ML25155A465

TMI-2 ISFSI Biennial Update Report 2025_Redacted

ML25155A465
Person / Time
Site:	07200020
Issue date:	03/03/2025
From:	Idaho Environmental Coalition NRC Licensed Facilities
To:	Office of Nuclear Material Safety and Safeguards
References
NLF-RPT-309
Download: ML25155A465 (1)

Text

TMI-2 ISFSI Biennial Update Report for 2025 Prepared by Idaho Environmental Coalition NRC Licensed Facilities Report: NLF-RPT-309 March 2025

NLF-RPT-309, TMI-2 ISFSI Biennial Update Report for 2025 Page 1 of 4

1.0 Purpose and Scope

The Department of Energy's Idaho Cleanup Project (DOE ICP) was issued license SNM-2508 to operate the Three Mile Island Unit 2 (TMI-2) Independent Spent Fuel Storage Installation (ISFSI) by the Nuclear Regulatory Commission (NRC). The date of issuance was March 19, 1999. A renewed license was issued on September 16, 2019. This update report provides the following reports, which were last updated in March 2023.

• The biennial Safety Analysis Report (SAR) Update Report pursuant to 10 CFR 72.70 (b) and (c). The SAR Update Report is provided in Section 2.0, Description of Changes to the Safety Analysis Report.

• The biennial 72.48 Evaluations Report pursuant to 10 CFR 72.48(d)(2). This report, comprised of summaries of evaluations of changes made pursuant to 10 CFR 72.48, is provided in Section 3.0, Changes, Tests, and Experiments.

• The biennial Technical Specifications (TS) Bases Evaluations Report pursuant to TS 5.5.1.d. This report is provided in Section 4.0, Changes to the Technical Specification Bases.

• The biennial Essential Program Evaluations Report pursuant to TS 5.5.2.6. This report covers changes made during this reporting period to the DOE ICP essential programs as described in Section 5.0, Radiological Environmental Monitoring Program Changes, Section 6.0, Training Program Changes, and Section 7.0, Quality Assurance Program Changes.

NOTE: The changes to the Physical Protection Program and the Emergency Response Program are provided separately because of the reporting time frames required by 10 CFR 72.44 (e) and (f), respectively.

This report is provided in a combined format because many of the changes described in the SAR Update are also covered by reviews of changes made without NRC approval pursuant to 10 CFR 72.48, TS 5.5.1, and TS 5.5.2. TS 5.5.1 requires 72.48 reviews to be performed for any change to the TS Bases. TS 5.5.2 requires an evaluation of the change in program effectiveness [similar to the requirements of 10 CFR 72.44(e) and (f)] for changes to the ISFSI Radiological Environmental Monitoring Program, Training Program, and Quality Assurance Program. These three programs are contained in the SAR.

NLF-RPT-309, TMI-2 ISFSI Biennial Update Report for 2025 Page 2 of 4 2.0 Description of Changes to the Safety Analysis Report The previous update of the TMI-2 SAR made pursuant to 10 CFR 72.70 was provided in March 2023 (Revision 13).

Since that time, in March of 2023 changes were made to the TMI-2 SAR Chapter 7 and Chapter 8 as described below.

• Both Chapter 7 and Chapter 8 were updated to incorporate changes required by revisions to Engineering Design File (EDF)-4728, Radiological Evaluation of TMI-2 ISFSI Technical Specification 3.1.1. EDF-4728 was revised in 2021 and 2022 to correct a calculational error and include clarifications and further explanation of the calculations. While the changes in the revised EDF did not modify the conclusions reached, applicable portions of the SAR were updated in March 2023 to include expanded description of the material characterization and to update calculated values as a result of EDF revisions. The revised calculations showed that the combined gaseous and particulate exposures at the ISFSI and Idaho National Laboratory boundaries maintain a conservative margin for normal, off-normal, and accident conditions.

Attached to this report are the replacement pages for SAR chapters (Chapter 7 and

8) that were revised as required by 10 CFR 72.70.

3.0 Changes, Tests, and Experiments The previous report of changes made pursuant to 10 CFR 72.48 was provided in March 2023.

Fourteen screens were prepared pursuant to 10 CFR 72.48 during the subsequent 24-month period as described below. All changes were screened out, and no further 10 CFR 72.48 evaluations were required. Therefore, the changes were made without a license amendment.

1. In March 2023, a screen was completed documenting changes made to Chapters 7 and 8 of the SAR as described in the previous section.

2. In June 2023, a screen was completed for a work order to apply crack sealant to the asphalt pad where applicable surrounding the TMI-2 base mat.

3. In July 2023, a screen was completed in support of a Field Design Change (FDC) to re-locate approximately 30 feet of conduit to a newly trenched underground location and re-adjust the installed jersey barrier to allow personal access into CPP-1774 at the ground level, thus alleviating the need to maintain a temporary stair set.

NLF-RPT-309, TMI-2 ISFSI Biennial Update Report for 2025 Page 3 of 4

4. In July 2023, a screen was completed for a work order and FDC to replace the asphalt underneath the Delta barrier at the TMI-2 vehicle entrance with a concrete pad to increase the serviceable lifespan of the barrier and allow equipment to cross the barrier more easily.

5. In August 2023, a screen was completed for a work order/corrective action plan to remedy issues identified on non-conformance reports (NCRs) written on the Horizontal Storage Modules (HSMs). The work order defined the methods of repair for concrete and coating issues identified during aging management inspections including the FDCs completed to document the repairs that were made to the end shield wall, the concrete portions of the HSMs, and the painting that was applied to the metal door portions of the HSMs.

6. In October 2023, a screen was completed for the HSM Repair Work Order and two updated FDCs to address the use of approved grout products for the concrete repairs.

7. In April 2024, a screen was completed for a work order to repair the remaining 21 roof pockets in the TMI-2 HSMs that were not completed in 2023.

8. In April 2024, a screen was completed for a work order to repair/replace a junction box for a TMI-2 security electrical panel.

9. In July 2024, a screen was completed for a work order to apply an epoxy sealer/crack filler on the concrete TMI-2 basemat.

10. In July 2024, a screen was completed for a work order to apply a fresh coat of silane sealant on the concrete exterior of all 30 HSMs.

11. In September 2024, a screen was completed for a work order in support of a security camera upgrade/addition.

12. In October 2024, a screen was completed in support of an NCR disposition for the TMI-2 Dry Shielded Canister steel coatings.

13. In October 2024, a screen was completed in support of an NCR disposition for the TMI-2 HSM concrete and steel coatings.

14. In October 2024, a screen was completed in support of an NCR disposition for the TMI-2 HSM roof bolt.

NLF-RPT-309, TMI-2 ISFSI Biennial Update Report for 2025 Page 4 of 4 4.0 Changes to the Technical Specification Bases The previous update of the TS Bases made pursuant to TS 5.5.1 was provided in March 2023. There was no change made to the TS Bases during the subsequent 24-month period.

5.0 Radiological Environmental Monitoring Program Changes The previous update of the Radiological Environmental Monitoring Program made pursuant to TS 5.5.2 was provided in March 2023. There was no change made to the Radiological Environmental Monitoring Program described in the SAR during the subsequent 24-month period.

6.0 Training Program Changes The previous update of the Training Program made pursuant to TS 5.5.2 was provided in March 2023. There were no changes made to the Training Program described in the SAR, Section 9.3 during the subsequent 24-month period.

7.0 Quality Assurance Program Changes The previous update of the Quality Assurance Program made pursuant to TS 5.5.2 was provided in March 2023. There were no changes made to the Quality Assurance Program described in Chapter 11 of the SAR during the subsequent 24-month period.

8.0 Attachments Attachment A, TMI-2 SAR-II-8.4 Chapter 7 Attachment B, TMI-2 SAR-II-8.4 Chapter 8

NLF-RPT-309, TMI-2 ISFSI Biennial Update Report for 2025 Attachment A TMI-2 SAR-II-8.4 Chapter 7 (PDF attached)

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SAR-II-8.4 CH 7 Rev. 7 7-1 of 7-35 CHAPTER 7 RADIATION PROTECTION

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SAR-II-8.4 CH 7 Rev. 7 7-2 of 7-35 CONTENTS

7.

RADIATION PROTECTION.......................................................................................................... 7-4 7.1 Ensuring That Occupational Radiation Exposures Are As-Low-As-Reasonably-Achievable............................................................................................................................ 7-4 7.1.1 Policy Considerations.............................................................................................. 7-4 7.1.2 Design Considerations............................................................................................. 7-5 7.1.3 Operational Considerations..................................................................................... 7-5 7.2 Radiation Sources................................................................................................................. 7-6 7.2.1 Characterization of Sources..................................................................................... 7-6 7.2.2 Airborne Radioactive Material Sources................................................................... 7-7 7.3 Radiation Protection Design Features................................................................................. 7-14 7.3.1 Installation Design Features.................................................................................. 7-14 7.3.2 Shielding................................................................................................................ 7-14 7.3.3 Ventilation............................................................................................................. 7-16 7.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation............. 7-16 7.4 Estimated On-Site Collective Dose Assessment................................................................. 7-22 7.4.1 Operational Dose Assessment............................................................................... 7-22 7.4.2 Site Dose Assessment............................................................................................ 7-22 7.5 Health Physics Program...................................................................................................... 7-27 7.5.1 Organization.......................................................................................................... 7-27 7.5.2 Equipment, Instrumentation, and Facilities........................................................... 7-27 7.5.3 Procedures............................................................................................................. 7-27 7.6 Estimated Off-Site Collective Dose Assessment................................................................ 7-29 7.6.1 Effluent and Environmental Monitoring Program................................................. 7-29 7.6.2 Analysis of Multiple Contribution......................................................................... 7-30 7.6.3 Estimated Dose Equivalents to the MEI................................................................ 7-30 7.6.4 Estimated Dose Equivalents to the Occupational Worker..................................... 7-31 7.6.5 Liquid Release....................................................................................................... 7-31 7.7 References........................................................................................................................... 7-34

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SAR-II-8.4 CH 7 Rev. 7 7-3 of 7-35 FIGURES 7.3-1.

DSC and HSM Shielding Geometry........................................................................... 7-18 7.3-2.

DSC and Cask Shielding Geometry........................................................................... 7-19 7.3-3.

DSC Ventilation Port Geometry................................................................................. 7-20 7.3-4.

HSM Shielding Analysis Results................................................................................ 7-21 7.4-1.

ISFSI Geometry for Site Dose Calculations............................................................... 7-25 7.4-2.

Dose Rate Versus Distance from the ISFSI................................................................ 7-26 TABLES 7.2-1.

Neutron Energy Spectrum and Flux-to-Dose Conversion Factors for PWR Spent Fuel......................................................................................................... 7-11 7.2-2.

Gamma Energy Spectrum and Flux-to-Dose Conversion Factors for PWR Fuel...... 7-12 7.3-1.

Shielding Analysis Results......................................................................................... 7-17 7.4-1.

NUHOMS System Operations Enveloping Times for Loading One DSC.............. 7-24 7.4-2.

Dose Rates in the Vicinity of the TMI-2 ISFSI at INL.............................................. 7-25 7.6-1.

Normal Operation Estimated Effluent Dose Equivalents........................................... 7-32 7.6-2.

Off-normal Condition Estimated Effluent Dose Equivalents..................................... 7-32 7.6-3.

Accident Condition Estimated Effluent Dose Equivalents......................................... 7-33

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SAR-II-8.4 CH 7 Rev. 7 7-4 of 7-35

7.

RADIATION PROTECTION This chapter presents the radiation protection features of the NUHOMS-12T system using the NRC certified MP-187 for transportation from TAN to INTEC. Appendix E of this SAR presents the radiation protection features of the NUHOMS-12T system using the NRC 10 CFR 72 approved OS-197 Transfer Cask for transportation from TAN to INTEC.

7.1 Ensuring That Occupational Radiation Exposures Are As-Low-As-Reasonably-Achievable 7.1.1 Policy Considerations It is the policy of the DOE Idaho Operations Office (DOE-ID) to take every precaution to control radiation and the spread of radioactive contamination in the performance of work, to be in full compliance with the requirements established by the NRC, to prevent unnecessary radiation exposure to employees and the public, and to prevent harmful effects to the environment. Radiological operations at the ISFSI will be conducted in a manner consistent with those at the Idaho National Laboratory (INL). Radiation exposures to workers and the public and releases of radioactivity to the environment are maintained below regulatory limits, and deliberate efforts are taken to further reduce exposures and releases to As-Low-As-Reasonably-Achievable (ALARA) levels.

To comply with this policy, all levels of line management are accountable for radiological performance. The responsibility for compliance with the radiological protection requirements and for minimizing personnel radiation exposure begins at the worker level and broadens as it progresses upward through the line organization. Line managers are responsible for taking all necessary actions to ensure that requirements are implemented and that performance is monitored and corrected as necessary.

Radiological Control Technicians (RCTs) assist line management by routinely evaluating and monitoring all radiological conditions. Also, RCTs oversee activities to ensure that all reasonable precautions are taken by personnel.

The requirements for the ALARA policy and program are provided by 10 CFR Part 20, Standards for Protection Against Radiation [7.1]. DOE-ID is committed to reducing safety and health risks associated with hazardous substances (including ionizing radiation) by promoting ALARA policy awareness and reducing and keeping radiation exposures to ALARA levels. The following methods are used to achieve ALARA objectives.

A.

Establishing employee and organizational level ALARA goals, tracking employee exposure, and maintaining associated records.

B.

Allocating the appropriate technical, administrative, and supervisory resources.

C.

Appointing an ALARA committee to oversee and evaluate efforts, and to provide technical assistance for identifying needed improvements. The TMI-2 ISFSI Facility Safety Officer acts as the ISFSI representative on the INTEC ALARA Committee. Radiation safety issues that arise are brought to the ALARA Committee by the Facility Safety Officer.

D.

Controlling access to radiation and radioactive contamination areas.

E.

Minimizing the working time required in high radiation areas and high surface contamination areas, as appropriate.

F.

Using engineered controls (e.g., ventilation, remote handling, and shielding) and monitoring equipment (e.g., continuous air monitors and remote area monitors).

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Requiring an ALARA review of procedures and work packages for the TMI-2 ISFSI that involve radiological work resulting in individual and/or collective radiation exposure exceeding thresholds established by the ALARA Committee or requiring entry into 1 rem/h radiation fields.

7.1.2 Design Considerations The NUHOMS-12T design installed at INL incorporates design features and improvements from previously constructed NUHOMS installations at H.B. Robinson, Oconee, Calvert Cliffs, and Davis-Besse. The NUHOMS design ensures ALARA by minimizing required maintenance operations, minimizing radiation levels and operating times, and providing contamination control during handling, transfer, and storage of radioactive material. The ALARA design criteria for the NUHOMS-12T system results in generally lower dose rates and exposures than those of the previously licensed Standardized NUHOMS system [7.3]. Specific features of the NUHOMS system that are directed toward ensuring ALARA include:

A.

Thick concrete walls and roof on the HSM to minimize the on-site and off-site dose contribution from the ISFSI B.

A thick shield plug on each end of the DSC to reduce the dose to plant workers while performing venting and sealing operations, and during transfer and storage of the DSC in the HSM C.

Use of a heavy shielded cask for DSC handling and transfer operations to ensure that the dose to employees and the general public is minimized D.

Fuel loading procedures which follow accepted practice and build on existing experience E.

A recess in the HSM access opening to dock and secure the cask during DSC transfer so as to reduce direct and scattered radiation exposure F.

Use of a heavy shielded door on the HSM to minimize direct and scattered radiation exposure G.

Use of a passive system design for long-term storage that requires minimal maintenance H.

Use of proven procedures and experience to control contamination during canister handling and transfer operations I.

Use of temporary shielding during DSC closure operations as necessary to further reduce the direct and scattered dose J.

A labyrinth design for the DSC vent and purge ports to minimize exposures during welding and filter installation and change-out.

Further ALARA measures may be implemented, as necessary, by the DOE-ID.

7.1.3 Operational Considerations Consistent with the DOE-IDs overall commitment to keep occupational radiation exposures ALARA, specific plans and procedures will be followed by ISFSI operations personnel to ensure that ALARA goals are achieved consistent with the intent of Regulatory Guides 8.8 [7.4] and 8.10 [7.5] and the requirements of 10 CFR Part 20. Since the ISFSI is a passive system, minimal maintenance is expected on a normal basis. Maintenance activities that could involve significant radiation exposure of

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SAR-II-8.4 CH 7 Rev. 7 7-6 of 7-35 personnel will be carefully planned utilizing previous operating experience. Maintenance activities will be performed using well-trained and certified personnel and proper equipment. Where applicable, formal ALARA reviews will be prepared which specify radiation exposure reduction techniques, such as those set out in Regulatory Guide 8.8.

7.2 Radiation Sources 7.2.1 Characterization of Sources Each DSC stored in the NUHOMS-12T ISFSI at INL contains up to 12 TMI-2 core debris canisters. At the time of the accident (March 28, 1979), the TMI-2 core was in its initial fuel cycle at 97%

of full power with an average core burnup of 3,175 megawatt days per metric ton of uranium (MWd/MTU). The core material includes fuel assemblies, including all fuel rods, control rods, axial power shaping rods, guide tubes, instrument tubes, spacer sleeves, spacer grids, and end fittings. Core material also includes primary startup sources, control rod spiders, coupling mechanisms and other miscellaneous material as noted in Reference 3.3.

The core had 177 Babcock and Wilcox 15x15 Mark B-4 fuel assemblies, identical in mechanical construction. Each assembly had 208 fuel rods, 16 control rod guide tubes, one instrument tube, seven spacer sleeves, eight spacer grids, and two end fittings. The guide tubes, spacer grids, and end fittings formed the structural cage which arranged the rods and tubes in a 15x15 array. The overall fuel assembly length was 165.62 inches, and the zircaloy-clad fuel rod length was 153.125 inches long. Three initial enrichments were used in the core, ranging from 1.98 weight percent (w/o) 235U to 2.98 w/o 235U.

The TMI-2 reactor accident left an intact reactor vessel containing damaged, partial length fuel assemblies in the outer region of the first core loading of the reactor. However, melting occurred in the central region of the core, producing a complex mix of melted/solidified corium, partial fuel assemblies, and loose debris such as pellets. Molten corium covered the lower core support structure at some time in the accident and damaged the bottom ends of the fuel assemblies. During decommissioning and cleanup, after the (corium) rocks had been reduced in size, the partial fuel assemblies still standing around the periphery of the core were severed at the debris bed level and sized with shears (if necessary) to fit into fuel canisters. Thus, the partial-length fuel assemblies do not have a continuous zircaloy fuel cladding to retain gaseous fission products that could later be released during dry storage.

The morphology of the corium rock and partial-length fuel assemblies core debris is complex.

Examination of the rocks showed that the uranium/iron/zirconium matrices contain cesium and iodine in stable compounds. The matrices remain stable up to 1600°C. Volatile contents including Cs-137 and I-129 are trapped as stable compounds within the corium uranium/zirconia matrix or the UO2 fuel pellet matrix remaining in the partial-length fuel assemblies.

