which satisfies the requirements of 10 CFR 72.236(b), 72.236(f), 72.236(g), and 72.236(h).

4.5 HI-STORM FW System Thermal Model The applicant used FLUENT computer-based analysis program to evaluate the thermal performance of the HI-STORM FW spent fuel storage system. FLUENT is a finite volume computational fluid dynamics (CFD) program with capabilities to predict fluid flow and heat transfer phenomena in two and three dimensions. Section 4.4.1.1 of the application provides a detailed description of the thermal model developed by the applicant for the new canister.

The staff reviewed the applicants description of the HI-STORM FW storage system thermal model. Based on the information provided in the application regarding the thermal model, the staff determined that the application is consistent with guidance provided in Section 4.4.4 (Analytical Methods, Models, and Calculations) of NUREG-1536. Therefore, the staff concludes that the description of the thermal model is acceptable as the description is consistent with NUREG-1536 which satisfies the requirements of 72.236(b), 72.236(f), 72.236(g), and 72.236(h).

4.6 Thermal Evaluation for Normal Conditions of Storage The applicant used the 3-D model, described in the previous section, to determine temperature distributions under long-term normal storage conditions for the MPC-32ML. The applicant performed screening calculations to determine which MPC type and loading pattern would result in the highest peak cladding temperature. The comparisons in FSAR Tables 4.4.2 and 4.4.3 show that the MPC-37 is the limiting MPC.

The applicant calculated the maximum gas pressure in the MPC for a postulated release of fission product gases from fuel rods into the MPC free space. For these scenarios, the amounts of each of the release gas constituents in the MPC cavity are summed and the resulting total pressures determined from the ideal gas law. Based on fission gases release fractions (NUREG 1536 criteria), rods net free volume, and initial fill gas pressure, maximum gas pressures with 1% (normal), 10% (off-normal), and 100% (accident condition) rod rupture are given in FSAR Table 4.4.5 for MPC-37, MPC-89, and MPC-32ML. The maximum computed gas pressures reported in FSAR Table 4.4.5 are all below the MPC internal design pressures for normal, off-normal, and accident conditions as specified in the FSAR Table 2.2.1.

The staff reviewed the applicants thermal evaluation of the HI-STORM FW storage system during normal conditions of storage. Based on the information provided in the application regarding the thermal model and evaluation, the staff determined that the application is consistent with guidance provided in Section 4.4.4 (Analytical Methods, Models, and Calculations) of NUREG-1536.

4.7 Thermal Evaluation for Short-Term Operations For the MPC-32ML containing moderate burnup (MBU) fuel assemblies only, drying operations may be performed using the conventional vacuum drying approach up to design basis heat load. Vacuum drying of MPC-32ML containing high burnup fuel assemblies is permitted up to threshold heat loads defined in Table 4.5.16 of the FSAR.

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The vacuum drying option is evaluated for the two limiting scenarios defined in FSAR Section 4.5.2.3 to address MBU fuel under limiting heat load in FSAR Table 1.2.3b (MPC-32ML) and high burnup (HBU) fuel under threshold heat load defined in FSAR Table 4.5.16 (MPC-32ML).

The principal objective of the analysis is to ensure consistence with SFST-ISG-11, Revision 3, temperature limits; therefore, 3-D thermal analysis models of the MPC-32ML canister were constructed in the FLUENT CFD code, as described in FSAR Section 4.5.2.3, and bounding steady state temperatures were computed. The results are tabulated in FSAR Table 4.5.17 and show that the cladding temperatures are below 400°C (752°F) for normal conditions of storage and short-term loading operations and 570°C (1,058°F) for off-normal and accident conditions, which is consistent with the guidance in SFST-ISG-11, Revision 3, for MBU and HBU fuel. The analysis presented above supports the MPC drying options summarized in Table 4.5.18.

The staff reviewed the applicants thermal evaluation of the HI-STORM FW storage system during drying operations. Based on the staffs audit of the applicants thermal model and the information provided in the application regarding the thermal analysis model and evaluation that the cladding temperatures are below the limits described in SFST-ISG-11, Revision 3, the staff determined that the application is consistent with guidance provided in Section 4.4.4 of NUREG-1536.

4.8 Off-Normal and Accident Events 4.8.1 Off-Normal Events The proposed changes do not impact this section which was previously reviewed by the staff in Amendment No. 0. MPC-37 analysis results continue to bound all MPC types, including the new canister, MPC-32ML. Therefore, the previous evaluation continues to be acceptable to the staff and an evaluation is not required.

4.8.2 Accident Events The proposed changes do not impact this section which was previously reviewed by the staff in Amendment No. 0. MPC-37 analysis continues to bound all MPC types, including the new canister, MPC-32ML. Therefore, the previous evaluation continues to be acceptable to the staff and an evaluation is not required.

4.9 Confirmatory Analyses The staff reviewed the applicants thermal models used in the analyses, checked the code input in the calculation packages, and confirmed that the proper material properties and boundary conditions were used. The staff verified that the applicants selected code models and assumptions were adequate for the flow and heat transfer characteristics prevailing in the HI-STORM FW geometry and analyzed conditions.

Engineering drawings were also consulted to verify that adequate geometry dimensions were translated to the analysis models. The material properties presented in the FSAR were reviewed to verify that they were appropriately referenced and applied. The staff assured that the applicant performed appropriate sensitivity analysis calculations to obtain mesh-independent results that would provide bounding predictions for all conditions analyzed in the application.

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4.10 Evaluation Findings F4.1 Chapter 2 of the FSAR describes SSCs important to safety to enable an evaluation of their thermal effectiveness. Cask SSCs important to safety remain within their operating temperature ranges.

F4.2 The HI-STORM FW storage system is designed with a heat-removal capability having verifiability and reliability consistent with its importance to safety. The cask is designed to provide adequate heat removal capacity without active cooling systems.

F4.3 The spent fuel cladding is protected against degradation leading to gross ruptures under long-term storage by maintaining cladding temperatures below 752°F (400°C).

Protection of the cladding against degradation is expected to allow ready retrieval of spent fuel for further processing or disposal.

F4.4 The spent fuel cladding is protected against degradation leading to gross ruptures under off-normal and accident conditions by maintaining cladding temperatures below 1,058°F (570°C). Protection of the cladding against degradation is expected to allow ready retrieval of spent fuel for further processing or disposal.

F4.5 The staff finds that the thermal design of the HI-STORM FW storage system is in compliance with 10 CFR Part 72 and that the applicable design and acceptance criteria have been satisfied. The evaluation of the thermal design provides reasonable assurance that the cask will allow safe storage of spent fuel. This finding is reached based on a review that considered the regulation itself, appropriate regulatory guides, applicable codes and standards, and accepted engineering practices.

5.0 CONFINEMENT EVALUATION The staff reviewed the proposed changes in HI-STORM FW Amendment No. 4 to ensure that any possible radiological releases to the environment continue to remain within the limits established in 10 CFR 72.104(a) and 10 CFR 72.106(b). The following proposed change is applicable to the confinement evaluation:

• Proposed Change #1: Add the MPC-32ML canister to the HI-STORM FW storage system.

5.1 Confinement System The confinement boundary of MPC-32ML includes the canister shell, baseplate, closure lid, vent port cover plate, drain port cover plate, and closure ring. The confinement boundary of MPC-32ML contains no valves or other pressure relief devices.

The applicant stated in the application that the canisters to be stored in the HI-STORM FW storage system meet the guidance of SFST-ISG-18, Revision 1, and the leakage of radiological matter from the confinement boundary is non-credible. As such, a dose analysis of leakage from the canister is not required. The applicant verified that the confinement function of the MPCs is maintained through pressure testing; helium leak testing of the MPC shell, base plate, and lid material along with the shell to base plate and shell to shell seam welds; and a rigorous weld examination regimen executed in accordance with the acceptance test program described in FSAR Chapter 10, Acceptance Criteria and Maintenance Program.

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The staff reviewed the application and finds that the confinement design of the MPC-32ML remains unchanged from MPC-32 that was reviewed and approved by the NRC for use in HI-STORM 100 system, Amendment No. 1 (NRC, 2002). The design of HI-STORM FW system is similar to HI-STORM 100 system with additional features for flood and wind protection. The staff finds the proposed addition of MPC-32ML has no impact to the confinement design, and the confinement boundary of the MPC-32ML remains leaktight during normal, off-normal, and accident conditions. Therefore, the staff concludes that the MPC-32ML meets the guidance of SFST-ISG-18, Revision 1. Based on these findings, the staff concludes that the spent fuel cladding and fuel assemblies will be sufficiently protected against degradation that might otherwise lead to gross rupture during storage.

The applicant did not request changes to the confinement criteria/design and confinement components. For this reason, the staff concludes the confinement criteria continue to comply with the general criteria for confinement performance of a storage system established in 10 CFR Part 72.