Particulates were vacuumed with a submerged system inside the reactor vessel and captured within sintered metal filter cartridges. These filter cartridges are stored in TMI-2 Filter Canisters and comprise almost all of the releasable particulates in the TMI-2 ISFSI even though the total mass in the Filter Canisters is only 4 percent of the ISFSI total mass (including uncertainties). The corium and fuel solid pieces constitute 96 percent of the mass but are essentially clean of particulates due to the action of the submerged vacuum system.

The TMI-2 core debris is contained in three types of canisters: fuel, knockout, and filter. The internal structure of each canister type is the differentiating feature and is dictated by the canister function. Fuel canisters contain a mixture of large pieces of core debris up to partial-length, full cross-

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SAR-II-8.4 CH 7 Rev. 7 7-7 of 7-35 section fuel assemblies. Knockout canisters contain solids characterized as loose core rubble of a size small enough to be vacuumed up from the rubble bed. Filter canisters contain the filter elements and the small fuel fines (particulates) removed by the many filters which cleaned the water circulated through the submerged vacuum defueling system and the defueling water cleanup system.

Design basis neutron and gamma-ray sources for the TMI-2 debris canisters were calculated using the ORIGEN2.1 computer code [7.6]. An intact B&W 15x15 fuel assembly was modeled with a burnup of 3,175 MWd/MTU, an initial enrichment of 1.98 w/o 235U (use of the minimum enrichment results in conservative sources), and a specific power of 27.137 MW/MTU. These parameters correspond to a cycle length of 117 days prior to the accident. The specific power used in the ORIGEN model was increased by a factor of 1.879 to convert from a core average burnup to a peak assembly burnup. The resulting sources were decayed for 19 years (to March 1998) and scaled from the 1515 lb fuel assembly weight (including zircaloy cladding, grids, and end fittings) to the 1908 lb weight of the peak canister fuel debris payload.

This design basis source was then conservatively used in the shielding calculations for all of the TMI-2 canisters. The ORIGEN calculations resulted in a total isotopic inventory of 11,403,383 Curies. Assuming 30 DSCs each containing 12 canisters equals 3.17E+04 Curies per canister.

The neutron source, spectrum, and flux-to-dose rate conversion factors [7.8] are provided in Table 7.2-1. The energy structure is that of the CASK-81 cross-section library [7.9] and the energy spectrum corresponds to the neutron spectrum from the spontaneous fission of 240Pu, which is the primary neutron source. Similarly, the gamma-ray source, spectrum, and flux-to-dose rate factors are provided in Table 7.2-2. The gamma source spectrum was mapped from the 18 ORIGEN2.1 energy groups to the CASK-81 energy groups by assuming that the photon energies are logarithmically distributed within each energy group.

The total TMI-2 reactor core loading was 82,985.9 kg of uranium. Sources were calculated assuming that all 82,985.9 kg of uranium was burned at maximum power to produce the maximum radiation source. Thus, every kg of uranium contains the same maximum source.

The effluent dose calculations are based upon release of a fraction of the total isotopic inventory of the ISFSI. Releases are either gas, particulates from the TMI-2 Filter Canisters, or particulates from the TMI-2 Knockout and Fuel canisters (abbreviated as solids, meaning particulates released from solid materials contained in Knockout and Fuel canisters).

The airborne radiological effluent calculations for the off-normal and Accident conditions employ a bounding DSC containing the entire particulate mass of the 62 TMI-2 Filter Canisters (2,511.8 kgU) including 20% uncertainty (3,014.2 kgU). The total initial uranium mass of TMI-2 fuel at INTEC is 82,985.9 kgU so that the mass fraction of the bounding DSC is 0.04 (i.e., the 62 canisters contain 4% of the total fuel mass stored at the ISFSI, thus the bounding DSC used for the off-normal and Accident conditions contains 4% of the total fuel mass stored at the ISFSI). The radionuclide activity in the canisters is proportional to the weight of the canisters because all of the reactor fuel mass is assumed to have experienced the bounding irradiation conditions for the ORIGEN2 radioactive source calculations.

Thus, the total Curie source in the bounding DSC is 4% of the total Curie content in the ISFSI.

7.2.2 Airborne Radioactive Material Sources The potential for airborne radioactive material sources existed during fuel handling at the TAN facility, evacuating and sealing of the DSC, and DSC transfer and storage. Potential airborne releases from handling of the TMI-2 debris canisters at the TAN facility were handled in accordance with existing DOE-ID practices. DSC evacuating and sealing operations were performed using procedures which

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SAR-II-8.4 CH 7 Rev. 7 7-8 of 7-35 prevent airborne leakage. During these operations, all vent lines were routed to the existing radwaste systems at the TAN facility.

During storage, the DSC cavity is vented to the atmosphere through HEPA grade filters. No significant releases are expected from the DSC for the following reasons: (1) Much of the volatile fission product inventory was released during the accident and the remainder is entrapped within the fuel matrix as determined by extensive examinations performed on the core materials following the accident; (2) No differential pressure exists between the DSC and the atmosphere to provide a driving force for a significant release; (3) The DSC HEPA filters will prevent the release of solids and particulates as specified in Section 4.3.1.1. Although no significant releases are expected, both normal and accident (confinement failure) releases have been postulated with release fractions calculated using the methodology of Reference [7.7] to demonstrate the safety of the system. The postulated releases are non-mechanistic in that there is no conceivable event which would result in a significant release of the TMI-2 core debris material.

During normal operation, all 29 DSCs and the spare HSM are assumed to vent to the atmosphere.

Although excessive temperatures would be required to release the volatile fission products from the fuel matrix, 10% of the gas inventory (Kr-85 and H-3) is assumed to be released directly to the environment each year. Additionally, per Reference [7.7], 0.1% and 0.0001% per year of particulates and solids, respectively, are assumed to be released from the canisters to the DSC. Of these releases, 1% of the solids and particulates are released from the DSC to the HEPA filter system which removes all but 0.03% of particulates. The 62 filter canisters contain particulate matter and the remaining 282 canisters, which contain the larger debris material, contain primarily all solids with few particulates.

The general equation for the Source Term per Reference [7.23] is:

Source Term (Curies) = MAR

• DR

• ARF

• RF

• LPF The Material-at-Risk (MAR) for Normal Operation is the total particulate activity in the ISFSI. The MAR for off-normal operation is a bounding DSC with 12 debris canisters containing the historical particulate loading of all 62 TMI-2 Filter Canisters equal to 3,014.2 kgU (including a 20% uncertainty) out of the total initial TMI-2 core loading of 82,985.9 kgU, for a total mass fraction of 0.04. Because every kg of uranium contains the maximum possible radioactive inventory, the activity fraction is 0.04.

The Fuel and Knock-out Canisters are essentially clean of particulates because the submerged vacuum system removed particulates produced or disturbed by the removal operations of corium pieces or partial-length fuel assemblies or loose debris.

DR (damage ratio) is 1 for all canister types under all conditions.

The Airborne Release Fraction (ARF) is set based on the type of materials in the TMI-2 canisters as follows:

The particulate inventory is set at 1.0E-03 for Normal, Off-normal, and Accident release conditions For Normal release conditions, the remaining solid inventory (1-0.04) is set at 1.0E-6 The Off-normal and Accident release calculation assumes a release from particulates only.

The Resuspension Factor (RF) is 1.0E-02 for particulates and solids. Thus 1% of the airborne material is released from the DSC to the HEPA filter system which removes all but 0.03% of particulates.

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SAR-II-8.4 CH 7 Rev. 7 7-9 of 7-35 The Leak Path Factor (LPF) for Normal operation is the filtration of the HEPA filters that limit particulate release to 3.0E-04 of the source, as specified for the type of HEPA filter used in the DSCs. The LPF for particulates in off-normal conditions is 1.0E-02 as defined in the SAR Chapter 8 Section 8.1.4.

The LPF for particulates in the Accident condition is 1.0 as defined in the SAR Chapter 8 Section 8.2.7.

The LPF for gases is always 1.0.

As explained in Reference 7.22, the confinement barrier limits the release of radioactive material to the environment, but the barrier itself does not completely eliminate a release at the TMI-2 ISFSI under the various operating conditions. Under Normal operations, the seals perform their total confinement function such that all releases are channeled through the HEPA filters as calculated in the airborne release using an LPF of 0.0003. At the opposite extreme under the Accident condition, there exists a total failure of the confinement barrier wherein either the HEPA filters are completely missing (i.e., there exists an open hole to the environment) or the HEPA filters are completely blocked, and the seals have been breached to provide an open hole to the environment. Thus, the airborne release calculated for an Accident condition uses an LPF of 1. For the Off-normal condition, the HEPA filters are not blocked and remain fully functioning, and the seals continue to limit the release of radioactive material to the environment although at a reduced level from the Normal condition. Failure of a leak test does not constitute total failure of the confinement barrier, and the seal leak test is intentionally conservative with a very small acceptance criterion. Thus, for the airborne release calculation in an off-normal condition, a conservative LPF of 0.01 is used.

Table 7.2-3 provides the total quantity of potential airborne sources in each canister, the postulated annual release during normal operation, the postulated off-normal release, and the postulated accident release. Source activities have been taken from the design basis ORIGEN2.1 model discussed in Section 7.2.1. The postulated accident condition is a failure or breach of the HEPA filters of one bounding DSC containing all 62 TMI-2 Filter canisters (100% particulate). Release fractions for the accident are identical to those of normal operation with the exception that no credit is taken for the HEPA filters so that the LPF is 1.0. The periodic surveillances and radiological control surveys will limit the duration of undetected releases resulting from the postulated accident.

The postulated off-normal condition is leakage of the HEPA filter housing metallic seals of one bounding DSC containing 62 TMI-2 Filter Canisters (100% particulate). Release fractions for off-normal are identical to those of normal operation with the exception that a leakage LPF of 0.01 is applied. The duration of the off-normal condition is one year.

The source activity for each scenario is based on the MAR/Activity fraction and the LPF of the scenario.

The release equations for the calculation of Normal operation, off-normal, and Accident released Curies are:

Particulate Total ISFSI Curies = Activity/Can

• 12 Cans/DSC

• 30 DSCs = 11,403,383 Curies in ISFSI Normal = Total ISFSI Activity

• 0.04 * [(1.0E-3*1.0E-2)*(0.0003)] +

Total ISFSI Activity * (1-0.04) * [(1.0E-6*1.0E-2)*(0.0003)]

Off-normal = Total ISFSI Activity

• 0.04 * (1.0E-3*1.0E-2)*(0.01)

Accident = Total ISFSI Activity

• 0.04 * (1/12 months) * (1.0E-3*1.0E-2)*(1.0)

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SAR-II-8.4 CH 7 Rev. 7 7-10 of 7-35 Gas Normal = Total ISFSI Activity * [0.10

• 1.0

• 1.0]

Off-normal = Total ISFSI Activity

• 0.04 * [0.10

• 1.0

• 1.0]

Accident = Total ISFSI Activity

• 0.04 * (1/12 months) * [0.10

• 1.0

• 1.0]

The gas equations use a ten percent release for all scenarios, but the MAR is the entire ISFSI activity for Normal operations, and the ISFSI activity multiplied by 0.04 for off-normal, and the ISFSI activity times 0.04 times 1/12 for the accident condition.

The Normal operation, off-normal, and Accident curies are provided in Table 7.2-3, Postulated Airborne Radioactive Material Sources.

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SAR-II-8.4 CH 7 Rev. 7 7-11 of 7-35 Table 7.2-1. Neutron Energy Spectrum and Flux-to-Dose Conversion Factors for PWR Spent Fuel.a Cask Group Eupper (MeV)

Spectrum Source (n/sec/can)

Flux-to-Dose Factor (mrem/hr per n/cm2/sec) 1 1.492E+01 7.005E-05 4.830E+01 1.945E-01 2

1.220E+01 5.129E-04 3.536E+02 1.597E-01 3

1.000E+01 2.456E-03 1.693E+03 1.471E-01 4

8.180E+00 1.170E-02 8.067E+03 1.477E-01 5

6.360E+00 3.197E-02 2.204E+04 1.534E-01 6

4.960E+00 4.775E-02 3.292E+04 1.506E-01 7

4.060E+00 1.101E-01 7.590E+04 1.389E-01 8

3.010E+00 9.376E-02 6.465E+04 1.284E-01 9

2.460E+00 2.250E-02 1.551E+04 1.253E-01 10 2.350E+00 1.250E-01 8.615E+04 1.263E-01 11 1.830E+00 2.224E-01 1.534E+05 1.289E-01 12 1.110E+00 1.938E-01 1.336E+05 1.169E-01 13 5.500E-01 1.238E-01 8.538E+04 6.521E-02 14 1.110E-01 1.413E-02 9.741E+03 9.188E-03 15 3.350E-03 7.106E-05 4.900E+01 3.713E-03 16 5.830E-04 5.166E-06 3.562E+00 4.009E-03 17 1.010E-04 3.398E-07 2.343E-01 4.295E-03 18 2.900E-05 4.909E-08 3.385E-02 4.476E-03 19 1.010E-05 1.058E-08 7.294E-03 4.567E-03 20 3.060E-06 1.649E-09 1.137E-03 4.536E-03 21 1.120E-06 3.634E-10 2.505E-04 4.370E-03 22 4.140E-07 1.050E-10 7.237E-05 3.714E-03 Total 1.000E+00 6.895E+05

a. An additional source term for neutrons of 4.5 MeV average energy for each of two canisters, one stored in DSC 1/HSM 4 and another stored in DSC 5/HSM 22, is 7.3E6 n/sec/canister due to the presence of AmBeCm startup source material.

[7.20]

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SAR-II-8.4 CH 7 Rev. 7 7-12 of 7-35 Table 7.2-2. Gamma Energy Spectrum and Flux-to-Dose Conversion Factors for PWR Fuel.

Cask Group Eupper (MeV)

Source

(/sec/can)

Flux-to-Dose Factor (mrem/hr per /cm2/sec) 23 10.00 2.401E+02 8.772E-03 24 8.00 1.523E+03 7.479E-03 25 6.50 8.988E+03 6.375E-03 26 5.00 1.027E+04 5.414E-03 27 4.00 1.924E+05 4.622E-03 28 3.00 7.196E+06 3.960E-03 29 2.50 7.066E+07 3.469E-03 30 2.00 3.533E+10 3.019E-03 31 1.66 3.504E+12 2.628E-03 32 1.33 8.250E+12 2.205E-03 33 1.00 1.364E+12 1.833E-03 34 0.80 9.471E+13 1.523E-03 35 0.60 1.773E+14 1.173E-03 36 0.40 5.260E+12 8.759E-04 37 0.30 9.887E+12 6.306E-04 38 0.20 2.068E+13 3.834E-04 39 0.10 4.836E+13 2.669E-04 40 0.05 2.679E+14 9.348E-04 Total 6.372E+14

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SAR-II-8.4 CH 7 Rev. 7 7-13 of 7-35 Table 7.2-3. Postulated Airborne Radioactive Material Sources.

Isotope Activity (Ci/Can)

Activity Percent ISFSI (30 DSC)

Total Ci Normal Operations Release (1)

(Ci/yr)

Off-Normal Release (2)

(Ci/y)

Accident Release (3)

(Ci)

Cs-137 7.34E+03 23.154%

2.64E+06 3.25E-04 1.06E-02 8.81E-02 Ba-137m 6.94E+03 21.901%

2.50E+06 3.07E-04 9.99E-03 8.33E-02 Y-90 5.87E+03 18.525%

2.11E+06 2.60E-04 8.45E-03 7.04E-02 Sr-90 5.87E+03 18.519%

2.11E+06 2.60E-04 8.45E-03 7.04E-02 Pu-241 4.20E+03 13.256%

1.51E+06 1.86E-04 6.05E-03 5.04E-02 Kr-85 3.57E+02 1.125%

1.29E+05 1.29E+04 5.14E+02 4.28E+01 Pm-147 2.53E+02 0.797%

9.11E+04 1.12E-05 3.64E-04 3.04E-03 Am-241(4) 2.07E+02 0.653%

7.45E+04 9.16E-06 2.98E-04 2.48E-03 Co-60 1.42E+02 0.448%

5.11E+04 6.28E-06 2.04E-04 1.70E-03 Pu-239 1.14E+02 0.360%

4.10E+04 5.04E-06 1.64E-04 1.37E-03 Sm-151 9.38E+01 0.296%

3.38E+04 4.15E-06 1.35E-04 1.13E-03 Pu-240 6.01E+01 0.190%

2.16E+04 2.66E-06 8.65E-05 7.21E-04 Ni-63 5.57E+01 0.176%

2.01E+04 2.46E-06 8.02E-05 6.68E-04 Eu-154 4.30E+01 0.136%

1.55E+04 1.90E-06 6.19E-05 5.16E-04 H-3 4.27E+01 0.135%

1.54E+04 1.54E+03 6.15E+01 5.12E+00 Eu-155 3.10E+01 0.098%

1.12E+04 1.37E-06 4.46E-05 3.72E-04 Pu-238 2.95E+01 0.093%

1.06E+04 1.30E-06 4.25E-05 3.54E-04 Sb-125 1.93E+01 0.061%

6.95E+03 8.54E-07 2.78E-05 2.32E-04 Cs-134 7.96E+00 0.025%

2.87E+03 3.52E-07 1.15E-05 9.55E-05 I-129 3.28E-03 0.000%

1.18E+00 1.45E-10 4.72E-09 3.94E-08 Total 3.17E+04 1.14E+07 1.44E+04 5.76E+02 4.83E+01 Particulate 1.38E-03 4.50E-02 3.75E-01 Gas 1.44E+04 5.76E+02 4.80E+01 (1)

Normal operation release source for all 30 HSMs in the ISFSI, 29 DSCs plus one spare HSM.

(2)

Off-normal release source assumes that the HEPA filter housing seals leak with a Leak Path Factor of 0.01 for a period of 12 months.

(3)

Accident release source assumes that all five HEPA filters in a single DSC fail for a period of one month.

(4)

An additional 3.3 Ci of Am-241 is postulated for each of two canisters, one stored in DSC 1/HSM 4 and another stored in DSC 5/HSM 22, due to presence of AmBeCm startup source material. The corresponding additional release per canister for normal operation and accident conditions is postulated to be 2.2E-8 Ci/y and 3.2E-5 Ci, respectively [7.20].

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SAR-II-8.4 CH 7 Rev. 7 7-14 of 7-35 7.3 Radiation Protection Design Features 7.3.1 Installation Design Features The design considerations listed in Section 7.1.2 ensure that occupational exposures to radiation are ALARA and that a high degree of integrity is achieved through the confinement of radioactive materials inside the DSC. Applicable portions of Regulatory Position 2 of Regulatory Guide 8.8 [7.4] have been used as guidance.

A.

Access control to the ISFSI is through a controlled gate in the perimeter fence.

B.

Radiation shielding substantially reduces the exposure of personnel during system operations and storage.

C.

The NUHOMS system is a passive storage system; no process instrumentation or controls are necessary during storage. The only required surveillance or instrumentation is the periodic HEPA filter monitoring and periodic aging management inspections.

D.

Airborne contaminants are confined by the high integrity welded DSC assembly and the integral DSC HEPA filtration system.

E.

No crud is produced by the NUHOMS system.

F.

The necessity for decontamination is reduced by maintaining the cleanliness of the DSC and the cask during fuel loading and unloading operations (see Section 5.1). Additionally, the DSC and cask surfaces are smooth, nonporous, and are generally free of crevices, cracks, and sharp corners.

7.3.2 Shielding 7.3.2.1 Radiation Shielding Design Features. Radiation shielding is an integral part of all NUHOMS component designs. The features described in this section assure that doses to personnel and the public are ALARA. The following paragraphs and figures describe the radial and axial shielding provided by the NUHOMS-12T system.

Radial shielding during loading and transfer is provided primarily by the cask. This shielding includes a stainless-steel inner liner, lead, and a stainless-steel structural shell. Neutron shielding in the radial direction is provided by an outer metal jacket which forms an annulus with the cask structural shell.

This annulus is filled with a solid neutron absorbing material to provide neutron dose attenuation. During storage, radial shielding is primarily provided by the thick concrete walls of the HSM.

Axial shielding during loading and transfer is provided by the thick steel DSC shield plugs and the cask steel top and bottom cover plates. Two penetrations in the top shield plug provide a means for evacuating and venting the DSC. The penetrations are located on the perimeter of the DSC away from the TMI-2 canisters and contain sharp bends to minimize radiation streaming. During storage, axial shielding is provided by the concrete HSM rear wall, the concrete and steel HSM door, and the DSC shield plugs.

A thick steel cover provides shielding for the ventilation system.

The shielding geometry for the HSM and the cask are shown in Figure 7.3-1 and Figure 7.3-2, respectively. The geometry of the DSC vent is shown in Figure 7.3-3. Additional portable shielding during DSC handling, transport and transfer operations will be used by the DOE contractor as needed in accordance with existing ALARA practices. If used, the base of the welding machine used to place the top cover welds includes an integral neutron and gamma shield to minimize exposures during closure operations.

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SAR-II-8.4 CH 7 Rev. 7 7-15 of 7-35 7.3.2.2 Shielding Analysis. This section describes the radiation shielding analytical methods and assumptions used in calculating the NUHOMS system dose rates during the handling and storage operations. The dose rates of interest are calculated at the locations listed in Table 7.3-1 for the design basis TMI-2 core debris. These results are shown graphically in Figure 7.3-4 for the HSM. The computer codes used for analysis are described below, each with a brief description of the input parameters generic to its use. Descriptions of the individual analytical models used in the analysis are also provided.

A.

Computer Codes: Surface dose rates for the HSM and cask were calculated using the two-dimensional discrete ordinates transport computer code DORT [7.10]. The CASK cross section library, which contains 22 neutron energy groups and 18 gamma energy groups, is applied in an S16P3 approximation in DORT. Calculated radiation fluxes are multiplied by flux-to-dose conversion factors (Table 7.2-1 and Table 7.2-2) to obtain final dose rates. The DORT calculations use coupled neutron and gamma libraries. Therefore, dose rates from both primary and secondary gammas are calculated in each run.

Dose rates around the DSC vent and the DSC top covers were calculated using the three-dimensional monte-carlo computer code MCNP [7.11]. The MCNP code uses continuous energy cross section data for both neutron and gamma-ray transport. Calculated fluxes were converted to dose rates using the ANSI/ANS 6.1.1-1977 flux-to-dose factors [7.8]. The MCNP code was also used for the site dose calculations presented in Section 7.4.2.

Exposures due to postulated radioactive releases are calculated using the RSAC-7.2 computer code

[7.13]. The RSAC code was developed at INL to calculate the consequences of the release of radionuclides to the atmosphere. The code has the capability to generate a fission product inventory, decay and ingrow the inventory during transport through processes, facilities, and the environment, model the downwind dispersion of the activity, and calculate doses to downwind individuals. Atmospheric dispersion was calculated using the Markee model of Gaussian plume diffusion, as included in the RSAC code.

B.

HSM Surface Dose Rates: Three DORT models were used to model the fuel debris, DSC, and HSM. Two axisymmetric models, shown in Figure 7.3-1, model the HSM roof, front wall, and back wall. The roof model includes everything in Figure 7.3-1 above the DSC centerline and the floor model includes the remainder. The fuel debris is modeled as a homogenous cylinder and no credit is taken for the DSC basket. The 1/4 inch thick debris canister shell has been added to the DSC shell thickness. Using axisymmetric models for the HSM results in conservative dose rates over the bulk of the HSM surface. A third, Cartesian model, is used to estimate the dose rates along the gap between modules. Material properties for concrete are taken from ANSI/ANS-6.4 [7.14]

and no credit is taken for reinforcing bars. The neutron and gamma-ray dose rate results for the HSM are reported in Table 7.3-1 and Figure 7.3-4.

C.

Cask Dose Rates: The NUHOMS-MP187 cask will be used during loading, on-site transfer, and off-site transportation of the DSCs. An axisymmetric DORT model of the MP187 cask, using dimensions and materials described in the Rancho Seco ISFSI SAR [7.15], was generated to calculate the neutron and gamma dose rates on the surface of the cask. The fuel debris is modeled as a homogenous cylinder and no credit is taken for the DSC basket. The 1/4 inch thick debris canister shell has been added to the DSC shell thickness. Cask surface dose rates are reported in Table 7.3-1.

D.

DSC Vent and Cover Dose Rates: The DSC top shield plug and vent port dose rates were calculated using a three-dimensional MCNP model of the fuel debris, canisters, DSC basket, and DSC. The geometry of the model in the vicinity of the vent opening is shown in Figure 7.3-3. The TMI-2 debris is assumed to be homogenized within each TMI-2 canister, and the design basis gamma source term is applied to each canister. Neutron dose rates were not calculated because the HSM results provide assurance that the neutron doses are negligible. Dose rates at the surface of

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SAR-II-8.4 CH 7 Rev. 7 7-16 of 7-35 the vent were calculated using a point detector located as shown in Figure 7.3-3. Dose rates on the shield plug surface were calculated using a surface crossing tally located at the center of the plug.

Dose rates on the surface of the DSC shield plug and external to the vent and purge ports are provided in Table 7.3-1. The DSC shield plug dose rates are well below those of the Standardized NUHOMS system [7.3]. The peak dose rate external to the vent and purge ports is of a magnitude similar to the annulus dose rates previously observed during loading of NUHOMS DSCs.

Therefore, operational exposures for welding the NUHOMS-12T DSC will be similar to those observed for the Standardized NUHOMS system. To keep exposures ALARA, temporary shielding and remote handling equipment may be utilized when access to the vent and purge ports is required.

7.3.3 Ventilation As stated in Section 7.2.2, all process flows from the DSC during loading operations at TAN will be handled in accordance with DOE-IDs current practices. During storage, the DSC will be vented to the atmosphere through the venting system described in Section 4.3.1. As stated in Section 7.2.2 and in Section 4.3.1, no significant radioactive releases are expected through the venting system. The majority of volatile fission products were released during the TMI-2 accident or during the 10 years of storage at TAN. Because the DSC is vented to the atmosphere, there is no driving pressure to force material into the environment. In the event that any material does escape the DSC, the venting system includes HEPA grade filters (removal efficiencies of 99.97% for 0.3 micron particles).

Although no significant releases are expected from the DSC, Section 7.6 includes an exposure contribution from a postulated normal operation release. Chapter 8 includes an evaluation of an Accident estimating the consequences of a total failure of the HEPA filters. Chapter 8, Section 8.1.4 also includes an evaluation of an off-normal condition of a failed vent or purge port HEPA filter housing seal test. The results of these analyses show exposures to employees and the general public well below the applicable limits.

7.3.4 Area Radiation and Airborne Radioactivity Monitoring Instrumentation Radiological monitoring and contamination control at the ISFSI are performed to ensure that radiation exposure and release limits contained in 10 CFR Part 20 are not exceeded. The ISFSI was added to the existing INL radiological control program which monitors, as appropriate, radiation levels, contamination levels and airborne radioactivity.

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SAR-II-8.4 CH 7 Rev. 7 7-17 of 7-35 Table 7.3-1. Shielding Analysis Results.

Peak Dose Rate (mrem/hr)

Average Dose rate (mrem/hr)

Location Neutron Gamma Neutron Gamma DSC in HSM Roof/Side Walls 0.002 10.4 0.001 6.5 Front Wall 0.081 12.7 0.027 4.9 Rear Wall 0.235 104.5 0.034 7.6 Module Gap 0.002 4.34 n/a n/a DSC in Cask Cask Side 1.35 1.85 0.16 0.66 Cask Top 0.33 0.30 0.24 0.21 Cask Bottom 0.31 0.16 0.22 0.11 DSC Shield Plug n/a1 17.1 n/a n/a DSC Vent Port 422 926 n/a n/a (1)

Analysis performed only for gamma-ray doses. Neutron doses represent less than 1.5% of the total doses at these locations based on the DORT model described in section 7.3.2.2(C) and have been neglected.

(2)

An estimated peak neutron dose equivalent rate for each of two canisters, one stored in DSC 1/HSM 4 and another stored in DSC 5/HSM 22, attributed to AmBeCm startup source material [7.20].

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SAR-II-8.4 CH 7 Rev. 7 7-18 of 7-35 Figure 7.3-1. DSC and HSM Shielding Geometry.

Security-Related Information Figure Withheld Under 10 CFR 2.390.

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SAR-II-8.4 CH 7 Rev. 7 7-19 of 7-35 Figure 7.3-2. DSC and Cask Shielding Geometry.

Security-Related Information Figure Withheld Under 10 CFR 2.390.

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SAR-II-8.4 CH 7 Rev. 7 7-21 of 7-35 Figure 7.3-4. HSM Shielding Analysis Results.

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SAR-II-8.4 CH 7 Rev. 7 7-22 of 7-35 7.4 Estimated On-Site Collective Dose Assessment 7.4.1 Operational Dose Assessment This SAR section establishes the anticipated cumulative dose exposure to site personnel during the fuel handling and transfer activities associated with utilizing one NUHOMS HSM for storage of one DSC. Chapter 5 describes in detail the NUHOMS operational procedures, a number of which involve potential radiation exposure to personnel.

A summary of the operational procedures which result in radiation exposure to personnel is given in Table 7.4-1. The cumulative dose can be calculated by estimating the number of individuals performing each task and the amount of time associated with the operation. The resulting man-hour figures can then be multiplied by appropriate dose rates near the transfer cask surface, the exposed DSC top surface, or the HSM front wall. Dose rates are referenced in Table 7.3-1 for the DSC, cask, and HSM.

Every operational aspect of the NUHOMS system, from canister loading through sealing, transport, transfer, and operation is designed to assure that exposure to personnel is ALARA. Dose rates are kept ALARA by the shielded DSC end plugs and shielded cask. The vent and purge ports have been designed with bends and shield plates to minimize streaming during DSC sealing and filter change-outs.

Exposures are kept ALARA by performing most operations remotely as follows: (1) The debris canister drying and loading into the DSC were performed remotely in TAN Hot Shop; (2) If it is used, the welding machine was pre-installed on the top shield plug and top cover plates, away from the DSC, and operated remotely; (3) Transfer operations were performed inside the heavily shielded MP187 cask and trailer; and (4) Cask alignment operations were performed using a remote hand-held pendant and the hydraulic ram was operated using a remote power unit. In addition, many engineered design features are incorporated into the NUHOMS system which minimize occupational exposure to plant personnel during placement of fuel in dry storage as well as off-site dose to the nearest neighbor during storage. The resulting dose at the ISFSI site boundary is well within the limits specified by 10 CFR 72.

Because the predicted dose rates for the NUHOMS-12T system are well below those predicted for previous NUHOMS systems, occupational exposures for the TMI-2 ISFSI will be bounded by those observed at other installations. Based on experience from operating NUHOMS systems at Oconee, Calvert Cliffs, and Davis-Besse, the occupational dose for placing a DSC with TMI-2 core debris into dry storage for the operational steps listed in Table 7.4-1 will be much less than one person-rem. With the use of effective procedures and experienced ISFSI personnel, the total accumulated dose can be reduced below 500 person-mrem per DSC.

If a DSC vent housing seal fails a leak rate test, the estimated collective occupational exposure to reseat or replace the seals while the DSC remains in place is 60 person-mrem [7.22]. Subsequent to a DSC failing a seal leak rate test, any decontamination in a DSC vent housing area will increase the collective occupational exposure at a rate of 5 person-mrem/h.

7.4.2 Site Dose Assessment A site dose assessment for the ISFSI has been performed using the average HSM surface dose rates presented in Figure 7.3-4 as input. Locations of interest for the assessment include the INTEC fence (100 m) and the INL site boundary (13.7 km). The INTEC fence serves as the restricted area boundary for demonstrating compliance with the 10 CFR 20.1201 limit of 5 rem per year. Additionally, this assessment evaluated compliance with 10 CFR 20.1502 limit of 0.5 rem per year, requiring individual monitoring of external and internal occupational dose.

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SAR-II-8.4 CH 7 Rev. 7 7-23 of 7-35 The INL site boundary serves as the controlled area boundary for compliance with 10 CFR 72.104 for normal and anticipated occurrences where any real individual located beyond the controlled area may receive exposures related to effective dose equivalents as well as total effective dose equivalents. The INL site boundary also serves as the controlled area boundary for compliance with 10 CFR 72.106 for any design basis accident. Also included in this assessment are evaluations to ensure any member of the public, where full time occupancy could occur, are commensurate with 10 CFR 20.1301, where the TEDE to any maximally exposed individual (MEI) does not exceed 0.1 rem in a year. As the TMI-2 ISFSI is co-located within the controlled area boundaries of the INL, additional consideration was given to the overall emissions of radionuclides to the ambient air from Department of Energy facilities (INL) that shall not exceed an effective dose equivalent of 10 mrem/yr per 40 CFR 61.92 [7.21]. Direct and air-scattered radiation doses from all 30 HSMs at these locations have been calculated using the MCNP monte-carlo computer code [7.11].

The MCNP code was used to model the two rows of HSMs, the concrete basemat, and the surrounding land and air as shown in Figure 7.4-1. Source particles are started on the surfaces of the arrays which are modeled as solid concrete simply to account for self-shielding and scattering. No credit is taken for shielding by nearby structures or terrain.

Based on the low neutron dose rates on the surface of the HSMs (less than 0.05 mrem/hr as shown in Table 7.3-1), only gamma-rays are considered in the site dose assessment. Source particles are assumed to leave the surfaces of the modules with an angular distribution approximating a cosine function and an energy distribution of photons shielded by three feet of concrete. The total activity (photons/second) of each array face is used as input to the MCNP surface source.

Dose rates have been calculated using point detectors located around the ISFSI fence and the INTEC restricted area fence. Ring detectors are used to calculate dose rates at distances from 200 meters to 1000 meters. The 1000-meter dose rate is assumed to apply to all distances greater than 1000 meters.

Ring detectors have been used at these distances to improve the statistical accuracy of the calculations.

Because the air-scattered dose rates, which are relatively independent of the orientation relative to the ISFSI, dominate the results at these distances, this assumption has little effect on the results. Dose rates at various locations around the ISFSI and the acceptance criteria are provided in Table 7.4-2. As can be seen in Table 7.4-2, the site dose rates are well below the applicable 10 CFR 20.1201 (occupational worker) and 10 CFR 20.1502 (requirements to monitor occupational doses to workers) limits.

Table 7.4-2 also provides the external dose rate at the INL Site Boundary that may be received by an MEI as part of the assessment to ensure compliance with 10 CFR 72.104 and 10 CFR 72.106. The INL Site Boundary (13.7 km) is controlled by DOE-contracted security forces, and DOE exercises authority over its use. Figure 7.4-2 provides the dose rate as a function of distance from the ISFSI.

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SAR-II-8.4 CH 7 Rev. 7 7-24 of 7-35 Table 7.4-1. NUHOMS System Operations Enveloping Times for Loading One DSC.

Number of Workers Completion Time(3) (hours)

Location: TAN Facility Hot Cell Vacuum dry 12 TMI-2 Canisters(4) 2 168 Ready the DSC and MP187 Cask for Service 4

4.0 Place the DSC into the Cask in the TAN Hot Shop 4

1.0 Verify and Load the Dry TMI-2 Canisters(4) into the DSC 2

2.0 Open the TMI-2 canister vents(4) 2 1.0 Place the Top Shield Plug on the DSC(4) 3 1.0 Decontaminate the Outer Surface of the Cask(2) 3 1.0 Set-up Welder 3

1.5 Weld the Shield Plug to the DSC Shell and Perform Examination (1) 3 6.0 Purge the DSC and Backfill the DSC with Helium(1)(5) 2 4.0 Helium Leak Test the Shield Plug Weld(5) 2 1.0 Weld the Top Cover Plate to DSC Shell and Perform Examination(1) 3 16.0 Install Vent System Filter Assemblies and Transportation Covers 2

1.0 Evacuate and Backfill DSC with Helium(5) 2 4.0 Install the Top Internal Spacer and Cask Lid 2

1.0 Ready the Cask Support Skid and Transport Trailer for Service(2) 2 2.0 Place the Cask onto the Skid and Trailer 4

0.5 Install Skid Frame, Impact Limiters, and Personnel Barrier(5) 4

2.0 Location

ISFSI Site Ready the HSM and Hydraulic Ram System for Service(2) 2 2.0 Transport the Cask to the ISFSI 4

6.0 Remove Impact Limiters(5) 3 1.0 Position the Cask in Close Proximity with the HSM 3

1.0 Remove the Cask Lid (note: the top internal spacer is attached to the lid) 3 1.0 Align and Dock the Cask with the HSM 3

2.0 Position and Align Ram with Cask 3

1.0 Transfer the DSC from the Cask to the HSM 3

0.5 Move the Ram Clear of Cask and Un-Dock the Cask from the HSM 3

1.0 Install the HSM Access Door and Seismic Restraint 3

1.0 Open Rear Wall Access Door, Remove the Filter Transportation Covers(5), Visually Check HEPA Filters, and Close Access Door 2

1.0 Perform Radiation Survey 2

1.0 (1)

Monitoring operation - personnel may leave the radiation work area.

(2)

Operation may be performed in parallel with other activities.

(3)

Time shown for each operation is enveloping (i.e. these are operational times and not necessarily exposure time).

Actual times for similar operations have been considerably less.

(4)

Performed remotely in TAN Hot Shop.

(5)

Operations applicable only to the MP-187 transportation cask.

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SAR-II-8.4 CH 7 Rev. 7 7-25 of 7-35 Table 7.4-2. Dose Rates in the Vicinity of the TMI-2 ISFSI at INL Location Dose Rate (mrem/hr)

Acceptance Criteria INTEC Fence (max)(1) 4.78E-02 5 rem/year (10 CFR 20.1201)

INL Site Boundary(2) 3.68E-06 0.1 rem/yr (10 CFR 20.1301)

Notes:

(1) Represents the restricted area boundary. Individuals entering this area are monitored for exposure to radiation and radioactive material at levels sufficient to demonstrate compliance occupational dose limits.