5.2 Evaluation Findings

F5.1 Based on the staff's evaluation of information provided in this amendment request, the staff finds that there are no design changes to the confinement system and therefore the addition of the MPC-32ML to the HI-STORM FW system in the amendment has no adverse impact on the confinement effectiveness of the cask confinement system. The staff therefore concludes that HI-STORM FW Cask System continues to ensure that potential radiological release to the environment remains in compliance with 10 CFR 72.104(a) and 10 CFR 72.106(b).

6.0 SHIELDING EVALUATION The objective of the shielding review is to evaluate the ability of the proposed shielding design features of the HI-STORM FW system in Amendment No. 4 to determine if it would provide adequate protection against direct radiation from the dry storage system (DSS) contents, and ensure the shielding design is capable of meeting the operational dose requirements of 10 CFR 72.104 and 72.106 in accordance with the regulatory requirements of 10 CFR 72.236(d).

The staff reviewed the application and the associated shielding analyses for the revised system design with the new MPC-32ML. The staff followed the guidance provided in NURG-1536 in its review.

The following proposed changes are applicable to the shielding evaluation:

• Proposed Change #1: Add a new Westinghouse 16x16D type PWR spent fuel for storage in a new fuel canister design, MPC-32ML

• Proposed Change #2: Add fuel assembly class 16x16E content for MPC-37.

6.1 Addition of Westinghouse 16x16D type PWR spent fuel for new MPC-32ML Canister The applicant requested approval to add a new fuel canister design, MPC-32ML, to store a new Westinghouse 16x16D type PWR spent fuel. The new fuel contains more uranium and different fuel characteristics, e.g., larger fuel rod size and longer length, in comparison to the previously authorized contents of the HI-STORM FW dry cask spent fuel storage system. MPC-32ML can 10

also accommodate up to 8 damaged fuel assemblies in DFCs. DFCs may be stored in fuel storage locations 1-1, 1-4, 1-5, 1-10, 1-23, 1-28, 1-29, and 1-32 (Figure 2.1.1b of the SAR). The remaining fuel storage locations may be filled with PWR undamaged fuel assemblies meeting the applicable specifications. The damaged fuel assemblies must meet the undamaged fuels criteria for array/class 16x16D in Table 2.1.2 of the FSAR. Therefore, in the applicants shielding analysis, damaged fuel is treated as undamaged.

6.1.1 Design Criteria The dose to any individual at or beyond the controlled area boundary is required to be below 25 mrem per year. The minimum distance to the controlled area boundary is 100 meters from the ISFSI. Table 5.1.3 presents the annual dose to an individual from a single HI-STORM FW cask with MPC-37 (the bounding MPC) and various storage cask array configurations, assuming an 8,760 hours0.0088 days <br />0.211 hours <br />0.00126 weeks <br />2.8918e-4 months <br /> annual occupancy at the dose point location. The minimum distance required for the corresponding dose is also listed. The final determination of site-specific ISFSI dose rates at the site boundary and demonstration of compliance with regulatory limits is to be performed in accordance with 10 CFR 72.212.

The overpack is designed to limit the calculated surface dose rates on the cask for all MPC designs as defined in Section 2.3.5 of the FSAR. The overpack is also designed to maintain occupational exposures ALARA, in accordance with 10 CFR Part 20. The calculated overpack dose rates are determined in Section 5.1 of the FSAR and Appendix S of Holtec proprietary report, HI-2094431, HI-STORM FW and HI-TRAC VW Shielding Analysis, Revision 18 (Holtec, 2019b, Enclosure 2).

6.1.2 Shielding Design Features The applicant states that the shielding from the gamma radiation of the HI-STORM FW system during loading, unloading, and transfer operations is provided by the stainless-steel structure and the basket of the MPC and in the HI-TRAC-VW transfer cask. For storage, the applicant states that the gamma shielding is provided by the MPC, and the steel and concrete of the overpack. Shielding from neutron radiation is provided by the concrete of the overpack during storage and is provided by the water for the HI-TRAC-VW during transfer cask during loading, unloading, and transfer operations. To accommodate the transfer of this new canister and fuel type, the applicant added a transfer cask design that has 3.25 inches of lead instead of the 2.75 inches lead as shown in the FSAR Table 3.2.2 for the HI-TRAC transfer cask. The applicant performed the shielding analysis for the FW system containing an MPC-32ML and loaded with intact design-basis fuel 16x16D for normal conditions. The calculated dose rates for the positions shown in FSAR Figure 5.1.1 are shown in FSAR Tables 5.1.10 and 5.1.11, and for the positions shown in FSAR Figure 5.1.2 are shown in FSAR Table 5.1.12.

6.1.3 Source term Specification The design-basis source specifications for bounding calculations are presented in Section 5.2 of the FSAR. The neutron and gamma source terms, decay heat values, and quantities of radionuclides available for release were calculated with the TRITON and ORIGAMI modules of the SCALE 6.2.1 system for MPC-32ML. The applicant states that the gamma source term was comprised of three distinct sources. The first was a gamma source term from the active fuel region due to decay of fission products. The second source term was from Co60 activity of the stainless-steel structural material in the fuel element above and below the active fuel region.

The third source was from (n,) reactions. Table 5.2.1 of the FSAR provides a description of the 11

design basis fuel for the source term calculations. The determination of the MPC-32ML design basis fuel assemblies (enrichment, burnup, and cooling time combination) is discussed in detail in Section 5.2.7 of the FSAR.

The applicant calculated the source terms (neutron and gamma) for the new Westinghouse 16x16D spent fuel loaded in the MPC-32ML canister. FSAR Table 5.2.17 presents the description of 16x16D design basis fuel. Tables 5.2.18 and 5.2.19 show calculated 16X16D (MPC-32ML) PWR fuel gamma source per assembly for 45,000 MWD/MTU design basis burnup and 5-year cooling under normal conditions and 62,500 MWD/MTU design basis burnup and 8-year cooling time for accident conditions, respectively. Table 5.2.21 presents calculated Co60 source per assembly for 16X16D (MPC-32ML) at selected design basis burnup and cooling time combinations for normal and accident conditions. Table 5.2.22 presents calculated 16X16D (MPC-32ML) PWR neutron source per assembly at selected design basis burnup and cooling time combinations for normal and accident conditions.

6.1.4 Computer Codes and Modeling The applicant used the MCNP-5 code for the shielding analyses. MCNP calculations were performed for each of the three source terms (decay gamma, Co60, and neutron). The axial distribution of the fuel source term is presented in Table 2.1.5 and Figures 2.1.3 and 2.1.4 of the FSAR. Table 5.3.1 shows the axial distribution source term of the new Westinghouse 16x16D fuel assemblies in MPC-32ML. The assumptions used in the model were previously reviewed and found acceptable by the staff for HI-STORM 100 Amendment No. 5.

The applicant performed calculations for (1) dose rates adjacent to and one meter from the HI-STORM FW system with MPC-32ML, considering all loading patterns; and (2) dose rates from a single HI-STORM FW cask, and from various arrays of HI-STORM FW, for various fuel burnup, cooling time, and enrichment combinations.

6.1.5 Shielding Analyses The applicant performed shielding analyses for new Westinghouse 16x16D fuel inside MPC-32ML using the MCNP5 computer code and modeled it as three regions. As part of the shielding analysis, the applicant developed the burnup, enrichment, and cooling time combination (BECT) loading with Westinghouse 16x16D design fuel source term obtained from the SCALE 6.2.1 code for burnup from 15,000 to 70,000 MWD/MTU in increments of 5,000 MWD/MTU. A detailed description of the MCNP models and the source term calculations are presented in Sections 5.3 and 5.2, respectively.

There is only one heat load pattern with uniform loading configuration for MPC-32ML. For the fuel to be loaded into the MPC-32ML canister, the burnup and cooling time combinations for each region of the cask must satisfy the polynomial equation below:

Ct = A Bu3 + B Bu2 + C Bu + D

where, Ct = minimum cooling time in years Bu = Assembly-average burnup in MWD/MTU The polynomial coefficients are 12

A = 6.7667E-14 B = -3.6726E-09 C = 8.1319E-05 D = 2.7951E+00 The equation is based on which combinations would have the highest dose rates at a given location. The applicant used the lower bound enrichment to obtain the highest dose rate for each burnup increment in order to bound the fuel assembly inventory to be covered. For each selected combination, the maximum dose rates are presented. The coefficients are determined using the curve fitting technique.

Then the applicant calculated the cooling time for each burnup using the equation and polynomial coefficients. The cooling time is rounded down to nearest cooling time available in calculated source term library. The design basis BECT combinations are shown in the Table 5.0.3 of the FSAR for normal conditions. The maximum dose rates for the patterns in Table 5.0.3 is given in Section 5.1 of the FSAR. Table 5.4.9 and Table 5.4.10 show adjacent and 1-m dose rates for selected BECT combinations from Table 5.0.3, respectively. The dose rates for arrays of HI-STORM FWs with MPC-32ML are provided in Table 5.4.11 for the most bounding loading pattern from Table 5.0.3.