(2) Represents the controlled area boundary. Maximally Exposed Individual (MEI) is assumed to be at this location. This area is controlled by DOE-contracted security forces and DOE exercises authority over its use.

Figure 7.4-1. ISFSI Geometry for Site Dose Calculations.

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SAR-II-8.4 CH 7 Rev. 7 7-26 of 7-35 1E-9 1E-8 1E-7 1E-6 1E-5 1E-4 1E-3 1E-2 1E-1 1E+0 1E+1 10 100 1000 10000 Distance from ISFSI Center (ft)

Dose Rate (mrem/hr)

Figure 7.4-2. Dose Rate Versus Distance from the ISFSI.

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SAR-II-8.4 CH 7 Rev. 7 7-27 of 7-35 7.5 Health Physics Program 7.5.1 Organization The radiation protection program at the TMI-2 ISFSI is described in a Radiation Protection Plan (RPP) which has been developed in accordance with 10 CFR 20 [7.1]. The program is implemented through the RPP as well as several INL radiation protection procedures referenced in the RPP. A Facility Safety Officer (FSO) that is independent of the facility operations organizational element is responsible for implementation of the radiation protection program. The FSO reports to the program manager at a professional level equivalent to operations personnel. Several matrixed radiation protection professionals (engineers and technicians) maintain qualification to implement the radiation protection program under the direction of the FSO.

7.5.2 Equipment, Instrumentation, and Facilities The radiological equipment, instrumentation, and facilities for the ISFSI will be those currently used at INL, which includes complete health physics facilities to support the ISFSI radiation protection program. Properly selected, operated, maintained, and calibrated radiological instrumentation is employed at TMI-2 ISFSI in order to implement an effective radiation protection program. Typical instruments include:

A.

A variety of portable beta-gamma detectors and suitable rate meters are used for surface contamination monitoring B.

Low-background alpha-beta counters indicate the level of contamination on smears and air filters C.

A variety of portable gamma and neutron dose rate meters D.

Gamma spectroscopy capability for radionuclide identification.

The INL radiation protection procedures referenced in the RPP provide requirements for the calibration, response check, operational inspection, maintenance, and repair of standard radiological instruments used at INL. These procedures are applicable to both fixed and portable instruments.

7.5.3 Procedures 7.5.3.1 Radiation Protection Practices. Radiation protection practices employed at the TMI-2 ISFSI include personnel protective equipment and permanent or temporary shielding as necessary to minimize the potential for personnel contamination and radiation exposure. Additional radiation protection practices used to control exposure include posting radiological areas, controlling activities within the radiologically controlled areas, and ensuring entry and exit control.

Posting Areas: NRC-approved signs, labels, and radiation symbols are conspicuously posted for radiologically controlled areas as required by 10 CFR Part 20. Each access point to a radiologically controlled area is posted. The size of the area is determined using the guidelines in 10 CFR Part 20.

The alteration or removal of control barriers is performed by, or under the direction of, the RCTs.

Radiological Work Permits (RWPs): The RWP is an administrative mechanism used to establish radiological controls for intended work activities. The RWP informs workers of area radiological conditions and entry requirements and provides a mechanism to relate worker exposure to specific work activities. An RWP contains pertinent information for performing the intended work safely and within ALARA guidelines. The information includes, but is not limited to, a description of the

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SAR-II-8.4 CH 7 Rev. 7 7-28 of 7-35 work, radiological conditions of the work area, dosimetry requirements, stay time controls, and special dose or contamination reduction considerations.

RWPs are required to enter high and very high radiation areas, high contamination areas, and airborne radioactivity areas. Job-specific RWPs are used for non-routine operations or work in areas with changing radiological conditions. Job-specific RWPs remain in effect only for the duration of the job, whereas general RWPs are used for routine or repetitive activities, such as tours and inspections, or minor work activities in areas with well-characterized and stable radiological conditions. General RWPs can be approved for periods of up to one year.

The RWP is approved by the job supervisor and job controller for the work area as well as the appropriate radiation protection management representative and facility/area manager. Radiological surveys are reviewed to evaluate the adequacy of the RWP requirements. Workers acknowledge that they have read, understand, and will comply with the RWP before their initial entry to the area and after any revisions to the RWP.

Entry and Exit Control: Personnel entry control is maintained for each radiological area per 10 CFR Part 20. The degree of control is commensurate with existing and potential radiological hazards within the area. One or more of the following methods are used to ensure control:

radiological posting, control devices on entrances, conspicuous visual and audible alarms, locked entrance ways, or administrative controls.

7.5.3.2 Dosimetry. 10 CFR Part 20 establishes the policy, requirements, and training necessary for assignment and use of external dosimetry. External dosimetry devices used for monitoring occupational whole-body exposure are accredited by the DOE Laboratory Accreditation Program for the appropriate radiation types and categories.

External dosimetry provides indication of the radiation exposures received by personnel, equipment, and the environment. External dosimetry devices are capable of indicating both penetrating and nonpenetrating radiation exposure that contribute to a persons occupational exposure. External dosimetry for equipment and the environment provides an indication of the general radiation field in the ISFSI areas. All external dosimetry devices used at the ISFSI are analyzed by the INL Radiation Dosimetry and Records group.

Personnel dosimetry badges (either direct reading and/or electronic dosimetry) are issued to all personnel performing radiological work within the ISFSI, as required by the applicable RWP. Each employee is responsible for wearing his or her assigned badge while within the ISFSI. The optically stimulated luminescence (OSL) badges are analyzed to provide input into a computerized record system that accumulates employee and visitor exposure information. The reports are transmitted to management to inform them of the exposure status of all employees. In addition to external dosimetry, all ISFSI personnel who are likely to receive intakes resulting in an effective dose equivalent greater than 100 mrem undergo initial, periodic, and termination baseline whole-body counts or bioassays.

7.5.3.3 Respiratory Protection. The RPP and referenced procedures provide guidelines for selecting respiratory equipment for protection against airborne radioactivity. These documents incorporate the requirements of ANSI Z88.2, Practices for Respiratory Protection. Respirators for radiological exposure control are used in accordance with 10 CFR Part 20. ISFSI personnel are formally trained and qualified before using respiratory equipment.

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SAR-II-8.4 CH 7 Rev. 7 7-29 of 7-35 7.5.3.4 Radiological Protection Training. All individuals requiring access to radiological controlled areas at the ISFSI receive training that emphasizes the nature of radiological conditions and the control of radiation exposure. 10 CFR Part 20 provides instructions for determining the training requirements, based on activities and responsibilities of the workers.

Levels of training for RCTs are commensurate with the technicians assignment. Qualifications for RCTs consist of standardized course material, on-the-job training, and both a comprehensive written examination and final examination with the Oral Examination Board. The level of RCT qualification is based on the education, experience, training, orientation, and other qualification achieved and maintained by the individual. RCT qualifications follow a 2-year cycle of continuing training evaluated with a written and oral examination.

Training is also provided for other radiological support personnel (who provide health physics and radiological engineering, dosimetry, bioassay, independent oversight, and instrumentation calibration functions) to ensure that these personnel have the technical qualifications pertinent to their assigned duties. Health physicists and other radiation protection professionals with baccalaureate degrees in science or engineering or equivalent work experience are available to support the radiation protection program.

7.6 Estimated Off-Site Collective Dose Assessment 7.6.1 Effluent and Environmental Monitoring Program The INL environmental surveillance program maintains a network of low-volume samplers to monitor for airborne radioactivity [7.21]. The network includes 14 onsite locations, 7 INL boundary locations, and 6 offsite locations. One of the onsite samplers is located about 1100 feet northwest of the ISFSI site near the INTEC entrance and west perimeter road. Samplers for monitoring tritium in the atmosphere are also located at two onsite and four offsite locations.

The INL environmental surveillance program also includes direct measurements of ambient (environmental) radiation levels using OSL dosimeters. These devices measure ionizing radiation exposure rates due to the combined sources of natural radioactivity in the air and soil, cosmic rays, residual fallout from nuclear weapons tests, and radioactivity from INL site operations. Dosimeters are located at 135 onsite locations and 27 offsite locations.

The ISFSI specific radiological environmental monitoring program includes monthly airborne radioactivity sampling within the ISFSI perimeter fence, direct radiation monitoring with OSL dosimetry placed along the ISFSI perimeter fence, and periodic loose surface radioactive contamination monitoring adjacent to each DSC vent and purge port and each HSM drain line.

The INL meteorological and environmental surveillance programs will be continued through the life of the TMI-2 ISFSI. The ISFSI specific radiological environmental monitoring program will also continue through the life of the TMI-2 ISFSI. Only the results of the ISFSI specific radiological environmental monitoring program will be reported to meet the 60-day reporting requirement of 10 CFR 72.44. The results of the INL environmental surveillance program will be available under separate cover, but outside the scope of the 60-day reporting requirement of 10 CFR 72.44.

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SAR-II-8.4 CH 7 Rev. 7 7-30 of 7-35 7.6.2 Analysis of Multiple Contribution Commensurate with the requirements of 10 CFR 72.104(a), the following exposure paths to the MEI have been assessed to ensure the total effective dose equivalents remain at or below the criteria for radioactive materials in effluents and direct radiation levels from the TMI-2 ISFSI:

The annual dose calculated for the maximally exposed individual as a result of current and projected INL sitewide emissions is about 0.05 mrem/year The ISFSI direct and air-scattered radiation at the controlled area boundary from Table 7.4-2 is 0.03 mrem/year The TMI-2 ISFSI effective dose equivalent from radioactive effluents at the controlled area boundary (13.7 km), from Table 7.6-1, Normal Operations, is 0.01 mrem/year The TMI-2 ISFSI effective dose equivalent from radioactive effluents at the controlled area boundary (13.7 km), from Table 7.6-2, Off-normal Operations, is 0.02 mrem/year.

By summation of the above contributing dose emission, the total effective dose equivalent to the MEI from normal operations and off-normal conditions is 0.11 mrem/year.

40 CFR 61.92 effective dose equivalent limit is 10 mrem/year 10 CFR 72.104 total effective dose equivalent limit is 25 mrem/year.

7.6.3 Estimated Dose Equivalents to the MEI Dose equivalents from effluents as a function of distance from the ISFSI have been calculated using the RSAC-7.2 computer code [7.13]. Meteorological parameters were generated using hourly meteorological data taken over a six-year period. The greatest annual average atmospheric dispersion factor over this period is 4.81x10-8 s/m3, established for Atomic City, 18 km from the INTEC. This corresponds to a 4.45 m/s wind speed with class C stability, which was assumed for the effluent dose calculations. Normal operation nuclide releases from Table 7.2-3 were input to the code for calculations of exposure from inhalation, ingestion, ground surface dose, and immersion.

RSAC-7.2 calculates inhalation doses using the ICRP 30 [7.16] and Federal Guidance Report No. 11 [7.17] dose conversion factors. The Committed Dose Equivalent (CDE) for each organ or tissue is multiplied by the appropriate weighting factor and summed to determine the Committed Effective Dose Equivalent (CEDE). Ingestion doses are calculated based on the models and equations of Regulatory Guide 1.109 [7.18]. The dose from radioactivity deposited on the ground surface is calculated using Federal Guidance Report No. 12 [7.19] dose-rate conversion factors. Immersion doses are calculated using a finite plume model.

Table 7.6-1, Normal Operation provides the thyroid organ dose and the effective dose equivalent (applicable to the whole body) at distances of 100 meters (assumed INTEC boundary), 1000 meters, 13.7 km (INL boundary, controlled area boundary, and assumed MEI), and 18 km (Atomic City).

Table 7.6-1 also provides the calculated /Q for each distance. Additionally, to assess total effective dose equivalent congruent with 10 CFR 72.104, the applicable dose rates from Table 7.4-2 are utilized. Doses for off-site locations related to the MEI assessment utilize a stay time of 8,760 hours0.0088 days <br />0.211 hours <br />0.00126 weeks <br />2.8918e-4 months <br /> per year.

Table 7.6-2, Off-normal Condition provides the thyroid organ dose and the effective dose equivalent (applicable to the whole body) for the off-normal condition of HEPA filter housing seal leakage.

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SAR-II-8.4 CH 7 Rev. 7 7-31 of 7-35 The direct and air-scattered radiation dose rates from Table 7.4-2 for the 13.7 km (INL boundary, controlled area boundary, assumed MEI) is 0.03 mrem/year. Therefore, the total effective dose equivalent as stated in 7.6-2 is 0.11 mrem/year, well below the most conservative standard as stated in 40 CFR 61.62 of 10 mrem/year effective dose equivalent, as well as 10 CFR 72.104 total effective dose equivalent of 25 mrem/year to the MEI located beyond the controlled area.

Table 7.6-3, Accident Condition provides the thyroid organ dose and the effective dose equivalent (applicable to the whole body) for the Accident condition of HEPA filter failure. Applicable to 10 CFR 72.106, the receptor location for the MEI (INL Boundary), the assessment utilized a stay time of 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br />, as related to a 30 day exposure period. The direct and air-scattered radiation dose rates from Table 7.4-2 for the 13.7 km (assumed MEI location), is 0.003 mrem/month. The effective dose equivalent at the MEI receptor point is 0.5 mrem/month, therefore the total effective dose equivalent rounded up is 0.6 mrem/month. 10 CFR 72.106 assumes an annual exposure limit of 5 rem TEDE, therefore, this assessment assumes the dose would relate to a 30-day accident exposure period resulting in a TEDE to the MEI of 0.6 mrem.

7.6.4 Estimated Dose Equivalents to the Occupational Worker Table 7.6-1, Normal Operations provides the thyroid organ dose and the effective dose equivalent at 4 receptor points (applicable to the whole body) to include 100 meters (assumed INTEC boundary),

1000 meters, 13.7 km (INL boundary, controlled area boundary, and assumed MEI), and 18 km (Atomic City). Table 7.6-1 also provides the calculated /Q for each distance. To assess total effective dose equivalent congruent with 10 CFR 20.1201, the applicable dose rates from Table 7.4-2 are utilized.

Table 7.6-2, Off-normal Condition provides the thyroid organ dose and the effective dose equivalent (applicable to the whole body) for the off-normal condition of HEPA filter housing seal leakage. In accordance with 10 CFR 20.1201, the applicable receptor locations for the occupational dose limits for adults is the 100 meters (INTEC Fence), with a maximum stay time of 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />. The combined effective dose equivalents from Tables 7.6-1 and 7.6-2 at the 100 meters receptor point, and the direct and air-scattered radiation dose rates from Table 7.4-2, where an occupational worker could be assumed to be located for up to 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year (full time occupancy) is rounded up to 115 mrem/year. As per 10 CFR 202.1502, where occupational exposure is requiring for individual monitoring is 10% of the 10 CFR 20.1201 values of 5 rem, the standard is not met, requiring individual monitoring. Table 7.6-3, Accident Condition provides the thyroid organ dose and the effective dose equivalent (applicable to the whole body) for the Accident Condition of HEPA filter failure. In accordance with 10 CFR 20.1201 the applicable receptor locations for the occupational dose limits for adults is the 100 meters (INTEC Fence),

with a maximum stay time of 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br />. The combined effective dose equivalents from Tables 7.6-3 at the 100 meters receptor point, and the direct and air-scattered radiation dose rates from Table 7.4-2, where an occupational worker could be assumed to be located for up to 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> per year (one month exposure) is rounded up to 151 mrem/month. Per 10 CFR 20.1201, the annual occupational dose limit is 5 rem. Assuming that the Accident condition is corrected within 30 days or additional controls related to ALARA are invoked, the annual exposure to this event is 151 mrem. Therefore, the limits of 10 CFR 20.1201 of 5 rem, nor the limits of 10 CFR 1502 of 0.5 rem are not exceeded and no additional controls are necessary.

7.6.5 Liquid Release Even though the HSM is provided with a drain to remove any moisture that may get into the HSM, no liquids are expected to be released from the TMI-2 ISFSI.

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SAR-II-8.4 CH 7 Rev. 7 7-33 of 7-35 Table 7.6-3. Accident Condition Estimated Effluent Dose Equivalents (rem per one-month duration, 100% occupancy for off-site locations)

RSAC-7 Accident Distance (meters)

Location Pathway 100 1,000 13,700 18,000 X/Q (s/m3) 6.88E-04 2.03E-05 4.11E-07 3.34E-07 Thyroid (I-129)

Inhalation (CDE) 1.51E-04 4.44E-06 2.98E-07 2.42E-07 Ingestion (CDE) 0.00E+00 0.00E+00 1.36E-04 1.10E-04 Ground Surface (DE) 7.70E-04 2.26E-05 1.90E-06 1.54E-06 Total Organ Dose 9.21E-04 2.70E-05 1.38E-04 1.12E-04 EDE (whole Body)

Inhalation (CEDE) 1.43E-01 4.20E-03 2.82E-04 2.29E-04 Ingestion (CEDE) 0.00E+00 0.00E+00 2.24E-04 1.82E-04 Ground Surface (EDE) 7.72E-04 2.26E-05 1.90E-06 1.54E-06 Cloud Gamma (EDE) 2.93E-07 4.25E-08 1.34E-08 1.08E-08 Total EDE 1.44E-01 4.22E-03 5.08E-04 4.13E-04

1)

No ingestion exposure is assumed within the INL site boundary.

2)

On-site exposures calculated using 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week occupancy.

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SAR-II-8.4 CH 7 Rev. 7 7-34 of 7-35 7.7 References 7.1 Title 10, Energy, Code of Federal Regulations, Part 20, Standards for Protection Against Radiation.

7.2 Title 10, Energy, Code of Federal Regulations, Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste.

7.3 Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel, VECTRA Technologies, Inc., Revision 4A, File Number NUH003.0103, 1996.

7.4 U. S. Nuclear Regulatory Commission, Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will be As Low As Reasonably Achievable,"

Regulatory Guide 8.8.

7.5 U. S. Nuclear Regulatory Commission, Operating Philosophy for Maintaining Occupational Radiation Exposures As Low as is Reasonably Achievable, Regulatory Guide 8.10.

7.6 ORIGEN2.1 - Isotope Generation and Depletion Code - Matrix Exponential Method, CCC-371, Oak Ridge National Laboratory, RSIC Computer Code Collection, August 1991.

7.7 Staley, C. S., Doses to Maximally Exposed Individuals due to Potential Airborne Releases from the INEL Storage of the TMI-2 Fuel Project, EDF Serial Number EMA-96-001, File Number 219-02.0034, February 16, 1996.

7.8 American Nuclear Society Standards Committee Working Group ANS-6.1.1, American National Standard Neutron and Gamma-Ray Flux-to-Dose-Rate Factors, ANSI/ANS-6.1.1-1977, American Nuclear Society, 1977.

7.9 Radiation Shielding Information Center, "CASK81 22 Neutron, 18 Gamma Ray Group, P3 Cross-Sections for Shipping Cask Analysis," DLC-23, June 1987.

7.10 DORT-PC - Two-Dimensional Discrete Ordinates Transport Code System, CCC-532, Oak Ridge National Laboratory, RSIC Computer Code Collection, October 1991.