6.2 Staffs Evaluation The staff verified the applicant code input in the calculation packages and confirmed that the proper material properties and boundary conditions were used. The staff also checked the engineering drawings to verify that proper geometry dimensions were translated to the analysis model. The staff reviewed the material properties stated in the FSAR to verify that they were appropriately referenced.

The staff performed confirmatory source term evaluations for new Westinghouse 16x16D fuel in the MPC-32ML using the SCALE 6.1 computer code with the ORIGEN-ARP isotopic depletion and decay sequence with the 44-GROUP ENDF/B-VI cross section library. The staff used the same irradiation parameter assumptions used by the applicant and obtained calculated source terms that were similar to those determined by the applicant. Based on staffs independent confirmatory calculation, the staff determined the applicants shielding analysis for the addition of the new Westinghouse 16x16D fuel in the MPC-32ML in the HI-STORM FW System is adequate.

The staff independently evaluates the applicants conclusion that the polynomial equation provides the bounding conditions for the fuel that can be loaded in the MPC-32L. The points on the curve represent the fuel that give highest dose rate for BECT at that conditions. The users can only load fuels that produce dose rates lower than the dose rate on the curve for the BECT combinations. Any fuel that give higher dose rate cannot be loaded into MPC-32L. The equation was established using all discharged fuel assemblys available data for PWR and PWR nuclear power plants. The fuel enrichment should cover most actual fuel assemblies.

Based on staffs independent evaluation and the review of the calculated dose rates for MPC-32ML with the design basis 16x16D fuel for normal operations, the staff found applicants calculated dose rates acceptable. The staff has reasonable assurance that compliance with 10 CFR Part 20 and 10 CFR 72.104(a) from direct radiation can be achieved by general licensees.

The actual doses to individuals beyond the controlled area boundary depend on several site-specific conditions, such as fuel characteristics, cask-array configurations, topography, 13

demographics, and distances. In addition, 10 CFR 72.104(a) includes doses from other fuel cycle activities, such as reactor operations. Each general licensee is responsible to verify compliance with 10 CFR 72.104(a) in accordance with 10 CFR 72.212. A general licensee will also have an established radiation protection program as required by 10 CFR Part 20, Subpart B and will demonstrate compliance with dose limits to individual members of the public and workers (including excavation activities), as required, by evaluation and measurements.

The staff reviewed the accident evaluation and found it acceptable for the requested addition of MPC-32ML. The staff has reasonable assurance that the direct radiation from the FW satisfies 10 CFR 72.106(b) requirements at or beyond a controlled boundary of 100 meters from the design-basis accidents. Site specific shielding evaluation is required for fuel with a burnup more than the design basis burnup, enrichment, and cooling time combination in FSAR Table 5.0.2 for MPC-32M.

6.3 Additional Content for MPC-37 The applicant requested to add 16x16E fuel assembly as the content for MPC-37. The applicant performed the shielding analysis for the HI-STORM FW overpack containing an MPC-37 and loaded with intact design basis fuel and determined dose rates for the positions shown in Figure 5.1.1 of the SAR. The shielding analyses were performed using the MCNP5 computer code, which was used in the previous amendments. The source terms for the design basis fuels were calculated with the SAS2H and ORIGEN-S sequences from the SCALE 5 system using the design basis zircaloy clad fuel assemblies Westinghouse 17x17 for PWR fuel types, which was previously evaluated and approved by the staff (NRC, 2011). The Westinghouse 17x17 fuel bound all other PWR fuel types, including CE 16x16A, B, C, and E.

Because 16x16E fuel is bounded by the 17X17 fuel, the staff determined that the storage of 16x16E in MPC-37 will not have a significant effect on dose and are bounded by the previous authorized content of HI-STORM FW System. The acceptable fuel characteristics, including the acceptable maximum burnup levels and minimum cooling times for 16x16E fuel class has been added to SAR Table 2.1.2, PWR Fuel Assembly Characteristic. Site specific shielding evaluations are required to verify each assembly meet the technical specification for storage in the MPC-37. Based on information provided by the applicant and staffs independent analysis, the staff has reasonable assurance that the shielding design features of the HI-STORM FW system can meet the radiological requirements of 10 CFR Part 20 and 10 CFR 72.

6.4 Evaluation Findings

Based on the NRC staff's review of information provided for the HI-STORM FW application, the staff finds the following:

F6.1 The FSAR sufficiently describes shielding design features and design criteria for the structures, systems, and components important to safety.

F6.2 Radiation shielding features of the HI-STORM FW are sufficient to meet the radiation protection requirements of 10 CFR Part 20, 10 CFR 72.104, and 10 CFR 72.106.

F6.3 Operational restrictions to meet dose and ALARA requirements in 10 CFR Part 20, 10 CFR 72.104 and 72.106 are the responsibility of each general licensee. The HI-STORM FW shielding features are designed to satisfy these requirements.

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F6.4 The staff finds the design addresses construction activities involving excavation (for ISFSI expansion) adjacent to the (operating) FW system sufficient to ensure that the shielding features will continue to be sufficient to meet the radiation protection requirements of 10 CFR Part 20, 10 CFR 72.104, and 10 CFR 72.106.

F6.5 The staff concludes that the addition of MPC-32 ML to the design of the radiation protection system of the HI-STORM FW can be operated in compliance with 10 CFR Part 72 and that the applicable design and acceptance criteria have been satisfied. The evaluation of the radiation protection system design provides reasonable assurance that the HI-STORM FW will provide safe storage of spent fuel. This finding is based on a review that considered the regulation itself, the appropriate regulatory guides, applicable codes and standards, the applicants analyses, the staffs confirmatory analyses, and acceptable engineering practices.

7.0 CRITICALITY EVALUATION

The NRC staff reviewed this application for amendment together with its method for criticality safety analyses, the results and conclusions. The objective of this review is to verify that the system design as amended continues to meet the regulatory requirements of 10 CFR Part 72 with respect to criticality safety.

The staffs evaluation of the criticality safety design of the HI-STORM FW System is based on the application and supplemental information. Specifically, the staff reviewed the information in Sections 1, 2 and 6 of the HI-STORM FW FSAR and verified that the information is consistent with all descriptions, drawings, figures and provides sufficient details to support an in-depth evaluation.

The following proposed changes are applicable to the criticality evaluation:

• Proposed Change #1: Add MPC-32ML for storage in HI-STORM FW system.

• Proposed Change #2: Add fuel assembly class 16x16E as content for MPC-37.

7.1 Criticality Design Criteria and Features The HI-STORM FW system is a welded metallic MPC contained in an overpack constructed from a combination of steel and concrete. The previously approved MPCs for the HI-STORM FW include MPC-32, MPC-37, and MPC-89.

The MPC-32ML is designed to hold the 16x16D class fuel assemblies. CoC Appendix B Table 2.1-2 provides the key characteristics of the 16x16 fuel designs. The 16x16D fuel can be up to 154.5-inch long, which is 4.5 inches longer than other 16x16 fuel classes. The MPC-32ML basket design is similar to the MPC-37 basket; the only major difference is the increased cell ID to hold the slightly larger 16x16D fuel assembly and a reduced number of fuel assemblies that can be held in the basket, i.e., reducing from 37 to 32.

The HI-STORM FW system uses the previously approved HI-TRAC VW transfer cask to facilitate loading, unloading, and transfer to and from the ISFSI pad. There is no change in the design of HI-TRAC VW.

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For PWR fuels, the HI-STORM FW system relies on soluble boron in the MPC for criticality safety control during loading and unloading operations. The minimal soluble boron concentration for PWR MPCs are also prescribed in the technical specifications (TS).

The HI-STORM FW system relies on the control of content quantity, favorable MPC geometry, minimal poison (B-10) loading in the MPC poison plates, and soluble boron for criticality safety control of the MPC-32ML. The MPC consists of a steel cylinder and Metamic-HT plates to form the fuel basket. The Metamic-HT plates are made of composite material containing a specific area density of B-10, which is a measure of the minimal quantity of B-10 per square inches.

The structural stability of the Metamic-HT has been evaluated and approved by the NRC in HI-STORM FW Amendment Nos. 0 through 3. The staffs evaluation of the structural performance of the MPC-32ML is documented in Chapter 3 of this SER. Drawing 10457 provides the structural arrangement and dimensions for the MPC-32ML canister design.

The HI-STORM FW designs allows for storage of damaged fuel in the MPC-32ML canister provided that the damaged fuels are contained in DFCs. The maximum number is 8 DFCs per MPC, and each DFC can hold up to maximum one fuel assembly. For the damaged fuel, the applicant assumed that fuel has lost its structural integrity and fills the DFC as random particulates. Figure 2.1.1b of the FSAR shows the allowable locations of damaged fuel in the MPC-32ML canister. For MPC-32ML with damaged fuel, elevated soluble boron concentrations are also required at the time of loading in the spent fuel pool to assure criticality safety, i.e., the system remains subcritical. Table 2.1.6 of the FSAR shows the required soluble boron concentrations for loading the MPC-32ML canister. These requirements are also included in the revised TS for the HI-STORM FW system.