7.11 MCNP 4 - Monte-Carlo Neutron and Photon Transport Code System, CCC-200A/B, Oak Ridge National Laboratory, RSIC Computer Code Collection, October 1991.

7.12 Title 40, Code of Federal Regulations, Part 61, National Emission Standards for Hazardous Air Pollutants.

7.13 Schrader, B. J., Radiological Safety Analysis Computer (RSAC) Program Version 7.2 Users Manual, INL/EXT-09-15275, Revision 1, Idaho National Laboratory, October 2010.

7.14 American Nuclear Society Standards Committee Working Group ANS-6.4, "American National Standard Guidelines on the Nuclear Analysis and Design of Concrete Radiation Shielding for Nuclear Power Plants," ANSI/ANS-6.4-1977, American Nuclear Society, 1978.

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SAR-II-8.4 CH 7 Rev. 7 7-35 of 7-35 7.15 Rancho Seco Independent Spent Fuel Storage Installation Safety Analysis Report, Sacramento Municipal Utility District, Docket No. 72-11.

7.16 ICRP, Limits for Intakes of Radionuclides by Workers, Part 1, ICRP Publication 30, Pergamon Press, Oxford, Great Britain, 1979.

7.17 Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, U.S. Environmental Protection Agency, EPA-5201/1-88-020, Washington D.C., 1988.

7.18 U. S. Nuclear Regulatory Commission, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, Regulatory Guide 1.109.

7.19 Federal Guidance Report No. 12, External Exposures to Radionuclides in Air; Water; and Soil, EPA-402/R-93-081, Washington D.C., 1993.

7.20 Hall, G. G., Impact of AmBeCm Sources on the TMI-2 ISFSI Design Basis, Engineering Design File No. 1793, Revision 4, March 15, 2001.

7.21 DOE/ID-12082, Idaho National Engineering and Environmental Laboratory Site Environmental Report, Calendar Year 2001. December 2002.

7.22 Radiological Evaluation of TMI-2 ISFSI Technical Specification 3.1.1, Engineering Design File No. 4728, Revision 5, November, 2022.

7.23 DOE-HDBK-3010-94 (reaffirmed 2013), Airborne Release Fractions/Rates and Respirable Fractions for Non-reactor Nuclear Facilities.

NLF-RPT-309, TMI-2 ISFSI Biennial Update Report for 2025 Attachment B TMI-2 SAR-II-8.4 Chapter 8 (PDF attached)

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SAR-II-8.4 CH 8 Rev. 6 8-1 of 8-98 CHAPTER 8 Analysis of Design Events

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SAR-II-8.4 CH 8 Rev. 6 8-2 of 8-98 CONTENTS

8.

Analysis of Design Events............................................................................................................. 8-5 8.1 Normal and Off-Normal Operations.................................................................................... 8-5 8.1.1 Normal Load Structural Analysis............................................................................ 8-5 8.1.2 Off-Normal Load Structural Analysis................................................................... 8-18 8.1.3 Thermal Hydraulic Analysis................................................................................. 8-21 8.1.4 Storage with Detected Leakage of Vent or Purge Port HEPA Filter Housing Seals...................................................................................................................... 8-26 8.2 Accident Analyses for the ISFSI........................................................................................ 8-61 8.2.1 Reduced HSM Self-Shielding............................................................................... 8-62 8.2.2 Tornado Winds/Tornado Missile........................................................................... 8-63 8.2.3 Earthquake............................................................................................................. 8-70 8.2.4 Flood..................................................................................................................... 8-75 8.2.5 Accidental Cask Drop........................................................................................... 8-75 8.2.6 Lightning............................................................................................................... 8-79 8.2.7 DSC Leakage......................................................................................................... 8-79 8.2.8 Accident Pressurization of DSC............................................................................ 8-81 8.2.9 Fire and Explosion................................................................................................ 8-81 8.2.10 Blockage of Space Between Adjacent HSMs....................................................... 8-82 8.2.11 Basaltic Lava Flow................................................................................................ 8-83 8.3 Load Combination Evaluation............................................................................................ 8-91 8.3.1 DSC Confinement Boundary Load Combination Evaluation............................... 8-91 8.3.2 DSC Confinement Boundary Fatigue Evaluation................................................. 8-91 8.3.3 MP187 Cask Load Combination Evaluation......................................................... 8-91 8.3.4 MP187 Cask Fatigue Evaluation........................................................................... 8-91 8.3.5 HSM Load Combination Evaluation..................................................................... 8-91 8.3.6 Thermal Cycling of the HSM................................................................................ 8-92 8.3.7 DSC Support Structure Load Combination Evaluation......................................... 8-92 8.4 Site Characteristics Affecting Safety Analysis................................................................... 8-95 8.5 References.......................................................................................................................... 8-96

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SAR-II-8.4 CH 8 Rev. 6 8-3 of 8-98 FIGURES 8.1-1.

Heating 7 Model of NUHOMS-12T DSC Used in HSM......................................... 8-38 8.1-2.

HSM HEATING 7 Results for 45F Ambient............................................................. 8-40 8.1-3.

NUHOMS-12T HSM Temperature Distribution for 45F Ambient......................... 8-44 8.1-4.

HEATING 7 Model of NUHOMS-12T DSC............................................................ 8-45 8.1-5.

NUHOMS-12T DSC Internal Temperature Distribution for 87F Ambient............. 8-46 8.1-6.

DSC Shell Stress Analysis Diagram............................................................................ 8-50 8.1-7.

DSC Shell Axisymmetric Analytical Model................................................................ 8-51 8.1-8.

DSC Bottom Cover Plate/Grapple Ring Analytical Model......................................... 8-52 8.1-9.

NUHOMS-12T DSC Spacer Disc Analytical Model................................................. 8-53 8.1-10. NUHOMS-12T DSC Spacer Disc Applied 8g Loading............................................ 8-54 8.1-11. DSC Axial Jam Condition........................................................................................... 8-55 8.1-12. DSC Binding (Pinching) Condition............................................................................. 8-55 8.1-13. Prefabricated HSM Analytical Model......................................................................... 8-56 8.1-14. Analytical Model for DSC Support Structure.............................................................. 8-57 8.1-15. Analytical Model for HSM Base Unit Thermal Conditions........................................ 8-58 8.1-16. Typical HSM Reinforcement....................................................................................... 8-59 8.1-17. MP187 Cask Handling Loads...................................................................................... 8-60 8.2-1.

MP187 Cask Postulated Drop Accident Scenarios...................................................... 8-87 8.2-2.

Tornado Missile Impact Model.................................................................................... 8-88 8.2-3.

Horizontal and Vertical Seismic Design Response Spectra......................................... 8-89 8.2-4.

TMI-2 ISFSI Volcanic Mitigation Map and Cross Section......................................... 8-90

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SAR-II-8.4 CH 8 Rev. 6 8-4 of 8-98 TABLES 8.1-1.

NUHOMS-12T Normal Operating Loading Identification................................. 8-27 8.1-2.

NUHOMS-12T Off-Normal Operating Loading Identification.......................... 8-27 8.1-3.

Mechanical Properties of Materials....................................................................... 8-28 8.1-4.

Estimated NUHOMS-12T Component Weights................................................. 8-30 8.1-5.

NUHOMS-12T DSC Operating and Accident Pressures.................................... 8-30 8.1-6.

Thermophyiscal Properties of Materials............................................................... 8-31 8.1-7.

Temperature Dependent Thermophysical Properties............................................ 8-31 8.1-8.

Thermal Load Case Definitions for NUHOMS-12T HSM Structural Analysis................................................................................................................. 8-32 8.1-9.

NUHOMS-12T DSC Thermal Analysis Results Summary................................ 8-33 8.1-10

. NUHOMS DSC Thermal Analysis Results Summary....................................... 8-34 8.1-11.

Maximum NUHOMS-12T DSC Stresses for Normal Loads.............................. 8-35 8.1-12.

Maximum NUHOMS-12T DSC Stresses for Off-Normal Loads....................... 8-36 8.1-13.

Maximum DSC Structure Stresses for Normal and Off-Normal Loads................ 8-36 8.1-14.

Maximum DSC Support Structure Vertical Displacements for Normal and Off-Normal Loads................................................................................................. 8-37 8.1-15.

Maximum NUHOMS-12T HSM Reinforced Concrete Bending Moments and Shear Forces for Normal and Off-Normal Loads........................................... 8-37 8.2-1.

Postulated Accident Loading Identification.......................................................... 8-85 8.2-2.

Maximum HSM Reinforced Concrete Bending Moments and Shear Force for Accident Loads...................................................................................................... 8-86 8.2-3.

Maximum NUHOMS-12T DSC Stresses for Drop Accident Loads................... 8-86 8.3-1.

NUHOMS-12T DSC Enveloping Load Combination Results for Normal and Off-Normal Loads (ASME Service Levels A and B)..................................... 8-93 8.3-2.

NUHOMS-12T DSC Enveloping Load Combination Results for Accident Loads (ASME Service Level C)............................................................................ 8-93 8.3-3.

NUHOMS-12T DSC Enveloping Load Combination Results for Accident Loads (ASME Service Level D)........................................................................... 8-94 8.3-4.

HSM Enveloping Load Combination Results....................................................... 8-94 8.3-5.

DSC Support Structure Enveloping Load Combination Results........................... 8-95

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8.

ANALYSIS OF DESIGN EVENTS In previous chapters of this SAR, the features of the TMI-2 ISFSI NUHOMS-12T system that are important to safety have been identified and discussed. The purpose of this chapter is to present the engineering analyses for normal and off-normal operating conditions, and to establish and qualify the system for a range of credible and hypothetical accidents. The term NUHOMS-12T refers to the TMI-2 ISFSI NUHOMS-12T system.

In accordance with NRC Regulatory Guide 3.48 [8.1], the design events identified by ANSI/ANS 57.9-1984 [8.2] form the basis for the accident analyses performed for the NUHOMS-12T system. Three categories of design events are defined. Normal operating and off-normal events are addressed in Section 8.1. Accident conditions as postulated in ANSI/ANS 57.9-1984 due to natural phenomena and manmade events are addressed in Section 8.2. These events provide a means of establishing that the NUHOMS-12T system design satisfies the applicable operational and safety acceptance criteria as delineated herein.

This chapter presents the engineering analysis for normal, off-normal, and accident conditions if the NUHOMS-12T system using the NRC 10 CFR 71 certified MP-187 for transportation from TAN to INTEC. Appendix E of this SAR presents the normal, off-normal and accident conditions of the NUHOMS-12T system using the NRC 10 CFR 72 approved OS-197 Transfer Cask for transportation from TAN to INTEC.

8.1 Normal and Off-Normal Operations Normal operating design conditions consist of a set of events that occur regularly, or frequently, in the course of normal operation of the NUHOMS-12T system. These normal operating conditions are addressed in Section 8.1.1. Off-normal operating design conditions are events that could occur with moderate frequency, possibly once during any calendar year of operation. These off-normal operating conditions are addressed in Section 8.1.2. The thermal-hydraulic, structural, and radiological analyses associated with these events are presented in the sections that follow.

8.1.1 Normal Load Structural Analysis Table 8.1-1 shows the normal operating loads for which the NUHOMS-12T safety-related components are designed. The table also lists the individual NUHOMS-12T components that are affected by each loading. The magnitude and characteristics of each load are described in Section 8.1.1.1.

The method of analysis and the analytical results for each load are described in Sections 8.1.1.2 through 8.1.1.8. The load combinations for the affected components are given in Table 8.1-1 and Table 8.1-2.

The mechanical properties of materials employed in the structural analysis of the NUHOMS-12T system components are presented in Table 8.1-3.

8.1.1.1 Normal Operating Loads. The normal operating loads for the NUHOMS-12T system components are:

1.

Dead Weight Loads

2.

Design Basis Internal Pressure Loads

3.

Design Basis Thermal Loads

4.

Operational Handling Loads

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5.

Design Basis Live Loads These loads are described in detail in the following paragraphs.

A.

Dead Weight Loads Table 8.1-4 shows the weights of various components of the NUHOMS-12T system. The dead weight of the component materials is determined based on nominal component dimensions.

A density value of 0.283 pound per cubic inch for carbon steel and 0.285 pound per cubic inch for stainless steel are used in the dead weight calculations. A nominal concrete density of 140 pounds per cubic foot is conservatively selected as a design basis for the shielding and thermal evaluations. A nominal density of 150 pounds per cubic foot is assumed for the structural evaluation of the HSM.

B.

Design Basis Internal Pressure The range of NUHOMS-12T DSC internal pressures for operating conditions and postulated accident conditions and the associated air temperatures are shown in Table 8.1-5. For normal and off-normal operating conditions, the NUHOMS-12T DSC is vented to the atmosphere and therefore the normal operating pressure is 0 psig.

The maximum fuel debris temperature under normal, off-normal, or accident conditions from Table 8.1-10 is 219F. To be conservative, the maximum fuel debris temperature is used for the air temperature in the DSC. To add additional conservatism, a temperature of 250F is used as the temperature of the air in the DSC for pressure calculation.

The free volume of the DSC is calculated to be 3.4 x 106 cc. From Appendix C, the maximum hydrogen production rate with 12 TMI-2 canisters is 7 x 12 = 84 cc/hr. Using the stoichiometric production rate of hydrogen and oxygen, the total gas production rate in the DSC from 12 TMI-2 canisters is 84 + 84/2 = 126 cc/hr. Assuming that the vents on the DSC are plugged and the DSC is left in the HSM for a one-year period, the total gas production in the DSC will be 1.1 x 106 cc/year.

Using the above temperatures, the DSC free volume and the total gas produced for a period of one year, the maximum pressure in the DSC is calculated to be 11.4 psig for the DSC in the HSM case.

Therefore, a design pressure of 15 psig is used for the DSC which bounds the pressure in the DSC during normal, off-normal, or accident conditions.

C.

Design Basis Thermal Loads The range of ambient temperature cases analyzed are defined in Section 8.1.3. The resulting temperature distributions in the HSM and DSC are determined by performing thermal analyses for these ambient conditions. Thermal properties for the materials are given in Table 8.1-6 and Table 8.1-7. The thermal analyses described in Section 8.1.3 provide temperature distributions for the HSM and DSC. They are shown in Table 8.1-8 through Table 8.1-10, and in Figure 8.1-1 through Figure 8.1-5. These temperature distributions are developed for the range of normal ambient temperatures shown in Table 8.1-10.

Table 8.1-9 summarizes the maximum calculated temperatures for normal operating conditions for the DSC shell. A more detailed tabulation of the thermal results used for the HSM structural design is shown in Table 8.1-8. The HSM reinforced concrete design is controlled by these thermal gradients.

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SAR-II-8.4 CH 8 Rev. 6 8-7 of 8-98 The effect of DSCs being emplaced at varying locations throughout the HSM array are represented by the single module analysis presented herein since the prefabricated HSMs which make up HSM arrays are free-standing units. The appropriate inside and outside surface temperatures corresponding to peak temperatures and thermal gradients from Table 8.1-8 are applied to the HSM analytical model shown in Figure 8.1-15 and the resulting forces and moments calculated.

The temperature distributions derived from the three normal operating cases (-20F, 45F, 87F) are considered in the structural analysis of the DSC and HSM as discussed in Sections 8.1.1.2 through 8.1.1.9. The temperature distributions for each component are used to determine the effects of thermal stresses and thermal cycling on the NUHOMS-12T components. These results are also used to evaluate the effects of creep on the HSM reinforced concrete.

D.

Operational Handling Loads The most significant operational loading condition for the NUHOMS-12T system components is sliding of the DSC from the transfer cask into the HSM. Sliding is achieved by the push/pull forces induced by the hydraulic ram system. These forces are applied to the grapple ring assembly, which is an integral part of the DSC bottom end assembly. The forces induced by the ram system are reacted by friction forces which develop between the sliding surfaces of the DSC, the transfer cask, and HSM support rails.

Based on the surface finish and contact angle of the DSC support rails inside the HSM (as described in Chapter 4), a bounding coefficient of friction is conservatively assumed to be 0.25. Therefore, the nominal ram load required to slide the DSC under normal operating conditions is:

P = 0.25 W Cos

= 0.29 W

Where:

P = Push/Pull Load W = Loaded DSC Weight

= 30 degrees, Angle of the Canister Support Rail The DSC bottom cover plate and grapple ring assembly are designed to withstand a normal operating push/pull force equal to 100% of the loaded DSC weight (conservatively set at 70,000 pounds). To ensure retrievability for a postulated jammed DSC condition, the ram is sized with a capacity for an enveloping load of 70,000 pounds, as described in Section 8.1.2. These ram forces bound those friction forces postulated to develop between the sliding surfaces of the DSC and transfer cask during worst case off-normal conditions.

E.

Design Basis Live Loads As discussed in Section 3.2.4, a live load of 130 pounds per square foot is conservatively selected to envelope all postulated live loads acting on the HSM, including the effects of snow and ice.

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SAR-II-8.4 CH 8 Rev. 6 8-8 of 8-98 8.1.1.2 Dry Shielded Canister Confinement Boundary Analysis. Stresses are evaluated in the NUHOMS DSC shell and closure plates due to:

1.

Dead Weight

2.

Design Basis Normal Operating Internal Pressure Loads

3.

Normal Operating Thermal Loads

4.

Normal Operation Handling Loads The methodology used to evaluate the effects of these normal loads is addressed in the following paragraphs. Table 8.1-11 summarizes the resulting stresses for normal operating loads.

A.

DSC Dead Load Analysis For the dead load analysis, the most limiting conditions are considered. Both the beam behavior and shell bending behavior of the DSC seated on the HSM support rails are evaluated. Conservatively considering the DSC to be supported at the ends, and distributing the DSC bounding weight of 70,000 pounds along its length, the maximum beam bending stresses in the DSC shell are derived. The analysis of the DSC cylindrical shell acting as a simply supported beam results in a maximum membrane stress intensity of 0.2 ksi which is insignificant compared to the allowable of 22.4 ksi.

For evaluation of local shell bending stresses, it is assumed that the total dead weight of the DSC, less the weight of the shield plugs, is uniformly supported by the continuous DSC support rails. The geometric boundary conditions assumed in this analysis are depicted in Figure 8.1-6.

The correlation from Bednar [8.3] is used in the evaluation of this condition as follows:

bx 3

2 S

= 1.75 Rt f t

Reference 8.3, p. 193 Where:

Sbx

=

Local Shell Bending Stress Intensity = 4.8 Ksi t

=

0.55 in., minimum acceptable DSC Shell Thickness R

=

33.62 in., DSC Outside Radius f3

=

191 lb./in., weight of loaded DSC less shielded end plugs per unit length for each support rail The DSC shell component vertical dead weight stresses are calculated from the DSC 75g bottom end drop analysis results discussed in Section 8.2.5. The stress results from the 75g bottom end drop linear elastic static analysis are factored by 1/75 to obtain the DSC shell component vertical dead weight stresses.

The horizontal dead weight stresses in the DSC, shield plugs, closure plates, and shell stresses at the top and bottom ends of the DSC, are calculated by ratioing the results from the 0 orientation DSC 75g horizontal side drop linear elastic static analysis results discussed in Section 8.2.5.2.A by 1/75.