The applicant calculated the neutron multiplication factor, keff, for MPC-32ML loaded with intact 16x16D class fuel and provides the results of the calculations in Table 6.1.1(b) of the FSAR.

The maximum keff value for MPC-32ML loaded with intact 16x16D class fuel at 5.0 wt.% U-235 enrichment is 0.9427, which includes all bias and bias uncertainties and manufactural tolerances that may impact the criticality safety of the system. The required minimal soluble boron concentration for the MPC-32ML loaded with the intact 16x16D class fuel at 4.0 wt.% U-235 enrichment is 1,500 ppm (parts per million) when the cask is submerged in water. The required minimal soluble boron concentration for the MPC-32ML loaded with the intact 16x16D class fuel at 5.0 wt.% U-235 enrichment is 2,000 ppm.

As noted above, the HI-STORM FW designs allows for damaged fuel contained in DFCs to be stored in the MPC-32ML, and the applicant assumed that fuel has lost its structural integrity and fills the DFC as random particulates. The applicant performed criticality safety analyses for the MPC-32ML cask containing 8 DFCs and provided the results in FSAR Table 6.1.4(b). The maximum keff value for MPC-32ML loaded with damaged 16x16D class fuel at 5.0 wt.% U-235 enrichment is 0.9380, but the required minimum soluble boron concentration increases to 2,100 ppm. The minimal soluble boron concentration for the MPC-32ML loaded with the damaged 16x16D class fuel at 4.0 wt.% U-235 enrichment is 1,600 ppm. The applicant notes that the licensee or user of this cask design may use a linear interpolation method to determine the required soluble boron concentration for U-235 enrichment between 4.0 wt.% and 5.0 wt.%.

The staff reviewed this method of determining the soluble boron by linear interpolation and finds it acceptable. Based on its own studies, the staff finds that for the same assembly design and keff value there is a linear relationship between enrichment and soluble boron concentration when enrichment and soluble boron concentration are within certain ranges. The maximum enrichment and soluble boron of the MPC-32ML are within the ranges of the staffs studies. On 16

this basis, the staff determined that the linear interpolation scheme proposed by the applicant is acceptable.

The staff confirmed that the applicant has included appropriate soluble boron and surveillance requirements in the TS for the PWR MPC-32ML canister for loading and unloading operations.

Administrative control on the soluble boron concentration during loading and unloading of the PWR MPCs consists of independent measurements which are prescribed in the HI-STORM FW TS. An accidental loss of soluble boron is therefore determined to be not credible and hence not considered. The staff finds this acceptable because the procedures in the TS provide specific instructions for assuring reliable measurement of the soluble boron in the cask while the cask is flooded with borated water.

7.2 Fuel Specification The applicant provided specification for the new fuels to be stored in the MPC-32ML canister in FSAR Table 2.1.2. These fuel characteristics include:

• Number of fuel rods

• Maximum Fuel Pellet outer diameter (OD)

• Minimum Fuel Clad outer diameter (OD)

• Maximum Fuel Clad inner diameter (ID)

• Maximum Fuel Rod Pitch

• Maximum Active Fuel Length

• Number of Guide and/or Instrument Tubes

• Maximum Guide and/or Instrument Tube Thickness

• Maximum U-235 enrichment The staff reviewed these fuel parameters and finds that they provide the information necessary for the staff to perform a detailed criticality safety review. Since the cask design does not take credit for the net loss of fissile materials and accumulation of fission products and non-fissile actinides, i.e., burnup credit, there is no need for specification of minimum burnup for the fuels with respect to criticality safety.

7.2.1 Non-fuel Hardware As specified in CoC Appendix B, Table 2.1-1, the allowable contents to be stored together with the spent fuel in MPC-32ML include non-fuel hardware. The applicants criticality safety calculations do not include the guide tube and assumed that the guide tubes are flooded with borated water. Although the hardware displaces moderator and provides some absorption, it also displaces soluble boron. The applicant analyzed this effect on reactivity and concluded that the displacement of soluble boron versus moderator has a negative reactivity effect when the soluble boron concentration is at 2,000 ppm boron level. The staff finds that this conclusion consistent with studies performed for other canister designs and the staffs experience.

The design also allows for storage of start-up neutron sources. The applicant states that none of the neutron sources contain fissile material. Therefore, a neutron source will not increase reactivity but will increase the absolute neutron flux for a subcritical system. Although the staff finds significant presence of beryllium will impact the systems reactivity, the staff determined the beryllium presence does not impose a concern on criticality safety for this analysis because only one neutron source is allowed in a single assembly within an MPC and the quantity of 17

beryllium is limited such that it does not impact the systems reactivity, keff. On this basis, the staff finds the inclusion of neutron sources acceptable based upon the justification and other conservative assumptions used in the HI-STORM FW criticality analyses. However, this conclusion needs be evaluated on a case by case basis. Significant presence of beryllium will impact the systems reactivity.

7.2.2 Fuel Condition The HI-STORM FW system is designed to store undamaged fuel assemblies, damaged fuel assemblies, and fuel debris in MPC-32ML. However, damaged fuel must be placed in DFCs and loaded in certain locations in the spent fuel canisters as specified in the TS.

The applicant performed criticality safety analyses for the MPC-32ML with 8 damaged fuel assemblies in DFCs. The results show that the MPC-32ML canister requires 2,100 ppm soluble boron in pool, an increase of 100 ppm boron compared with the same canister loaded with all intact fuel. But the keff decreases from 0.9427 to 0.9380. This is the net effect of geometry changes of damaged fuel and increased soluble boron concentration.

The staff reviewed the applicants criticality safety analyses for MPC-32ML with the 16x16D fuel types and finds that the system design meets the acceptance criteria of NUREG-1536 for criticality safety and therefore is acceptable.

The applicant used the same method and assumptions in criticality safety analyses for the MPC-32ML canister as it used for other fuel types and canister designs that have been approved in HI-STORM FW Amendment Nos. 0 through 3. Therefore, the staff does not perform further evaluation on the method of evaluation and assumptions.

7.3 Model Specification The applicant performed criticality safety analyses for the new PWR fuel canister design, MPC-32ML containing the 16x16D fuel type. For the canister and allowable fuel contents, the applicant assumed that fuel is unirradiated (i.e., fresh fuel). This assumption is the major conservatism in the spent fuel storage cask criticality design because the permanently discharged fuel contains much less fissile materials than the fresh fuel does.

The applicant modeled the MPCs with their design specifications and crediting 90% of the B-10 in the Metamic poison plates in the criticality safety analyses for the new MPC-32ML canister design.

The applicant modeled the MPCs in the transfer cask under normal, off-normal, and accident conditions. For normal conditions, the applicant includes fabrication tolerances in the criticality safety analysis models to account for the potential most reactive configurations. The applicant includes soluble boron concentrations in the models for the MPC-32ML with all intact 16x16D fuel assembly class fuel. The soluble boron concentration included in the model is consistent with the required soluble boron concentration as specified in the TS.

In the FSAR, the applicant states that the MPC-32ML basket design is similar to the MPC-37 basket with respect to the criticality safety. The major differences are the increased basket length, increased cell ID, and rod pitch. The applicant consequently reduced the number of assemblies that can be stored from 37 to 32. Based on these comparisons, the applicant states that all studies performed for MPC-37 are directly applicable to MPC-32ML.

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In response to staffs RAI (NRC, 2018, Attachment 2), the applicant performed further studies to demonstrate the applicability of the studies performed for the MPC-37 basket to MPC-32ML and presented the results in Tables 6.3.1, 6.3.2, 6.3.5, 6.3.6, 6.4.1, 6.4.2, 6.4.4 and 6.4.5 of the FSAR. The results confirm that the MPC-32ML basket design is similar to the MPC-37 basket with respect to the criticality safety features of the canister designs.

The staff reviewed the response to the RAI and the revised FSAR and finds that the applicant demonstrated the MPC-32ML loaded with the 16x16D class fuel is similar to that of the MPC-37 canister with respect to criticality safety although the 16x16D class fuel has a longer active fuel region, larger cell ID, and rod pitch. On this basis, the staff finds that all sensitivity studies performed for the MPC-37 canister are applicable to the MPC-32ML with respect to criticality safety analyses. The major sensitivity studies include rod pitch, enrichment level, fuel to moderator ratios, temperature, internal and external moderation effects, etc., discussed in FSAR Sections 6.3.1 and 6.4.2, are applicable to MPC-32ML cask design.

For the two PWR canister designs, MPC-37 and MPC-32ML, the applicant used the same modeling approach as used in the analyses for the previously approved canister designs. Full three-dimensional models were used for all calculations. The models explicitly define the fuel rods and cladding, the guide tubes, the neutron absorber walls of the basket cells, and the surrounding MPC shell and overpack. For the flooded condition (loading and unloading), fresh water was assumed to be present in the fuel rod pellet-to-clad gaps since this represents the bounding condition as demonstrated in FSAR Section 6.4.2.3.

To ensure the configuration with optimum moderation and highest reactivity is analyzed, the applicant further studied the effect of the amount of fuel loaded per DFC on the systems reactivity. This is achieved by changing the number of rods in the array and the rod pitch. The number of rods varies between 64 (8x8) rods and 676 (26x26) rods per PWR fuel DFC. The staff reviewed the study and finds it covers reasonably large ranges of damaged fuel that may be put into a DFC. The maximum number of fuel rods, i.e., 676 (26x26), represents the maximum capacity of the damaged fuel can, i.e., the DFC cavity is fully occupied by the individual rods, whereas the DFC loaded with 64 (8x8) is the lower limit.

From Table 6.4.11 of the FSAR, the maximum keff occurs at the case of loading of 22x22 rods with 5 wt.% U-235 enrichment. Based on its own calculation, the staff finds that a DFC loaded with 64 rods is over-moderated, i.e., too little fuel and too much water. Any loading with less than 64 PWR rods in a DFC will cause the reactivity to decrease further.

Based on the fuel load limit imposed on the TS for Amendment No. 4, the maximum quantity of fuel per DFC shall not exceed a full quantity of a fuel assembly, i.e., 264 rods per a Westinghouse 17x17 PWR fuel assembly. The rest of the data, i.e., greater than 17x17 rods, as presented in Table 6.4.11 are used to demonstrate the reactivity changes as a function of number of rods per DFC.

The staff reviewed the applicants model and compared it with the licensing drawings in Chapter 1 of the FSAR and confirmed that the applicant used correct dimension important to criticality safety. The staff verified that the applicant considered the manufacturing tolerances of the basket when constructing their criticality models for the two new canister designs. Since there is no change made in the modeling strategy, the staff did not perform further evaluation on the model specification used by the applicant.

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7.3.1 Model Configuration Based on structural analyses, the canister as well as the fuel in the canister will not experience any changes in their geometries under normal and off-normal operating conditions of the ISFSI.

Therefore, the applicant used the cask and fuel geometric dimensions as designed in the criticality safety analyses for the cask under normal and off normal conditions of operations.

The staff reviewed the applicants results and conclusions of the structural review of the system design and determined that this assumption is acceptable because the cask and fuel are not expected to change their geometric forms under normal and off-normal operation conditions.

The applicant discussed the potential for changes to the basket geometric configuration due to impact under accident conditions as prescribed in 10 CFR 72.122. The applicants structural analyses show that the fuel and basket will not deform under these accident conditions. The staff finds that the applicant has addressed these impacts and that the models used to evaluate the criticality safety of the cask design are consistent with the damaged conditions of the canister as shown in structural performance analyses of the canister and fuel.

The applicant modeled the MPC-32ML canister containing 16x16D fuel at 5 wt.% U-235 enrichment in accordance to the specifications provided in FSAR Table 2.1.2. These specifications are consistent with the fuel specifications prescribed in the TS. The applicant neglects all minor structural material and replaces it with borated or fresh water, as applicable.

The applicant studied the potential positive reactivity feedback caused by displacement of borated water in the PWR canisters and demonstrated this assumption at this level of soluble boron concentration is conservative with respect to criticality safety. The staff finds these assumptions acceptable as it simplifies the safety analyses and produce conservative results.

The applicant evaluated partial flooding for both the vertical and horizontal positions for both MPC designs and found that the fully flooded condition gives the highest reactivity. The applicant studied the impact of partial flooding of the canister. The term partial flooding refers to a condition in which the cask is not filled with full density water or not fully filled with full density water. Because of the similarity between the MPC-32ML cask and the MPC-37 cask in system nuclear characteristics, the applicant determined that all previous studies on the MPC-37 cask design with respect to partial flooding is applicable to the MPC-32ML cask. The staff finds that the conclusion is consistent with the basic nuclear theory and engineering principle and applicable to all of the assembly configurations that are to be stored in the HI-STORM FW.

The applicant studied the impact of moderator density on the reactivity. The applicant presents the results in FSAR Table 6.4.5. The result shows that the maximum reactivity occurs at fuel density of water with 2,000 ppm boron in the water. The staff finds this conclusion acceptable.

The applicant considered flooding in the fuel rod pellet-to-clad gap regions with unborated water. The applicant found that it is conservative to assume that the pellet-to-clad gap regions are flooded and therefore for all cases, the pellet-to-clad gap regions are assumed flooded with unborated water. This assumption is conservative and acceptable because assuming the pellet-to-clad gap is flooded with unborated water adds more moderator to the system. FSAR Table 6.2.3 shows the keff for all assemblies that are to be stored in the HI-STORM FW and the analysis is conservative with flooded pellet-to-clad gap. This is further shown in FSAR Table 6.4.3. The staff finds this assumption to be acceptable.

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The applicant calculated the neutron multiplication factor, keff, for MPC-32MLand provides the results of the calculations in FSAR Table 6.1.1(b). For the MPC-32ML loaded with the 16x16D class fuel at 5.0 wt.% U-235 enrichment, the minimal soluble boron concentration is 2,000 ppm and the maximum keff value is 0.9427 with all bias and bias uncertainty included. The staff finds that the calculated keff value is below the acceptance criterion for criticality safety as set in NUREG-1536. On this basis, the staff finds that the HI-STORM FW design for the storage of the 16x16D class fuel in the MPC-32ML canister meets the criticality safety requirement of 10 CFR 72.236(c).

7.3.2 Material Properties The staff verified that compositions and densities are provided for all materials used in the computational model as provided in FSAR Table 6.3.4. The staff verified the material properties provided by the applicant and finds them appropriate because these material properties are consistent with the PNNL-15870, Revision 1, Compendium of Material Composition Data for Radiation Transport Modeling, and the SCALE computer code material property library which are widely used as standard compositions for nuclear calculations.

The applicant assumes 90% of the B-10 content for the Metamic-HT fixed neutron absorber.

The staff verified that the applicant performs acceptance testing and assessment of the material to ensure its continued efficacy. The staff finds this consistent with the staff position on neutron poison plate boron credit and acceptable.

The applicant performed an analysis that assumed that the absorber material received a constant neutron source equivalent to the design basis fuel as determined in FSAR Section 5.2 and found there was no significant degradation of the B-10 for 60 years. The staff finds that this meets the requirements of 10 CFR 71.124(b) and finds this acceptable.

For simplification, the applicant used the same fuel density of the regular UO2 fuel for all fuel rods, including those containing gadolinia poison. Because fuel loaded with gadolinia has a lower fuel density, this approach assumes more fuel in the assembly than it actually has in the criticality safety analysis models and therefore will result in a larger calculated keff than the system actually has because assuming a higher than the actual density of the fuel leads to a higher amount of fuel per assembly. This approach is therefore conservative and acceptable.

The staff reviewed the material properties used in the criticality safety analysis models and finds them consistent with the material compositions typically used in criticality calculations and meet the acceptance criteria recommended in NUREG-1536 for sources of material properties. On this basis, the staff finds that the applicant has used appropriate material compositions for the structural and fuel materials. The material property for the Metamic, which is a Holtec registered trademark material, has been approved for use in dry storage system construction to serve both as a structure material and well as a neutron poison plate for criticality control.

7.4 Criticality Safety Analysis 7.4.1 Computer Programs The applicant performed the criticality analyses using the MCNP5 three-dimensional Monte Carlo code and continuous energy cross sections that are distributed with the code. The applicant lists the specific cross sections used in FSAR Table 6.A.4. The MCNP5 code and the cross sections are widely used in criticality analyses for almost all types of nuclear system and 21

is extensively validated and verified by the code developer, Los Alamos National Laboratory, and nuclear industry. It is one of the computer codes identified in NUREG-1536 as appropriate for criticality safety analysis for dry cask storage system criticality safety analyses. On these bases, the staff finds this code and the selected set of cross section are appropriate for this application and therefore acceptable.

7.4.2 Multiplication Factor The applicant presented its criticality safety analysis results for the MPC-32ML and MPC-32ML loaded with 8 DFCs in Tables 6.1.1(b) and 6.1.4(b) of the FSAR, respectively. The data in the table show that the maximum keff value is 0.9427 with fresh fuel assumption and at 2,000 ppm boron when intact fuel in the cask. The maximum keff value for MPC-32ML is 0.9380 when it is loaded with 8 DFCs. However, the minimal soluble boron density required when loading the cask with 8 DFCs is 2,100 ppm. These required soluble boron densities are incorporated in the TS of the HI-STORM FW system design.

The staff reviewed the results of the criticality safety analyses performed by the applicant. The results show that the maximum keff is below 0.95, which assures that the system remains subcritical under all operating conditions, normal, off-normal, and accidents. The maximum keff value meets the acceptance criterion for criticality safety for cask under normal, off-normal, and accident conditions of operation. On this basis, the staff determined that the HI-STORM FW design for the new MPC-32ML canister loaded with 16x16D class fuel as specified in the TS meets the regulatory requirement of 10 CFR 72.236(c).