The dead weight stresses in these components while inside the HSM are analyzed using the same analytical model. The boundary conditions are modified to support the DSC shell at the locations of the DSC support rails. A linear elastic analysis is performed for the 1g dead weight load.

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SAR-II-8.4 CH 8 Rev. 6 8-9 of 8-98 The controlling calculated dead weight stresses from the analyses described above are tabulated in Table 8.1-11.

B.

DSC Normal Operating Design Basis Pressure Analysis For evaluation of DSC stresses due to design basis normal operating internal pressures, an analytical model is developed using the ANSYS computer program. The DSC shell, the closure plates, and the shield plugs are included in the analytical model, as shown in Figure 8.1-7. Even though the normal and off-normal operating pressure is 0 psig, a bounding internal pressure of 15 psig is applied to the model as a uniform internal pressure loading and the DSC stresses calculated. Two loading distribution cases are considered; one with the inner closure plates pressurized and one with the outer closure plates pressurized. The resulting maximum stress intensities are reported in Table 8.1-11 and Table 8.1-12.

C.

DSC Normal Operating Thermal Stress Analysis The thermal analysis of the DSC is presented in Section 8.1.3. The results of this analysis show that for the range of normal operating ambient temperature conditions, no significant DSC through wall thermal gradients exist. There is also sufficient space provided in the axial direction between the internal basket assembly and the inner surfaces of the DSC shell assembly for free thermal expansion.

Similarly, sufficient radial gap is provided between the basket assembly and inside of the DSC shell to permit free thermal expansion. As a result, no thermal stresses are induced in the DSC shell or the basket assembly. This design feature also acts to minimize the effects of thermal cycling and fatigue on the DSC.

For normal operating conditions, a thermal stress analysis of the DSC shell is performed to establish the DSC shell stresses induced by variations in shell temperatures. The 87°F ambient case DSC shell temperature varies, as shown in Figure 8.1-5. The maximum DSC shell thermal gradient results from the 103°F ambient off-normal conditions are described in Section 8.1.3. The thermal stresses due to these temperature variations bound those due to the normal operating events and are conservatively used for the normal operating thermal analysis. The thermal bending and membrane stresses for the off-normal case shell gradients are evaluated using the three-dimensional ANSYS model of the DSC shell shown in Figure 8.1-7. The three-dimensional analytical model includes the full-length DSC shell and cover plates with the temperature distribution imposed on the DSC. The analysis results for this case are included in the formulation of normal operating load combination results in Section 8.3.3.

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SAR-II-8.4 CH 8 Rev. 6 8-10 of 8-98 D.

DSC Operational Handling Load Analysis The applied normal operating force from the hydraulic ram, specified in Section 8.1.1.1, is applied to the DSC assembly at the grapple ring location. The three-dimensional ANSYS finite element model shown in Figure 8.1-7 is used to calculate the stresses in the DSC shell assembly. A uniform pressure load is applied to the center of the DSC bottom cover plate over an area approximately equal the contact area of the ram grapple assembly to evaluate the DSC insertion loading condition. A uniform pressure load is applied to the inner surface of the DSC grapple ring plate to evaluate the DSC retrieval loading condition.

In addition to the uniformly distributed load assumption, the DSC bottom cover plate and grapple ring stresses for the DSC retrieval non-uniform load distribution of the ram grapple are analyzed using the 1/4 symmetry ANSYS model shown in Figure 8.1-8. The load is applied to the grapple ring plate nodes corresponding to the contact area between the ram grapple arms and the grapple ring plate. The edges of the DSC bottom cover plate are conservatively modeled as pinned. The analysis results show that the load transferred to the DSC shell is uniformly distributed by the bottom cover plate. Therefore, the results obtained using the three-dimensional model with a uniform pressure load applied to the grapple ring are valid for all the DSC shell components except the bottom cover plate and the grapple ring plate for the DSC retrieval condition.

The maximum stress intensities from these analyses are tabulated in Table 8.1-11.

E.

Evaluation of the Results The maximum calculated DSC shell stress intensities for the normal operating load conditions are shown in Table 8.1-11. The calculated stress intensities for each load case are combined in accordance with the load combinations described in Table 3.2-5. The resulting stress intensities for the controlling load combinations are reported in Section 8.3 and compared to the ASME Code allowable stress intensities.

8.1.1.3 DSC Internal Basket Analysis. The DSC internal basket assembly is a non-structural component, which is classified as not important to safety. As discussed in Section 3.3, it is not required to maintain subcriticality of the stored TMI-2 canisters. As such, the basket structure is not required to meet any structural limits. For information purposes only, the results of an analysis for an 8g dead load case, simulating worst case transfer loads, is provided to demonstrate the level of stresses in the basket and show it has no adverse effect on the DSC shell. The applied loading is shown in Figure 8.1-10. Differential thermal expansion effects are negligibly small due to the very small temperature gradients and the geometry of the basket assembly.

A.

DSC Internal Basket Dead Weight Analysis The basket dead weight stresses are calculated for the horizontal storage orientation inside the HSM.

The basket dead load stress intensities are calculated using the ANSYS spacer disc analytical model shown in Figure 8.1-9. The model boundary conditions are modified to reflect the support conditions provided by the transfer cask and DSC support rails. Linear elastic static analyses are performed. The applied loading is shown in Figure 8.1-10.

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SAR-II-8.4 CH 8 Rev. 6 8-11 of 8-98 B.

DSC Internal Basket Thermal Stress Analysis The effects of axial and radial thermal expansion are evaluated for the DSC internal basket components to ensure that adequate space exists in the cavity of the DSC for free thermal expansion.

To verify that adequate provision for free axial expansion of the TMI-2 canisters and other internal components of the basket are included, the differential expansion of each DSC component is calculated. The TMI-2 canisters are assumed to be at their maximum calculated accident temperature (Table 8.1-10) of 219°F (104°C) and the DSC shell is conservatively assumed to be at its long-term average normal operating temperature (45°F ambient air) of 95°F (Table 8.1-9). Calculated results for varying ambient temperatures show similar results as the T between the TMI-2 canisters and DSC does not change appreciably. The length of the TMI-2 canister when hot is:

LHC = (1 + sT)LT Where for the design basis TMI-2 canister:

LHC

=

Hot length of TMI-2 canister, in.

s

=

Stainless steel coefficient of thermal expansion is 9.08 x10-6 in./in.°F at 200°F T

=

219° - 70° (stress-free temperature) = 149°F LT

=

Length of TMI-2 canister at room temperature (70F)

=

150 in.

Therefore:

LHC

=

150.20 in.

The length of the DSC cavity at room temperature is 151.00 inches. The minimum length of the DSC cavity (LC) at 95F is:

LCH = LCCT + LC Where:

LC

=

151.00 in., DSC cavity at 70F T

=

95 - 70 = 25F C

=

5.42 x 10-6 in./in. F at 70F (SA 516, Gr. 70 Plate)

Therefore:

LCH

=

151.02 in.

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SAR-II-8.4 CH 8 Rev. 6 8-12 of 8-98 The minimum axial clearance between the end of the TMI-2 canisters and the inner surface of the shield plug assembly for the long-term base case ambient conditions (45°F) and the associated normal operating spent fuel temperature is 0.83 inch. Similarly, assuming that each spacer disc is uniformly heated to the maximum core debris temperature of 219°F and the DSC shell remains at 95°F, the radial gap will be reduced from 0.19 inches to 0.17 inches. Thus, adequate clearance has been provided between the TMI-2 canisters, the DSC shell, and internal basket assembly to permit free thermal expansion of all components for the maximum differential temperatures expected.

8.1.1.4 DSC Support Structure Analysis. The general description of the DSC support structure inside the HSM is provided in previous sections. The DSC support structure is shown in Figures 4.2-1 and 4.2-2. The DSC support rails are supported by the HSM front and rear walls. The DSC support structure design uses welded connections and allows thermal growth relative to the HSM. Normal operating condition loads on the DSC support structure consist of the DSC dead weight, the support structure dead weight, and the DSC operational handling loads. The normal operating frictional loads are reacted by the support rails, which are bolted to embedments located in the HSM front wall. Because of the dimensions of the HSM internal cavity, the DSC support rails only carry the DSC dead load during insertion and withdrawal operations. During storage conditions, the DSC closure plates transfer the loads to the front and back HSM walls while the DSC shell acts as a rigid beam.

A linear elastic model of the assembly is utilized to evaluate axial, shear, and bending stresses in the DSC support structure members. The geometry for the analytical model is shown in Figure 8.1-14.

A.

DSC Support Structure Dead Weight Analysis For the dead weight analysis, the weight of the DSC and contents is conservatively applied to the support rails as a uniformly distributed load. The dead weight of the DSC support structure is also included in this analysis.

B.

DSC Support Structure Operational Handling Analysis For the NUHOMS-12T system operational handling loads, a sliding force of 70,000 pounds is applied axially to the DSC support rails (35,000 lb/rail) to account for the sliding friction between the DSC shell and the support rails. This force is combined with the worst case of the normal condition loading of a 40,000 lb. (20,000 lbs/rail) point load or a uniformly distributed load representing the weight of the DSC and contents.

C.

DSC Support Structure Thermal Analysis The DSC support structure is a simple frame structure that permits free thermal expansion relative to the HSM. The connection details permit free thermal expansion of the DSC support structure relative to the HSM, eliminating any significant thermal stresses.

D.

Evaluation of DSC Support Structure Results The maximum calculated DSC support structure deflections for normal and off-normal operating loads are shown in Table 8.1-14. Specific information on the DSC support structure seismic loads can be found in Section 8.2.3. Additional details of the DSC support structure analysis are presented in Section 8.3.

8.1.1.5 HSM Loads Analysis. To qualify the design for the TMI-2 ISFSI, a single free-standing HSM is evaluated. The HSM structural analyses include evaluation of normal operating, off-normal, and postulated accident loads for the HSM. The frame and shear wall action of the HSM floor, walls, and roof slab are the

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SAR-II-8.4 CH 8 Rev. 6 8-13 of 8-98 primary structural system for transverse loads. The structural analysis of an individual module provides a conservative methodology for evaluating the response of the HSM structural elements under various static and dynamic loads for any HSM array configured in accordance with Section 4.1.1.

The HSM side walls and roof slab thicknesses are established on the basis of radiological shielding requirements. As such, all other thickness requirements such as the minimum barrier thickness requirements for tornado generated missiles specified by NUREG-0800, Section 3.5.3 [8.4] are bounded. The ultimate strength method is used to evaluate stresses in the HSM reinforced concrete walls, roof, and floor. The HSM reinforcement is designed to meet the minimum flexural and shear reinforcement requirements of ACI 349

[8.5]. The available design strength exceeds that required for the factored design loads specified in Section 3.2. The reinforcement layout for the prefabricated HSMs is shown on Figure 8.1-16 and the Appendix A licensing drawings. HSM construction details such as construction joints and reinforcing bar splices will be detailed on the construction drawings.

For all normal, off-normal, and environmental load cases except the thermal load cases, the reinforced concrete free-standing HSM module is evaluated using the analytical model shown in Figure 8.1-13. The DSC support structure model shown in Figure 8.1-14 is included in the model. The concrete structure is modeled using four layers of three-dimensional brick elements. The reinforced concrete pocket on the south wall is modeled using three layers of brick elements. The composite steel and concrete door on the north wall is modeled using brick elements with equivalent properties. The DSC and the support structure are modeled using beam elements and lumped mass elements. The three-dimensional ANSYS finite element model has 2991 nodes and 1517 brick elements. The load inputs for this analysis are described in Sections 8.2.2 through 8.2.4. The various normal operating mechanical loads are applied to the analytical model and the HSM internal forces and moments calculated by performing a linear elastic finite element analysis.

For the thermal load analysis, the three-dimensional model shown in Figure 8.1-15 is used. This analytical model does not include the DSC support structure or the HSM door. The model utilizes non-linear three-dimensional brick elements with concrete cracking capability. This allows modeling the self-relieving nature of thermal stress due to cracking. The analysis shows that a single HSM provides the governing case for load combinations containing tornado wind and missile loads and seismic load conditions. The postulated response investigated for each of these cases is the potential for sliding or overturning of a single free-standing HSM which envelopes that of an HSM array. The analysis also shows that thermal loads control the reinforcement requirements for the HSM walls, roof and floor. The load inputs for this analysis are described in Section 8.1.1.5, Paragraph C, Section 8.1.2.2, Paragraph A, and Section 8.2.10.2.

A description of the individual loads and load analyses are provided in the following sections.

A.

HSM Dead Load and Live Load Analyses The dead weight of the HSM plus the loaded DSC and the DSC support structure weights are applied to the analytical model shown in Figure 8.1-13. The loads applied to the analytical model include the dead weight of the loaded DSC and DSC support structure. The resulting calculated maximum dead load shears and moments are tabulated in Table 8.1-15. A live load of 130 psf is applied to the HSM roof to conservatively envelope all postulated live loads, including snow and ice. The resulting calculated maximum live load shears and moments are tabulated in Table 8.1-15.

B.

Concrete Creep and Shrinkage Analyses ACI 349 Section 9.2.2 states that where the effects of... creep or shrinkage may be significant, they shall be included with the dead load.... Since creep is mainly dependent on elastic strain due to dead loads, the loading contribution from creep is minimal as the dead loads are small in relation to the

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SAR-II-8.4 CH 8 Rev. 6 8-14 of 8-98 capacity of the HSM frame and shear walls formed by the walls, roof and floor. The creep strain is conservatively calculated using the ultimate creep strain value suggested by Wang and Salmon [8.6].

c =

u C c Where:

'c

=

Creep strain in./in.

Cu

=

Ultimate creep strain or ratio of creep to initial strain from dead weight

=

2.35 c

=

Initial strain from dead weight

=

3E-5 in./in. (approximate value)

Therefore:

'c

=

- in./in.

Shrinkage of the HSM concrete is conservatively calculated using an ultimate shrinkage strain value suggested by Wang and Salmon [8.6]. Additionally, since shrinkage is significantly affected by the surface area to volume ratio, the ultimate shrinkage strain value is reduced according to the method recommended by Fintel [8.7]. The combined shrinkage strain is:

s =

sh

Where:

s

=

Shrinkage strain (in./in.)

sh

=

Ultimate shrinkage strain = 8E-4 in./in.

=

Volume to surface area reduction = 0.5 (conservative)

Therefore:

s

=

4E-4 in./in.

For determination of moments and shears in the HSM due to creep and shrinkage, the total strain is converted to an axial change in length across the roof of a prefabricated HSM.

L = L (

+

)

s

c Where:

L

=

Axial length change (in.)

L

=

Length from center to center of HSM walls

=

99 in.

s

=

4E-4 in./in.

'c

=

4.65E-5 in./in.

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SAR-II-8.4 CH 8 Rev. 6 8-15 of 8-98 The resulting change in length of the HSM roof slab, L = 0.044 inch.

Shortening due to creep and shrinkage occurs gradually over a period of time, and the effects are lessened by plastic creep flow and microcracking of the members. Ambient humidity also acts to reduce the effects of creep and shrinkage. The PCI Design Handbook [8.8] suggests that the calculated creep and shrinkage shortening values be reduced by a factor of three to five for design. Also, creep and shrinkage forces act opposite to those of thermal expansion forces for the HSM. Hence, it is conservative to neglect creep and shrinkage effects. Therefore, creep and shrinkage are not considered further for the HSM design.

C.

HSM Thermal Loads Analysis To evaluate the effects of thermal loads on the HSM, heat transfer analyses for a range of normal ambient temperatures are performed and the limiting thermal gradients and temperature values at various locations in the HSM determined. A more detailed description of the heat transfer analyses is provided in Section 8.1.3. Structural analyses of the HSM for the maximum calculated floor, wall and roof temperature loads listed in Table 8.1-8 and Table 8.1-9 are performed for the HSM using the analytical models shown in Figure 8.1-15. The results of these analyses are summarized in Table 8.1-15. The basis for the HSM thermal analysis is discussed further in the following paragraphs.

ACI 349, Appendix A, provides a general methodology for designing reinforced concrete structures subjected to thermal loads. The commentary to this Appendix, Section A.3.3, defines a range of approaches utilized in the analysis of thermal loads. One approach accounts for the self-relieving nature of the thermal loads (relief is obtained by the occurrence of thermal cracking when the concrete modulus of rupture is reached). For the thermal analysis of the HSM for normal operating conditions, the thermal loads are calculated for the cracked cross section properties of the HSM walls, roof and floor.

To account for the seasonal effects of ambient temperature fluctuations on the outside surface of the HSM, ambient temperatures of -20F (winter) to 87F (summer) are considered in the heat transfer analysis for normal operating conditions. Analyses are performed for ambient temperatures of 45F and 87F to determine the limiting design conditions for the HSM. For the HSM roof slab, the results of the HSM heat transfer analysis for normal operating conditions for a lifetime average ambient temperature of 45F and with solar heat flux effects included are shown in Figure 8.1-3. The maximum temperature and gradients for the roof slab and side walls are shown in Table 8.1-8. The results demonstrate that the concrete temperatures are within the ACI 349, Section A.4.1 acceptance temperature for local areas of 200F.

For conservatism and consistency with the philosophy of ACI 349, Section A.4.3, the strength properties of the concrete and reinforcing steel used in the HSM structural analysis are taken at the postulated temperature range for each load case. For all normal operating load cases the concrete and reinforcing properties are assumed to be equal to the specified values (f'c = 5000 psi for concrete and fy = 60,000 psi for rebar).

Table 8.1-8 shows that the maximum HSM temperature for the lifetime average ambient temperature of 45F does not exceed the ACI limit of 200F for local area temperatures.

The design criteria utilized are adequate to ensure that the NUHOMS-12T HSM will perform its intended safety function for all design conditions.

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SAR-II-8.4 CH 8 Rev. 6 8-16 of 8-98 D.

Radiation Effects on HSM Concrete As described in Reference 8.9, the accumulated neutron flux over the 50-year service life of the HSM for five year cooled intact fuel (which envelopes the TMI-2 fuel) is estimated to be 1.7E14 neutrons/cm2. From the study by Hilsdorf, Kropp, and Koch [8.10], the compressive strength and modulus of elasticity of concrete is not affected by a neutron flux of this magnitude.

As described in Reference 8.9, the gamma energy flux deposited in the HSM concrete for five year cooled intact fuel (which envelopes the TMI-2 fuel) is 1.7 E9 MeV/cm2-sec. or 3.0 E-4 watt/cm2.

According to ANSI/ANS-6.4-1977 [8.11], the temperature rise in concrete due to this level of radiation is negligible. Thus, radiation effects on concrete strength are not evaluated further for the HSM design.

E.

HSM Design Analysis The mechanical design loads (dead load, live load, wind, etc.) were applied to the analytical model shown in Figure 8.1-13, and the thermal loads were applied to the analytical model shown in Figure 8.2-15. In both cases, the design forces and moments are calculated. The results of these analyses are summarized in Table 8.1-15 for the normal operating and off-normal loads.