7.4.3 Confirmatory Analyses The staff performed confirmatory analysis to verify the applicants calculation for the MPC-32ML with design basis fuel and soluble boron concentration. The staff used the information in the applicants output file for the MPC-32ML canister to reconstruct a SCALE Keno VI module. The staff used SCALE 6.1 and ENDF/B-VII continuous energy cross section library in its model. The staffs result confirms that the applicants criticality safety analyses with reasonable assurance are acceptable.

7.4.4 Computer Code Benchmarking The applicant performed benchmarking analyses for the MCNP code and the selected cross section in the previous amendments which are applicable to the MPC-32ML with the 16x16D fuel. The staff did not perform further review.

7.5 Evaluation Findings

Based on the evaluation documented above, the staff has the following findings with respect to the criticality safety of the HI-STORM FW design with the requested amendment:

F7.1 Structures, systems, and components important to criticality safety are described in sufficient detail in Chapters 1, 2 and 6 of the HI-STORM FW FSAR to enable an evaluation of their effectiveness.

F7.2 The cask and its spent fuel transfer systems meets the requirements of 10 CFR 72.236(c), i.e., the system remains subcritical under all expected norm and off-normal conditions of operations as well as designed basis accident conditions.

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F7.3 The criticality design is based on favorable geometry, and fixed neutron poisons, and soluble boron poison. An appraisal of the fixed neutron poisons has shown that they will remain effective and there is no credible means for the fixed neutron poisons to significantly degrade for the design basis duration of the cask; therefore, there is no need to provide a positive means to verify their continued efficacy as required by 10 CFR 72.124(b).

F7.4 The analysis and evaluation of the criticality design and performance have demonstrated that the cask will enable the storage of spent fuel for the term as specified in the CoC.

Based on its review, the staff concludes that the criticality design features for the HI-STORM FW as amended with this application are in compliance with 10 CFR 72.124(b) and 72.236(c).

The evaluation of the criticality design provides reasonable assurance that the HI-STORM FW will allow safe storage of spent fuel. These findings are reached on the basis of a review that considered the regulation itself, applicable regulatory guides, codes and standards, and accepted engineering practices.

8.0 MATERIALS EVALUATION The staff reviewed the proposed changes in Amendment No. 4 to the HI-STORM FW storage system to verify that the applicant performed adequate materials evaluation to ensure adequate material performance of components important to safety under normal, off-normal, and accident conditions. The staffs review also ensures that the materials performance will remain within the allowable values for thermal, confinement, shielding, and criticality evaluations.

The following proposed changes are applicable to the materials evaluation:

• Proposed Change #1: Add MPC-32ML for storage in HI-STORM FW system.

• Proposed Change #2: Add fuel assembly class 16x16E as content for MPC-37.

• Proposed Change #4: Add caution note in FSAR Section 9.2.1 to avoid exposure to air during loading operations.

• Proposed Change #5: Update the definition of undamaged fuel assembly in Glossary to be consistent with the definition in Appendix A and FSAR Table 2.1.3.

• Proposed Change #6: Revise Metamic-HT Sourcebook to include fracture toughness test program, to study crack propagation in material assumed to contain flaws or defects.

• Proposed Change #7: Include a caution note in FSAR Section 9.2.3 that states low-enriched fuel must be shown to be without known or suspected grossly breached rods.

8.1 Materials Selection 8.1.1 Addition of MPC-32ML The allowed fuel types to be stored in MPC-32ML are uranium oxide undamaged fuel assemblies, damaged fuel assemblies, and fuel debris meeting the limits in FSAR Table 2.1.1 for the 16x16D array/class only. The content includes irradiated non-fuel hardware. The applicant evaluated functional requirements in 10 CFR 72.236 (e.g., thermal or structural) of the package as needed with MPC-32ML. The staff finds that materials properties of MPC-32ML are properly used in assessing functional requirements, with respect to American Society of Mechanical Engineers Boiler and Pressure Vessel Code and American Concrete Institute Codes. Thermal conductivities are not changed with the addition of MPC-32ML because 23

thermal conductivities were conservatively assumed to be constant for the entire temperature range. For the addition of MPC-32ML, the applicant assessed gap pressures for 1%, 10%, and 100 % fuel failure of non-mechanistic range of fuel failure, and the pressures are all lower than the existing fuels.

The temperature of aluminum basket shims could be higher than ~ 270°C under off-normal, accident, or some short-term operations (e.g., FSAR Table 4.5.17 for vacuum drying operations). The staff raised an issue regarding the potential degradation of mechanical strength of aluminum basket shims due to prolonged exposure to a higher temperature (Aluminum Association, Inc. 2003). The aluminum alloy used, Alloy 2219, is a precipitation hardened alloy. With a prolonged exposure in this higher temperature range, the micro-structure of precipitates may be altered in a few hours by over-aging, potentially leading to degradation of mechanical strength. This may affect the structural analysis during non-mechanistic tip-over. The applicant stated that (i) the peak temperature is local, and (ii) the vertical design has negligible loading of the basket shims. Therefore, any short-term effects of temperature rise will not affect the package structural properties. The staff finds that there is no temperature effect of aluminum shims on structure stability on the basis of the low mechanical properties needed for the negligible loading.

8.1.2 Addition of Fuel Assembly Class as Content in MPC-37 The addition of fuel assembly class 16x16E as content for MPC-37 meets all functional requirements of the package, including requirements for thermal, confinement, shielding, and criticality evaluations, as documented in SER Sections 4-7.

8.1.3 Metamic-HTTM SourceBook Updates The staff raised an issue regarding valid fracture toughness of Metamic-HTTM in assessing potential crack propagation due to non-mechanistic tip-over. During the staffs review of HI-STORM 100 Amendment No. 11, the applicant obtained the values of fracture toughness using measured Charpy impact energy and a correlation derived for steels to obtain fracture toughness from Charpy impact energy and updated the Metamic-HT Qualification SourceBook (Holtec, 2017b). The applicant did not initially provide adequate justification for using the fracture toughness values. The staff questioned whether the correlation from steel is valid for Metamic-HTTM, considering available literature values of fracture toughness for aluminum metal matrix composites with ceramic particles, an analog for Metamic-HTTM. During the staffs review of HI-STORM FW Amendment No. 4, the applicant measured Metamic-HTTM fracture toughness values and used the measured values in the latest version of Metamic-HT Qualification SourceBook in Amendment No. 4. The staff reviewed the applicants test data and evaluation and finds that the applicant used appropriate fracture toughness values for Metamic-HTTM in the structural analysis.

8.2 Chemical and Galvanic Reactions 8.2.1 Potential Fuel (Pellet) Oxidation During loading of the casks, the cask will be filled only with inert gas after vacuum drying. In FSAR Chapter 9, Operating Procedures, the loading and unloading procedures state that the fuel will be either covered in water or exposed to an inert gas. During water removal and vacuum drying, the applicant stated that the Helium Backfill System will be purged to remove oxygen from the line. During unloading, with the re-flooding of the cask, air would potentially be 24

removed. The staff concludes that, with over 30 years of operating experience, the applicant has presented adequate protection against fuel oxidation during operations.

Consistent with SFST-ISG-22, the applicant assessed the potential oxidation in a transportation cask (HI-STAR 80, CoC No. 9374) based on (i) appropriate literature information and (ii) example calculations for the time restriction at various temperatures. The applicant determined that the oxidation process is highly sensitive to the temperature. Thus, the time limit for exposure to air/moisture could be higher for casks with lower heat load. The applicant also considered later stage of vacuum drying with less oxygen partial pressure which is likely to decrease oxidation rate. The staff finds this assessment to be acceptable based on the staffs previous research on this topic which is documented in NUREG-1565, Dry Oxidation and Fracture of LWR Spent Fuels (NRC, 1996). The staff requested that the applicant specify either (i) the fuel is always covered with inert gas or (ii) the time (and/or temperature) limit of oxygen exposure, with a caution note to the user that the fuel oxidation shall be prevented. As the result, the applicant added caution notes in Chapter 9 that inert gas must be used when the fuel is not covered with water to prevent oxidation. With the condition imposed by NRC to add helium backfill limits for MPC-32ML, the staff determined that handling of the potential fuel oxidation is acceptable.

8.2.2 Non-Fuel Contents The staff questioned whether non-fuel contents are potentially pyrophoric and compatible with each other. The applicant confirmed that BPRAs (Burnable Poison Matrix Materials), TPDs (Thimble Plug Devices), WABAs (Wet Annular Burnable Assemblies), CRAs (Control Rod Assemblies), RCCSs (Rod Cluster Control Assemblies), CEAs (Control Element Assemblies),

NSAs (Neutron Source Assemblies), and/or APSRs (Axial Power Shaping Rods) are non-pyrophoric and compatible. Therefore, the staff finds non-fuel contents acceptable to be included because these materials are mainly stainless steels that are not subject to pyrophoric reactions in their bulk form in the non-fuel contents.

8.3 Spent Fuel Cladding Integrity 8.3.1 High Burnup (HBU) Integral Fuel Burnable Absorber PWR Fuel As documented in draft report NUREG-2224, Dry Storage and Transportation of High Burnup Spent Nuclear Fuel, (NRC, 2018), NRC has accepted the use of HBU integral fuel burnable absorber PWR fuels for storage in the first 20 years. Potential hydride reorientation in cladding may occur due to high end-of-life gas pressure and accompanying increase in cladding hoop stress. Based on the study in draft report NUREG-2224, the staff determined that potential hydride reorientation and accompanying mechanical property changes of cladding are not significant to the cladding safety requirements in storage and transportation.

8.3.2 Copper or Crud-Induced Localized Corrosion (CILC) of Cladding Fuel The MPCs may be loaded with CILC fuels, which should be undamaged. The applicant has added a similar note in HI-STORM 100 Amendment No. 11 to confirm that CILC is undamaged prior to loading. Therefore, the evaluation of HI-STORM 100 Amendment No. 8, Revision No. 1 (NRC, 2016), continues to be acceptable to the staff and an evaluation is not required.

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8.3.3 HBU Fuel Cladding Temperature Limit With the addition of MPC-32ML, the temperature of HBU fuel cladding is below the maximum temperature of 400°C in SFST-ISG-11, Revision 3. Therefore, the previous evaluations in HI-STORM FW Amendment Nos. 0 through 3 continue to be acceptable to the staff and a new evaluation is not required.

8.3.4 Fuel Assembly The applicant updated the definition of undamaged fuel assembly in Glossary to be consistent with the definition in Appendix A and FSAR Table 2.1.3. The applicant defined undamaged fuel assembly as a BWR fuel assembly with an intact channel, a maximum planar average initial 3.3 wt%, without known or suspected grossly breached spent fuel rods, and which can be handled by normal means. The applicant also included a caution note in FSAR Section 9.2.3 that states low-enriched fuel must be shown to be without known or suspected grossly breached rods. The staff determines the definition and caution note acceptable as they are consistent with the provision of SFST-ISG-1, Revision 2.

8.4 Evaluation Findings

F8.1. The staff has reasonable assurance that the SSCs important to safety for the MPC-32ML and are described in sufficient detail in the FSAR. The materials properties comply with American Society of Mechanical Engineers Boiler and Pressure Vessel Code and American Concrete Institute Codes. The staff has reasonable assurance that adding fuel assembly class 16x16E as content in MPC-37 is described in sufficient details in the FSAR. No galvanic effects are expected.

F8.2. The applicant has met the requirements of 10 CFR 72.122(h)(1) and 236 (h) for the applicable changes to materials evaluation. The design of the dry cask storage system and selection of materials adequately protects the spent fuel cladding against degradation that might otherwise lead to gross rupture of the cladding. The temperature limits are met. The gas pressure generated from integral fuel burnable absorber fuels is within the limit. The CILC fuels will be confirmed as undamaged fuel with a users note.

The undamaged fuel is appropriately defined, consistent with SFST-ISG-1, Revision 2.

F8.3. The applicant has met the requirements of 10 CFR 72.236(h) and 236(m) for the applicable changes to materials evaluation. The material of construction for SSCs important to safety will be maintained during normal, off-normal, and accident conditions of operation so the spent fuel can be readily retrieved without posing operational safety problems.

F8.4. The applicant has met the requirements of 10 CFR 72.236 (g) for the applicable changes to materials evaluation. The materials of construction for SCCs important to safety will be maintained during all conditions of operation so the spent fuel can be stored for a minimum of 20 years and maintenance can be conducted as required.

F8.5. The applicant has met the requirements of 10 CFR 72.236 (h) for the applicable changes to materials evaluation. The HI-STORM FW Storage System employs materials compatible with wet and dry spent fuel loading and unloading operations and facilities.

These materials should not degrade over time or react with one another during any conditions of storage. The staff has reasonable assurance that the applicants new 26

evaluations are adequate for Metamic-HTTM fracture, aluminum shims strength, and potential fuel oxidation.

9.0 OPERATING PROCEDURES EVALUATION The staff reviewed the proposed changes to the operating procedures to ensure that the applicant's FSAR presents acceptable operating sequences, guidance, and generic procedures for the key operations. The staff also ensures that the FSAR incorporates and is compatible with the applicable operating control limits in the TS.

The applicant proposed to add a few caution notes to the operating procedures. The following proposed changes are applicable to the operating procedure evaluation:

• Proposed Change #4: Add a caution note in FSAR Section 9.2.1 that states fuel cladding is not exposed to air during loading operations.

• Proposed Change #7: Include a caution note in FSAR Section 9.2.3 that states low-enriched fuel must be shown to be without known or suspected grossly breached rods.

The staff documents the evaluation of Proposed Changes #4 and #7 in Sections 8.2.1 and 8.3.4 of this SER, respectively, and determines these caution notes are adequate.

10.0 ACCEPTANCE TESTS AND MAINTANANCE PROGRAM EVALUATION The applicant did not propose any changes that affect the staffs acceptance tests and maintenance program evaluation provided in the previous SERs for CoC No. 1032, Amendment Nos. 0 through 3. Therefore, the staff determined that a new evaluation was not required.

11.0 RADIATION PROTECTION EVALUATION The staff reviewed proposed changes in Amendment No. 4 to evaluate the impact on the radiation protection design features, design criteria, and operating procedures of the HI-STORM FW system to ensure that it will continue to meet the regulatory dose requirements of 10 CFR Part 20, 10 CFR 72.104(a), 10 CFR 72.106(b), 10 CFR 72.212(b), and 10 CFR 72.236(d). This proposed amendment was also reviewed to determine whether the HI-STORM FW system continues to fulfill the acceptance criteria listed in Section 11 of NUREG-1536. The staffs review is based on the application to Amendment No. 4 to the HI-STORM FW system and supplemental information.

11.1 Radiation Protection Design Criteria and Features The applicable radiological protection design criteria are the limits and requirements of 10 CFR Part 20, 10 CFR 72.104, and 10 CFR 72.106. As required by 10 CFR Part 20 and 10 CFR 72.212, each general licensee is responsible for demonstrating site-specific compliance with these requirements. The HI-STORM FW system TS also establishes dose rates limits for the transfer cask and overpack that are based on calculated dose rate values used to ensure occupational and off-site radiological exposures will meet regulatory limits.

There are no proposed changes to design features for the HI-STORM FW system.

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11.2 Occupational Exposures Previous amendments discussed general operating procedures that general licensees will use for fuel loading, dry storage cask/transfer cask operations and transfer into the overpack, and fuel unloading. For the proposed Amendment No. 4, the applicant proposed to add MPC-32ML and a new fuel type 16X16D as authorized content, which have different fuel rod dimensions and a different layout of the guide tubes from the previously approved contents, and to add fuel assembly class 16X16E as authorized content of MPC-37.

The shielding analysis performed for the proposed changes is evaluated in Section 6 of this SER. Dose rate limits at the HI-TRAC VW and HI-STORM overpack surfaces are provided in the FSAR. These rates were calculated to ensure that the HI-STORM FW system is operated within design basis conditions and those ALARA goals will be met. Worker dose rate monitoring, in conjunction with trained personnel and well-planned activities, will significantly reduce the overall dose received by the workers.

CoC Appendix B, Section 5.3.4.2 states the measured dose rates on a loaded OVERPACK or TRANSFER CASK shall not exceed the following values for MPC-32L:

a. 6 mrem/hr (gamma + neutron) on the top of the overpack

b. 250 mrem/hr (gamma + neutron) on the side of the overpack

c. 2,500 mrem/hr (gamma + neutron) on the side of the transfer cask Table 5.1.12, Maximum Dose Rates from the HI-TRAC VW for Normal Conditions, MPC-32ML with 16x16D Fuel, Loading Patterns (See Table 5.0.3), shows that the maximum calculated dose rate at the surface of the side of the transfer cask is 1,666 mrem/hr. Table 5.1.10, Maximum Dose Rates Adjacent to HI-STORM FW Overpack for Normal Conditions, MPC-32ML with 16x16D Fuel, shows that the maximum calculated dose rate for the overpack at the center of the lid is 2 mrem/hr and at the side is 173 mrem/hr. These dose rates are lower than the TS limits of 2,500 mrem/hr, 6 mrem/hr, and 250 mrem/hr, respectively, making these TS limits non-conservative for the MPC-32ML. A loaded transfer cask and/or overpack with maximum allowable measured TS dose rates would exceed the calculated design basis dose rate. However, the staff found these TS limits acceptable because the MPC-32ML is not the design basis canister and the lower limit is still conservative compared to CoC Appendix A, Section 5.3.4.1, the design basis limits of 3,500 mrem/hr, 30 mrem/hr, and 300 mrem/hr for the HI-STORM FW with the MPC-89 and MPC-37 canisters which have been demonstrated in Amendment Nos. 0 through 3 to meet the regulatory dose limits in 10 CFR 72.104 and 72.106.