The ultimate axial force, moment and shear capacities of a typical 12-inch wide section of the HSM walls, roof, or floor are computed using the following relationships:

Ultimate Tension Capacity:

Ptu = Pt = Astfy Ultimate Compression Capacity:

Pcu = Pn = 0.8[0.85fc(Ag-Asc)+fyAsc]

Ultimate Moment Capacity:

Mu = Mn = As fy (d-a/2) where a = (As fy)/(0.85fc b)

Ultimate In-Plane Shear Capacity: Vui = Vi = Asc(2fc + nfy)

Ultimate Out-Plane Shear Capacity: Vuo = Vc = 2 fc (bd)

The thicker front and rear wall sections qualify as deep flexure members and the allowable out-of plane shear capacity may be calculated in accordance with Section 11.8.6 of ACI using the formula:

Vuo = *Vc = (3.5-2.5 Mu/Vu d) (1.9fc + 2500 wVud / Mu ) b d but not to exceed 6 fc b d and (3.5-2.5 Mu/Vu d) not to exceed 2.5 Vu and Mu are shear force and moments at factored loads Where:

= Strength reduction factor (Table 3.2-3) fy

= 60,000 psi, Rebar design strength at 150F fc

= 5,000 psi, Design compressive strength of concrete at 150F As

= Area of tensile reinforcing steel (see Appendix A drawings) b

= 12 width section d

= depth of the section minus rebar cover (see Appendix A drawings)

Ast

= Area of axial tension reinforcement = 2*As Asc

= Area of axial compression reinforcement = 2*As

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SAR-II-8.4 CH 8 Rev. 6 8-17 of 8-98 Ag

= Gross area of concrete = (b*T) n

= (2*As)/(b*T)

T

= Depth of the section including the rebar cover w

= As/(bd)

The computed ultimate capacities are tabulated below:

Component Ptu kips/ft Pcu kips/ft Mu kip-in/ft Vui kips/ft Vuo kips/ft Roof 128 817 1427 159 33 Floor 102 745 1046 131 31 Side Walls 102 541 1046 131 31 Front Wall 128 931 1683 164 101 Back Wall 161 1121 2576 203 145 8.1.1.6 HSM Door Analyses. The access opening for transferring the DSC into and out of the HSM is protected by a steel door with a core of concrete shielding material. The door is recessed into the HSM front wall as shown in Figure 4.2-3.

For normal system operation, the door assembly is only subjected to its own dead weight which is transmitted by bearing directly into the HSM front wall, and handling loads resulting from installation and removal of the door during DSC transfer operations.

The concrete bearing strength required to support a bearing load on the front wall concrete for a weight of 5900 pounds is a small fraction of the ACI 349 (Section 10.15) permissible concrete bearing strength of [0.85 fc'A1] = 0.7 [0.85 x 5 x 6 x 80.63] = 1440 kips. The embedded anchors for the HSM door are designed in accordance with ACI 349, Appendix B. The governing design load combination for the HSM door embedded anchors is the dead load plus seismic load combination. The dead load consists of the weight of the door. The seismic load consists of the longitudinal seismic accelerations acting on the door itself (the DSC seismic loads are transferred to the DSC support structure rails.) The seismic loads produce very small shear and tension loads in the anchors.

8.1.1.7 DSC Seismic Retainer. The details of the DSC seismic retainer are shown on the drawings in Appendix A. Additional details are provided in Section 8.2.3.2.

8.1.1.8 MP187 Cask Analysis. Any NRC-approved transfer cask may be utilized to transfer a loaded DSC to the TMI-2 ISFSI provided it meets the requirements for size, weight, and radiological protection required for the NUHOMS-12T DSCs.

The MP187 cask when used in the 10 CFR Part 72 on-site transfer mode, has been fully evaluated for normal operating condition loads which meet or exceed the loads expected at INL, and the results reported in the SMUD SAR [8.12]. These evaluations have demonstrated that the MP187 cask has adequate margins to safely perform all required operations for the transfer of a loaded DSC. As the MP187 is fabricated from austenitic stainless steel, exposure to the -20°F minimum normal operating temperature will not have any effect on the casks ability to protect the NUHOMS-12T DSCs.

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SAR-II-8.4 CH 8 Rev. 6 8-18 of 8-98 8.1.2 Off-Normal Load Structural Analysis Table 8.1-12 shows the off-normal operating loads for which the NUHOMS-12T safety-related components are designed. This section describes the design basis off-normal events for the NUHOMS-12T system and presents analyses that demonstrate the adequacy of the design safety features of a NUHOMS-12T system.

8.1.2.1 Jammed DSC During Transfer. The interfacing dimensions of the top end of the MP187 cask and the HSM access opening sleeve are specified so that docking of the cask with the HSM is not possible should gross misalignments between the cask and HSM exist. Furthermore, beveled lead-ins are provided on the ends of the cask, DSC, and DSC support rails to minimize the possibility of a jammed DSC during transfer. Nevertheless, it is postulated that if the cask is not accurately aligned with respect to the HSM, the DSC binds or becomes jammed during transfer operations. Based on the dimensions of the DSC, cask, and HSM, the maximum misalignment of the sliding surfaces is limited by operating procedures to 1/8 inch or less. Assuming a worst-case misalignment in positioning and docking the cask with the HSM access opening, the maximum possible misalignment which would permit transfer of the DSC to occur is 1/4 inch. Although unlikely, any greater misalignment may cause axial sticking and/or a rotation of the DSC to occur which may result in a binding condition.

A.

Detection of the Event When a jam of the DSC occurs during transfer the hydraulic pressure in the ram begins to increase.

The maximum normal operating ram push/pull forces are limited automatically by features in the ram system design to a maximum load equal to 70,000 lbs.

B.

Axial Sticking of the DSC The off-normal handling load event postulated to occur assumes that the leading edge of the DSC becomes jammed against some immovable feature because of gross misalignment of the cask. As the DSC motion is prevented, the hydraulic pressure increases, thereby increasing the force on the DSC until the system pressure limit is reached. To overcome potentially higher resistance loads due to binding of the DSC in either the cask or the HSM, the 70,000 lb. maximum ram force exceeds the weight of the loaded DSC. This force corresponds to a coefficient of friction of one and is the design basis for the hydraulic ram system. This postulated loading condition is illustrated in Figure 8.1-11.

The resulting ram load acting on the DSC grapple ring assembly and bottom cover plate are analyzed in the paragraphs which follow.

The DSC bottom cover plate and the grapple ring assembly are subjected to a maximum force of 70,000 pounds. The method of analysis is as described in Sections 8.1.1.1 and 8.1.1.2. The maximum bending stress intensity calculated for the DSC bottom cover plate, as listed in Table 8.1-11 is well within the ASME Code allowable limit.

It is conservatively assumed that the force created by the jammed DSC condition produces a force-couple of magnitude F x R, where: F is the imposed force of 70,000 pounds and R is 33.63 inches, the outside radius of the DSC shell. Thus, the DSC shell stress is:

Smx

= M S 1 Where:

M

=

2354 in.-kip, Bending moment

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SAR-II-8.4 CH 8 Rev. 6 8-19 of 8-98 S

=

2165 in.3, DSC section modulus Therefore:

Smx

=

1.09 ksi This magnitude of stress is negligible when compared to the allowable membrane stress intensity.

For a jammed DSC, the 70,000-pound ram load is postulated to be reacted by the DSC support rails inside the HSM at the most critical location. At the same time, a concentrated force of one half of the DSC weight is assumed to act vertically at mid-span of the DSC support rail member. The results of this analysis are reported in Table 8.1-13.

C.

Binding of the DSC If axial alignment within system operating specifications is not achieved, it may be possible to pinch the DSC shell as shown in Figure 8.1-12. The pinching force acting on the DSC shell and the cask inner liner is directly proportional to the angle of rotation. The maximum possible inclination angle established by various conservative geometric and operational assumptions is less than one degree. If this angle is conservatively assumed to be one degree, then the pinching force is taken as the product of the maximum ram loading of 70,000 pounds and the sine of the angle, or 1,225 pounds. This force is distributed around the circumference of the DSC shell and either the cask inner shell or HSM sleeve as a cosine distribution.

The 1,225-pound load is conservatively assumed to be applied as a point load at a location away from the ends of the cask or DSC. The resulting maximum stresses are given by Table 30, Case 8a of Roark

[8.13] as:

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SAR-II-8.4 CH 8 Rev. 6 8-20 of 8-98 Membrane stress:

= 0.4P t2 Bending stress:

=.4P t2 2

Therefore, the maximum stress is:

+

=. P t2 2 8 For the DSC shell, tmin = 0.55 inch. Substituting for t and P equal to 1,225 pounds, the maximum extreme fiber stresses in the DSC shell is 11.3 ksi. This is well within the ASME Code Service Level C allowable of 34.6 ksi for an off-normal jammed DSC event or the normal or Service Level A allowable of 23.1 ksi.

The tangential component of ram loading under the assumed condition is less than the force of the jammed condition calculated previously in Paragraph B and is not considered further. The stresses on the DSC are given in Table 8.1-12, and for the DSC support assembly inside the HSM for the jammed condition are reported in Table 8.1-13.

D.

Consequences of Jammed DSC In both scenarios for a jammed DSC, the DSC shell stress is much less than the material yield strength and ASME Code allowables. Therefore, no permanent deformation of the DSC shell or cask inner liner will occur and there is no potential for breach of the DSC confinement boundary or potential for release of radioactive material.

E.

Corrective Action In both cases, the required corrective action is to reverse the direction of the ram force being applied to the DSC and return the DSC to its previous position. Since no permanent deformation of the DSC or MP187 cask inner shell occurs, return of the DSC to its previous position is unimpeded. The cask alignment is then rechecked, and the cask repositioned as necessary before attempts at transfer are renewed.

8.1.2.2 Off-Normal Thermal Load Analysis. Off-normal ambient temperatures of -50°F (extreme winter) and 103°F (extreme summer) are extremes for the INL site. Handling or transporting a DSC containing TMI-2 Canisters will not be performed when DSC temperature is less than 20F or when the ambient air temperature is less than 0F. The MP187 is designed for operation at -20F. The system components affected by the postulated extreme ambient temperatures are the cask and DSC during transfer operations at the ISFSI site, and the HSM during storage of a DSC.

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SAR-II-8.4 CH 8 Rev. 6 8-21 of 8-98 The thermal stresses in the various NUHOMS-12T system components due to the off-normal temperatures are calculated in the same manner as described for the normal operating thermal loads. A description of these methods is provided in Sections 8.1.1.2 for the DSC shell and 8.1.1.3 for the DSC internals.

A.

HSM Off-Normal Thermal Load Analysis As described in Section 8.1.3.1, the HSM steady state temperatures are calculated for the off-normal extreme ambient temperature of 103°F. The resulting maximum HSM concrete temperatures and gradients for the roof and side walls are shown in Table 8.1-8. The short duration peak concrete temperatures easily meet the ACI 349 limit of 350°F. Since the ultimate shear and moment capacities for each module concrete section are assumed to be the same for off-normal and accident conditions, the maximum calculated shear forces and bending moments resulting from either case are conservatively used for design. The DSC support assembly is free to expand/contract with changing temperatures and, therefore, the changes in temperature have minimal effect on the DSC support structure.

B.

DSC Off-Normal Thermal Stress Analysis The off-normal thermal gradients and maximum temperatures for -50°F and 103°F ambient air are developed for the DSC resting in the cavity of the MP187 cask as described in Section 8.1.3.3. The maximum off-normal calculated surface temperature for the DSC shell is given in Table 8.1-9.

The off-normal thermal gradients and maximum temperatures are also developed for the DSC resting in the HSM, as described in Section 8.1.3.2. The maximum off-normal surface temperature calculated for the DSC shell is given in Table 8.1-9.

8.1.3 Thermal Hydraulic Analysis This section of the SAR describes the thermal analysis of the NUHOMS-12T HSM and DSC. The analytical models of the HSM and the DSC are described, and the calculation results summarized. The thermophysical properties of the NUHOMS-12T system components used in the thermal analysis are listed in Table 8.1-6 and Table 8.1-7. The following evaluations are performed for the NUHOMS-12T system:

1.

Thermal Analysis of the HSM

2.

Thermal Analysis of the DSC in the HSM

3.

Thermal Analysis of the DSC in the Transfer Cask The NUHOMS-12T components are evaluated for a range of design basis ambient temperatures as follows:

A.

Normal Operating Conditions The system components are evaluated for ambient temperatures in the range of -20°F to 87°F. The normal operating seasonal daily ambient temperatures fluctuate from a low of -20F (winter) to a high of 87°F (summer) [8.14]. The thermal analyses are carried out for a sufficient duration to establish steady state conditions in the NUHOMS-12T components. For the evaluation of thermal cycling and material properties, fluctuations in the ambient temperature from winter to summer conditions are assumed to occur once per year for the HSM. The lifetime average ambient temperature for the 50-year service life is taken as 45°F. The stress-free temperature for material properties is assumed to be 70°F.

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SAR-II-8.4 CH 8 Rev. 6 8-22 of 8-98 B.

Off-Normal and Accident Conditions The system components are evaluated for the extreme ambient temperatures of -50F (winter day) and 103F (summer day) [8.14]. Should these extreme conditions ever occur, they would be expected to last for a short duration of time (a few hours per day for 2 to 3 days in a row). Nevertheless, the thermal analyses are performed for a sufficient duration to cause a steady-state temperature distribution in the NUHOMS-12T system components.

8.1.3.1 Thermal Hydraulics of the HSM A.

Principles of HSM Cooling System The HSM is cooled by radiation, natural convection, and heat conduction through the HSM ceiling and walls. The HSM roof and side walls are the primary concrete surfaces conducting heat to the outside environment. For analytical purposes, an HSM centered in a group of HSMs, each loaded with a DSC, is assumed. For the thermal analyses of an HSM at the end of a multiple module array, an end shield wall is added to the side wall and the outer surface is exposed to the prevailing ambient conditions.

The ISFSI basemat is in contact with soil, which is assumed to be at a constant temperature of 65F (summer) and 45F (winter) at a combined depth of 10.5 feet [8.14]. Air infiltration from the HSM access opening door and HEPA filter door is conservatively neglected.

In the HSM HEATING 7 model, Boundary Type 1 (surface-to-boundary) is used to describe the natural circulation heat transfer between the HSM outer surfaces and the adjacent ambient air. The DSC temperatures from the HSM models are used as constant temperature boundary conditions to determine the temperature distributions for the DSC internals and the TMI-2 canister regions. These calculations are described in the following paragraphs.

B.

Computer Program The HEATING 7 computer program is used for the heat transfer analysis of the HSM and DSC. The HEATING 7 program is known as The HEATING Program, where HEATING is an acronym for Heat Engineering and Transfer In Nine Geometries. HEATING 7 is designed to be a functional module within the SCALE system of computer programs [8.15] for performing standardized computer analysis for licensing evaluations of nuclear spent fuel systems. Thus, its features are designed to perform thermal analyses on problems arising in licensing evaluations of shipping and storage of spent fuel containers, and its input format is designed to be compatible with that of other functional modules within the SCALE system. HEATING 7 is used in the previous NUHOMS design for thermal analysis of the DSC, HSM, and casks that were reviewed and approved by the USNRC for 10 CFR Part 71 and 10 CFR Part 72 requirements. HEATING 7 may also be used as a stand-alone heat transfer code. The SCALE system was developed by Oak Ridge National Laboratory for the U. S. NRC Transportation Branch.

C.

Thermal Model of the HSM The HEATING 7 thermal model of the HSM is depicted in Figure 8.1-1. The three-dimensional model represents the symmetric right half of an HSM and DSC. As shown in Table 3.1-1, the design basis heat load of 860 W/DSC is used for the NUHOMS-12T HSM design. This heat load is applied as a volumetric heat density of 9.60E-5 Btu/min-in3 over the fueled portion of the TMI-2 canister. The heat rejection through the TMI-2 canister ends and the DSC shield plugs and cover plates are included.

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SAR-II-8.4 CH 8 Rev. 6 8-23 of 8-98 The HEATING 7 analytical model of the HSM consists of a typical HSM cross-section. In addition, symmetry or an insulated boundary is assumed along the vertical centerline of the HSM shown in Figure 8.1-1. The HEATING 7 model includes regions for the concrete front, rear, and side walls, roof slab, floor, and basemat of the HSM. The soil below the ISFSI basemat is modeled as a 7-ft thick region with a constant temperature boundary at the edge of this region. Sufficient nodal refinement is used in the HSM analytical model to obtain accurate temperature distributions through the thickness of the HSM walls, roof, and floor slab.

The thermal analysis of a typical HSM is performed for a loaded DSC located in the interior of a multiple module array with a DSC present in the two adjacent HSMs. The HSM roof and side walls are modeled, the outer surfaces of which are assumed to be exposed to the prevailing ambient conditions.

For summer ambient conditions, a solar heat flux of 67 Btu/hr.-ft.2 for normal conditions, and 105 Btu/hr.-ft.2 for off-normal and accident conditions, is conservatively applied to the roof surface [8.14].

Solar heat loads are conservatively neglected for the HSM thermal analysis for winter ambient conditions for normal, off-normal and accident conditions.

The DSC cylindrical shell is approximated in the model by equivalent rectangular regions with the thickness and properties of the carbon steel DSC shell. The approximation using rectangular regions is necessary since HEATING 7 restricts the user to one geometry type in the same analytical model. The analytical model also includes regions to model the air gaps between the DSC and the HSM where heat transfer is mainly due to radiation.

The outer surface of the DSC shell dissipates heat to the HSM through radiation. The air surrounding the DSC is modeled as a gap filled with gas (air) with radiation heat transfer only, thus providing a mechanism for heat transfer from all HSM interior surfaces and the DSC outer surface. Any closed cavity convection occurring in the HSM cavity air surrounding the DSC shell is conservatively neglected.

With these temperatures and the equations for the heat transfer coefficients described below, the HEATING 7 program calculates the temperatures of the DSC exterior surface and the HSM interior and exterior surfaces.

The fuel debris decay heat is removed from the DSC outer surface through radiation. Horizontal surfaces with convection on their upper surfaces, such as the HSM roof outer surface, are assumed to be cooled by natural convection with a heat transfer coefficient of hplate.

The HSM concrete walls are assumed to be cooled by air with a heat transfer coefficient of hwall.

Radiation heat transfer is modeled between the DSC outer surface and the HSM concrete walls and ceiling, and between the DSC outer surface and the HSM floor. The external surface of the HSM roof is assumed to be cooled by external air with a heat transfer coefficient of hplate, and by radiation heat transfer to ambient air. The formulas used for the calculation of the heat transfer coefficients for natural convection are as follows [all in Btu/(hr. ft.2 F)] [8.16]:

Hplate

= 0.22 (T)1/3 Hwall

= 0.19 (T)1/3 The heat transfer coefficients are updated by HEATING 7 following each iteration using the resulting average temperature of the corresponding surface node. A sufficient number of iterations are performed until the temperatures differ by less than 0.001% from the previous temperature calculated in two consecutive iterations indicating that stable convergence is achieved. The effective thermal conductivity of LICON (low density concrete) material, which is a part of the TMI-2 fuel debris canister is calculated for the material composition of 11 weight percent (w/o) glass bubbles, 60 w/o

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SAR-II-8.4 CH 8 Rev. 6 8-24 of 8-98 cement and 29 w/o water. Note that small differences in the weight percentage of the LICON constituents has negligible impact on the thermal analysis results of the core debris and canisters. The effective thermal conductivity of LICON is calculated to be 0.726 Btu/hr.-ft.F. The remaining thermal-hydraulic parameters used in the HSM heat transfer calculations are given in Table 8.1-6 and Table 8.1-7.