The applicant states that the HI-TRAC VW transfer cask provides shielding to maintain occupational exposures ALARA in accordance with 10 CFR Part 20. The calculated dose rates for HI-TRAC VW for a set of reference conditions are reported in FSAR Tables 5.1.10 and 5.1.12. These dose rates were used to perform a generic occupational exposure estimate for MPC-32ML loading, closure, and transfer operations. The annular area between the MPC outer surface and the HI-TRAC VW inner surface can be isolated to minimize the potential for surface contamination of the MPC by spent fuel pool water during wet loading operations as stated in the FSAR. The HI-TRAC VW surfaces expected to require decontamination are coated with a suitable coating. The maximum permissible surface contamination for the HI-TRAC VW is in accordance with plant-specific procedures and ALARA requirements.

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11.3 Offsite Dose Calculation The applicant estimated offsite direct radiation dose rates at the site boundary for the HI-STORM FW system overpack loaded with 16x16D fuels. Based on results presented in Table S3-13 in Appendix S, the analyses indicated that the HI-STORM FW system (assuming design-basis fuel and 100% occupancy for 365 days) can meet the annual dose limit of 25 mrem at 400 meters.

The dose received by a person at 100 meters from the ISFSI assuming a duration of 100%

occupancy for 365 days and a distance from 100 to 600 meters is shown in Table S3-12 of Appendix S. The annual dose rates for a single cask and an array of 2x2 at 400 meters and arrays of 2x3, 2x4, and 2x5 at 500 meters is shown in Table S3-13 of Appendix S. All dose rates meet the 10 CFR Part 72.104 and 106.

The general licensee using the HI-STORM FW system must perform a site-specific evaluation, as required by 10 CFR 72.212(b), to demonstrate compliance with 10 CFR 72.104(a). The actual doses to an individual beyond the controlled area boundary depend on several site-specific conditions, such as fuel characteristics, cask-array configurations, topography, demographics, distances, and use of engineered features. In addition, the dose limits in 10 CFR 72.104(a) include doses from other fuel cycle activities such as reactor operations.

Consequently, final determination of compliance with 10 CFR 72.104(a) is the responsibility of each general licensee.

The general licensee should have an established radiation protection program as required by 10 CFR Part 20, Subpart B, and will demonstrate compliance with dose limits to individual members of the public, as required in 10 CFR Part 20, Subpart D, by evaluations and measurements.

Staff reviewed the proposed limits and measurements with consideration of these parameters and finds that they are acceptable.

11.3.1 Confirmatory Calculation The staff evaluated the public dose estimates during normal and off-normal conditions. The primary dose pathway to individuals beyond the controlled area is from direct radiation (including sky shine). A discussion of the staffs evaluation and confirmatory analysis of the shielding calculations are presented in Section 6 of the SER. The staff has reasonable assurance that the effects of direct radiation from bounding design basis accidents and natural phenomena will be below the regulatory limits in 10 CFR 72.106(b).

11.4 Evaluation Findings F11.1 The FSAR sufficiently describes the radiation protection design bases and design criteria for the SSCs important to safety.

F11.2 Radiation shielding, and confinement features are sufficient to meet the radiation protection requirements of 10 CFR Part 20, 10 CFR 72.104, and 10 CFR 72.106.

F11.3 The FSAR adequately evaluates the HI-STORM FW system with MPC-ML32 and its content 16X16D fuels and their systems important to safety to demonstrate that they will 29

reasonably maintain confinement of radioactive material under normal, off-normal, and accident conditions.

F11.4 The FSAR sufficiently describes the means for controlling and limiting occupational exposures within the dose and ALARA requirements of 10 CFR Part 20.

F11.5 Operational restrictions necessary to meet dose and ALARA requirements in 10 CFR Part 20, 10 CFR 72.104, and 10 CFR 72.106 are the responsibility of the site licensee.

The HI-STORM FW are designed to assist in meeting these requirements.

12.0 ACCIDENT ANALYSES EVALUATION The applicant did not propose any changes to the principal design criteria related to the SSCs important to safety. For this reason, the staff finds the applicant complied with the relevant general criteria established in 10 CFR Part 72 and does not require an accident analysis evaluation of the principal design criteria.

13.0 TECHNICAL SPECIFICATIONS AND OPERATING CONTROL AND LIMITS EVALUATION The staff reviewed the proposed amendment to determine that applicable changes made to the conditions in the CoC, and to the TSs for CoC No. 1032, Amendment No. 4 would be in accordance with the requirements of 10 CFR Part 72. The staff reviewed the proposed changes to the technical specifications to confirm the changes were properly evaluated and supported in the applicants revised safety analysis report.

The applicants proposed changes to the certificate of compliance and technical specifications are as follows:

Table 13 Conforming Changes to the Certificate of Compliance, Technical Specifications and Operating Control and Limits Page TS Proposed Description Number Reference Change CoC Issued To Update address. N/A Page 1 CoC Add MPC-32ML to the MPC models for Description 1 Page 2 HI-STORM FW storage system.

SR 3.1.2 Specify surveillance requirement for MPC-37 and Appendix A (Surveillance MPC-89 and add surveillance requirement for 1 3.1.2-2 Requirement) MPC-32ML.

LCO 3.3.1 Add soluble boron concentration limits for Appendix A (Limiting MPC-32ML and specify the 16x16 fuel types (A, 1, 2 3.3.1-1 Condition for B, C, and E) for MPC-37.

Operation)

Appendix A Add MPC cavity drying limits for different burnups Table 3-1 1 3.4-1 for MPC-32ML.

Appendix A Table 3-2 Add helium backfill limits for MPC-32ML. 1 3.4-3 30

Table 13 Conforming Changes to the Certificate of Compliance, Technical Specifications and Operating Control and Limits Appendix A Section 5.3.4 Update dose rate limit for all approved MPCs. 1 5.0-3 Table of Appendix B Updated. N/A Content Appendix B Add reference to Figure 2.1-3 for MPC-32ML cell Section 2.1.2 1 2-1 identification information.

Appendix B Add Figure 2.1-3 for MPC-32 cell identification Figure 2.1-3 1 2-4 information.

Appendix B Table 2.1-1, 2-9 through Add fuel assembly limits for MPC-32ML. 1 Section III 2-10 Appendix B Add fuel types 16x16D and 16x16E to PWR fuel 2-13 through Table 2.1-2 assembly characteristics. Update Note 3 and 1, 2 2-14 add Note 5 for MPC-32ML.

Appendix B Add reference to Table 2.3-5 for fuel loading Section 2.3.1 1 2-20 decay heat limit information for MPC-32ML.

Appendix B Table 2.3-5 Add Table 2.3-5 for MPC-32ML heat load data. 1 2-22 Appendix B Section 2.5 Add Section 2.5 and Table 2.5-1 to include fuel 1

2-26 Table 2.5-1 qualification information for MPC-32ML.

Appendix B Add Section 3.2.4 to include MPC-32ML design Section 3.2.4 1 3-2 features important for criticality control.

The staff finds that the proposed changes to the CoC and TS for the HI-STORM FW MPC Storage System conform to the changes requested in the amendment application and do not affect the ability of the cask system to meet the requirements of 10 CFR Part 72. The proposed changes provide reasonable assurance that the HI-STORM FW MPC Storage System will continue to allow safe storage of spent nuclear fuel.

14.0 QUALITY ASSURANCE EVALUATION The applicant did not propose any changes that affect the staffs quality assurance evaluation provided in the previous SERs for CoC No. 1032, Amendment Nos. 0 through 3. Therefore, the staff determined that a new evaluation was not required.

15.0 CONCLUSION

S The staff has performed a comprehensive review of the amendment application, during which the following requested changes to the HI-STORM FW MPC Storage System were considered:

1. Add MPC-32ML for storage in HI-STORM FW system and allow fuel assembly class 16x16D as content for MPC-32ML.

2. Add fuel assembly class 16x16E as content for MPC-37.

3. Separate the design pressure for the short-term operation from the off-normal condition to provide clarity in final safety analysis report (FSAR) Table 2.2.1.

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4. Add a caution note in FSAR Section 9.2.1 that states fuel cladding is not exposed to air during loading operations.

5. Update the definition of undamaged fuel assembly in FSAR Glossary to be aligned with the definition in Appendix A and FSAR Table 2.1.3 (Note 14).

6. Replace Charpy test program with fracture toughness test program from the revised Metamic-HT Sourcebook (Holtec, 2017b) in FSAR Sections 1.2.1.4.1 and 3.4.

7. Add a caution note in FSAR Section 9.2.3 that states low-enriched fuel must be shown to be without known or suspected grossly breached rods.

Based on the statements and representations provided by the applicant in its amendment application, as supplemented, the staff concludes that the changes described above to the HI-STORM FW MPC Storage System do not affect the ability of the cask system to meet the requirements of 10 CFR Part 72. Amendment No. 4 for the HI-STORM FW MPC Storage System should be approved.

Issued with Certificate of Compliance No. 1032, Amendment No. 4 On June 15, 2020.

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