The results of the HEATING 7 analysis for the HSM are in the form of temperature distribution profiles. Figure 8.1-2 shows an example of the HEATING 7 results for the HSM. The resulting temperature profile shows the steady state temperature distribution on the outer surface of the DSC, and at various locations throughout the HSM.

The calculated HSM wall and roof temperature gradients are used in the reinforced concrete structural analysis for long term thermal loads which occur during normal operating conditions, and the short-term thermal loads occurring during off-normal and postulated accident conditions. The HSM thermal analysis results are also used to obtain steady state temperature distributions for the outer surface of the DSC for the range of design basis ambient conditions. These steady state surface temperatures are used as a constant temperature boundary condition for the DSC model, described in Section 8.1.3.2.

D.

Description of the Cases Evaluated for the HSM The HSM thermal analyses are performed for the design basis ambient air temperatures. These include a total of three cases with ambient air at the following temperatures:

1.

45F (lifetime average), 87F (summer daily high), and -20F (winter daily low) for normal operating conditions which can be expected to occur for long periods of time,

2.

103F (extreme summer maximum) and -50F (extreme winter minimum) for off-normal conditions which can be expected to occur for short periods of time (a few hours per day for 2 to 3 days at a time), and

3.

103F daily high and -50F daily low extreme ambient temperatures with the gap between the modules blocked, and DSC HEPA filter vents blocked for sufficient duration to reach steady state conditions. This design basis condition is designated as an accident condition assumed to occur once in the service life of an HSM.

The results of these calculations are summarized in Table 8.1-8.

8.1.3.1 Thermal Analyses of the DSC Inside the HSM. For the DSC thermal analyses, the internal basket assembly of the DSC is modeled in detail. A worst case, two-dimensional slice of the DSC and TMI-2 canister cross section is modeled. Heat transfer effects along the axis of the DSC (third dimension) are conservatively neglected. The DSC is assumed to be cooled through radiation with the DSC shell surface specified as a constant temperature boundary condition equal to that calculated in the HSM thermal analysis.

The fuel region inside the DSC is modeled as a heat source equal to 1.9 times the conservative decay heat power of 41.9 W/TMI-2 canister. The peaking factor of 1.9 is based on Reference 8.17. For calculating the maximum debris temperature, all 12 TMI-2 canisters inside the DSC are conservatively assumed to contain this decay heat power.

The steady state outer surface temperatures for the DSC resting inside the HSM are calculated in the HSM thermal analysis, described in Section 8.1.3.1. The results for each HSM analysis case are used to obtain maximum DSC surface temperatures for each region in the analytical model representing the DSC cylindrical shell. These surface temperatures are used as boundary conditions for the DSC thermal analysis and are assumed to remain constant.

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SAR-II-8.4 CH 8 Rev. 6 8-25 of 8-98 8.1.3.2.1 NUHOMS-12T DSC Inside HSM A.

Thermal Model of the NUHOMS-12T DSC The HEATING 7 computer program is used to perform the thermal analysis of the DSC internal basket assembly and TMI-2 canister regions. The analytical model of the DSC is shown in Figure 8.1-4 with the individual region numbers indicated. The model includes regions for the fuel debris, boral region with stainless steel skin, which surrounds the fuel debris, LICON, TMI-2 fuel canister shell, and the DSC shell. The gaps between the TMI-2 canisters and DSC shell are assumed to be filled with air.

The heat generated by the debris in each TMI-2 canister is assumed to be transferred by conduction to the TMI-2 canister shell regions. For the narrow gaps between adjacent canisters, heat transfer is assumed to occur through conduction and radiation. The TMI-2 canisters are open to the DSC atmosphere because the Hanson couplings on them may be removed or blocked open. The DSC cavity is vented to the atmosphere through HEPA filters. For the thermal analysis, any convection to the air inside the TMI-2 canisters or DSC is conservatively neglected. For the space between the horizontal and vertical surfaces of the outer TMI-2 canister and the DSC shell, heat is assumed to be transferred through conduction and radiation only. Heat transfer through the DSC shell is achieved by conduction.

The base case decay heat power of 80W for each TMI-2 canister is assumed to be uniformly distributed over the fuel regions inside the TMI-2 canister cavity. The resulting volumetric heat density is 3.982E-4 Btu/min.-in.3. The thermal properties from Table 8.1-6 and Table 8.1-7 are used in the HEATING 7 model of the DSC.

Figure 8.1-5 shows the resulting temperature distribution inside the DSC for the 87F ambient case.

The resulting calculated temperature profiles for the DSC are used for the evaluation of TMI-2 fuel debris and other DSC internal component temperatures. The results are summarized in Table 8.1-10.

B.

Evaluation of NUHOMS-12T DSC inside the HSM From the 87F ambient temperature profile in Figure 8.1-5 for the DSC, it can be observed that the maximum temperatures occur for the fuel regions in the center-most TMI-2 canister just above the horizontal center line of the DSC. The maximum temperature occurs slightly above the midplane because the lower half of the DSC shell is at a lower temperature than the upper half.

For variations in ambient air temperatures for normal operating conditions, the maximum calculated fuel debris temperature is 174F (79C) for the 87F ambient air case. The maximum calculated fuel debris temperature is less than 174F (79C) for the 45F and -20F ambient air cases. Therefore, the maximum fuel debris temperature for 87F ambient is well below the design basis fuel debris storage temperature limit of 724F (384C) defined in Section 3.3.7.1 for long term dry storage. For extreme ambient conditions, or short-term operating conditions, the maximum fuel debris temperature is 191F (88C). This value is well below the short-term fuel debris temperature limit of 1058F (570C) defined in Section 3.3.7.1.

These temperatures are well within the filter tested temperature range of -194C to 140C. The temperatures are also within the operating range for both the metallic seals and the elastomerics used in the filter assembly. The limiting case is the elastomerics used in the filter assembly that has a normal operating range for a static seal between -50F and 250F with the capability of going beyond these limits for short periods.

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SAR-II-8.4 CH 8 Rev. 6 8-26 of 8-98 8.1.3.3 Thermal Analysis of the NUHOMS-12T DSC Inside the MP187 Cask A.

NUHOMS-12T DSC in Cask During Transfer Operation The design basis heat load for the MP187 cask for 10 CFR Part 72 conditions is 13.5kW [8.12]. The design basis heat load in the NUHOMS-12T DSC is 0.86kW. Therefore, the thermal analysis for the NUHOMS-12T DSC in the MP187 cask is bounded by the results presented in the MP187 10 CFR Part 71 SAR [8.18] and 10 CFR Part 72 SAR [8.12].

8.1.4 Storage with Detected Leakage of Vent or Purge Port HEPA Filter Housing Seals A Limiting Condition for Operation (LCO) during storage operations at the TMI-2 ISFSI is maintenance of the DSC vent and purge ports HEPA Filter housing seals leak rate below 1E-02 standard cc/s. Compliance with the LCO is demonstrated through periodic performance of a leak check of the HEPA filter housing double metallic seals on both the vent port and purge port for each DSC containing TMI-2 canisters.

The basis for verifying the integrity of the DSC is the need to maintain confinement of the radioactive material stored in each DSC. The HEPA filter housing seals for the vent port and purge port make up part of the DSC confinement barrier. Failure of the confinement barrier is considered in the accident analysis discussed in Section 8.2.7. Verification of the HEPA filter housing seal integrity ensures that the HEPA filtered vent system is the only vent path for the DSC. The required action to reseat or replace the HEPA filter housing seals within seven days of a failed leak test recognizes the low motive force available to transport radioactive material through the leaking seals.

If the seals cannot pass the leak test after being reseated or replaced, continued storage with increased radioactive contamination survey frequencies and standard contamination control practices would be implemented.

Neither of the two intended functions of the DSC HEPA filters will be compromised with leaking DSC HEPA filter housing double metallic seals. Leaking seals will not disrupt the diffusion path for hydrogen. The HEPA filters will continue to provide filtration. It can be concluded from the analysis summarized below that any particulate radioactive material release through leaking double metallic seals would be negligible without a significant motive force. The calculated total effective dose equivalent at the INL boundary for an off-normal release due to a leaking DSC HEPA filter housing seal as well as the direct radiation dose from Table 7.4-2 is 0.16 mrem per year, well below the 25 mrem 10 CFR 72.104 limit for the whole body from normal operations and anticipated occurrences.

8.1.4.1 Hydrogen Venting Analysis. Leakage of the vent port seals would not adversely affect the ability of the vent system to allow hydrogen to diffuse through the HEPA filters.

8.1.4.2 Confinement Analysis. Any particulate radioactive material release through leaking double metallic seals would be negligible without a significant motive force. The only motive forces for release through leaking double metallic seals are diffusion, temperature gradients, and atmospheric pressure changes, all of which are quite small. For a release of radioactive material from within a DSC to occur, the radioactive material would have to be so fine that it can pass through a canister seal leak, become airborne and migrate to the DSC seal area, then pass through leaking double metallic seals with a significant motive force to sustain it. Without HEPA filter blockage, most of any motive force would be dissipated through the HEPA filters rather than acting to pass radioactive material through the seals. Even if the double metallic seals were completely missing, the gaps between the vent and purge filter housings and the DSC lid are so small that it would be difficult for particulate radioactive material to pass though without significant motive force..

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SAR-II-8.4 CH 8 Rev. 6 8-27 of 8-98 8.1.4.3 Continued Storage Analysis. Table 7.6-2, Off-normal Condition, provides the thyroid doses and the whole-body doses for the off-normal condition of a vent or purge port HEPA filter housing seal leakage. The calculations for this Off-normal condition conservatively assume that all 62 TMI-2 Filter Canisters are contained in the DSC experiencing the off-normal condition. The dose contribution from normal operations as well as the anticipated off-normal occurrence form effluent and direct radiation contributions show a calculated total effective dose equivalent at the INTEC boundary, of 115 mrem, well below the 10 CFR 20.1201 occupational dose limits for adults. This calculated total effective dose equivalent at the INL boundary is 0.11 mrem per year, well below the 10 CFR 20.1301 dose limits for individual members of the public.

Table 8.1-1. NUHOMS-12T Normal Operating Loading Identification.

LOAD TYPE AFFECTED COMPONENT DSC SHELL ASSEMBLY DSC INTERNALS DSC SUPPORT STRUCTURE REINFORCED CONCRETE HSM Dead Weight X

X X

X Internal Pressure X

Normal Thermal X

X X

Normal Handling X

X X

Live Loads X

Table 8.1-2. NUHOMS-12T Off-Normal Operating Loading Identification.

LOAD TYPE AFFECTED COMPONENT DSC SHELL ASSEMBLY DSC SUPPORT STRUCTURE REINFORCED CONCRETE HSM Dead Weight X

X X

Internal Pressure X

Off-Normal Thermal X

X X

Off-Normal Handling X

X X

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SAR-II-8.4 CH 8 Rev. 6 8-28 of 8-98 Table 8.1-3. Mechanical Properties of Materials.

MATERIAL TEMPERATURE

(°F)

STRESS PROPERTIES(1)

(ksi)

ELASTIC(1)

MODULUS (x1.0E3 ksi)

(E)

MEAN COEFFICIENT OF THERMAL EXPANSION (1)

( in./in.-°F)

STRESS INTENSITY (Sm)

YIELD STRENGTH (Sy)

ULTIMATE STRENGTH (Su)

Carbon 70 19.3 36.0 58.0 29.5 5.53 Steel ASME 100 19.3 36.0 29.3 5.53 SA36 200 19.3 32.8 28.8 5.89 300 19.3 31.8 28.3 6.26 Carbon 70 23.3 38.0 70.0 29.5 5.53 Steel Plate 100 23.3 38.0 70.0 29.3 5.53 ASME SA516 200 23.1 34.6 70.0 28.8 5.89 Grade 70 300 22.5 33.7 70.0 28.3 6.26 MATERIAL TEMPERATURE

(°F) 28 DAY COMPRESSIVE STRENGTH (ksi)

MODULUS OF ELASTICITY (1.0E3 ksi)

Concrete Normal Wt.(2) 5000 psi Strength 100 5.0 4.03 200 5.0 4.03 300 4.8 3.30 MATERIAL TEMPERATURE

(°F)

YIELD STRENGTH (ksi)

MODULUS OF ELASTICITY (1.0E3 ksi)

Reinforcing Steel(2)

ASTM A615 Grade 60 100 60.0 29.0 200 57.0 28.4 300 54.0 27.8 MATERIAL TEMPERATURE

(°F)

ALLOWABLE STRESS VALUES FOR CLASS 2(1)

COMPONENTS (S) (ksi)

YIELD STRENGTH(1)

(ksi)

Structural Bolting Material ASTM A325 100 20.2 81.0 200 20.2 73.9 400 20.2 69.3

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SAR-II-8.4 CH 8 Rev. 6 8-29 of 8-98 Table 8.1-3.Mechanical Properties of Materials. (continued)

MATERIAL TEMPERATURE

(°F)

YIELD STRESS (1)

(ksi)

NORMAL CONDITION ALLOWABLES Ftb (3, 5)

(ksi)

Fvb (4, 5)

(ksi)

SI (6) (ksi)

Filter Housing Bolts SA 193 Grade B8 Normal Conditions 100 30 20.0 12.0 27.0 200 25.8 17.2 10.3 23.2 300 23.3 15.5 9.3 21.0 MATERIAL TEMPERATURE

(°F)

YIELD STRESS (1, 7)

(ksi) 0.6 Sy ACCIDENT CONDITION ALLOWABLES Ftb (5, 8)

(ksi)

Fvb (5, 9)

(ksi)

Filter Housing Bolts SA 193 Grade B8 Accident Conditions 100 30 18.0 30 18.0 200 25.8 15.5 25.8 15.5 300 23.3 14.0 23.3 14.0 NOTES:

1.

Steel data and thermal expansion coefficients are obtained from ASME Boiler and Pressure Vessel Code,Section II, Part D [8.19].

2.

Concrete and reinforcing steel data were obtained from Handbook of Concrete Engineering, by Mark Fintel. [8.7]

3.

Allowable tensile stress Ftb = (2/3)Sy

4.

Allowable shear stress Fvb = (0.6) (2/3)Sy

5.

Tension and shear stresses must be combined using the following interaction equation:

f F

f F

tb tb vb vb

+

2 2

10.

6.

Stress intensity from the combined tensile, shear and residual torsion loads must be less than the allowable stress intensity: SI = 1.35(2/3)Sy

7.

Yield stress values from [8.19]. Note that Su is 75 ksi at all temperatures of interest.

8.

Allowable tensile stress Ftb is the smaller of 0.7Su or Sy. Where: 0.7Su = 0.7(75 ksi) = 52.5 ksi.

9.

Allowable shear stress Fvb is the smaller of 0.42Su or 0.6Sy. Where: 0.42Su = 0.42(75 ksi) = 31.5 ksi.

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SAR-II-8.4 CH 8 Rev. 6 8-30 of 8-98 Table 8.1-4. Estimated NUHOMS-12T Component Weights.

COMPONENT DESCRIPTION CALCULATED WEIGHT (POUNDS)

Dry Shielded Canister Shell Assembly 12,300 DSC Top Shield Plug Assembly 4,700 DSC Internal Basket Assembly 4,800 DSC Top Cover Plate Assembly 1,600 12 heaviest TMI-2 Canisters 35,000 Total Dry DSC Loaded Weight 58,400 MP187 Cask Empty Weight (include internal spacers) 159,400 MP187 Cask Max. Loaded Weight 217,800 HSM Single Module Weight (empty) 272,000 Table 8.1-5. NUHOMS-12T DSC Operating and Accident Pressures.

CASE AMBIENT AIR TEMPERATURE

(°F)

NORMAL AND OFF-NORMAL CONDITION PRESSURE (1)

(psia)

ACCIDENT CONDITION PRESSURE (2,3)

(psia)

DESIGN BASIS PRESSURE (psia) 1 45

<26.1

<26.1 29.7 2

87

<26.1

<26.1 29.7 3

103

<26.1 26.1 29.7 4

DSC in Cask with ambient air @ 87

<29.7

<29.7 29.7 (1)

Even though the normal and off-normal operating pressure is 0 psig, a 11.4 psig pressure is assumed for Case 1, 2, and 3, and a 15.0 psig pressure is assumed for Case 4. Atmosphere pressure is conservatively assumed to be 14.7 psia.

(2)

Assumes HEPA filters are blocked, the gap between adjacent HSMs is blocked, and this condition is not discovered for twelve months.

(3)

For a postulated accident involving hydrogen burn, a pressure spike of 194 psig lasting a few milliseconds is also considered for the TMI-2 canister.

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SAR-II-8.4 CH 8 Rev. 6 8-31 of 8-98 Table 8.1-6. Thermophyiscal Properties of Materials.

MATERIAL EFFECTIVE THERMAL CONDUCT (k) (Btu/hr-ft-°F)

EMISSIVITY ()

FRACTION Carbon Steel/Stainless Steel Table 8.1-7 [8.19]

0.587 [8.20]

Concrete Table 8.1-7 [8.43]

0.9 [8.16]

Soil 0.5 [8.44]

LICON 0.726 (Section 8.1.3.1.C)

UO2 Debris 4.623 [8.21]

Table 8.1-7. Temperature Dependent Thermophysical Properties.

TEMPERATURE

(°F)

THERMAL CONDUCTIVITY (Btu/hr-ft-°F)

Carbon Steel 70 100 150 200 300 400 500 23.62 23.90 24.19 24.41 24.41 24.19 23.69 Concrete

-50 100 200 300 400 500 1.10 1.17 1.14 1.11 1.08 1.04 Stainless Steel 70 100 150 200 250 300 350 400 450 500 8.57 8.71 9.00 9.29 9.58 9.79 10.08 10.37 10.58 10.87 Air [8.16]

0 32 100 200 300 400 500 0.0133 0.0140 0.0154 0.0174 0.0193 0.0212 0.0231

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SAR-II-8.4 CH 8 Rev. 6 8-32 of 8-98 Table 8.1-8. Thermal Load Case Definitions for NUHOMS-12T HSM Structural Analysis. (1)

THERMAL CONDITION CASE NO.

AMBIENT TEMPERATURE

(°F)

MAXIMUM INNER SURFACE TEMPERATURE (°F)

MAXIMUM OUTER SURFACE TEMPERATURE

(°F)

MAXIMUM THERMAL GRADIENT (2) (°F)

ROOF WALL FLOOR ROOF WALL FLOOR (3)

ROOF WALL FLOOR (3)

Normal Operating (To) 1 2

3 45 87

-20 90.8 128.7 4.0 74.5 114.6 3.9 77.2 115.4 9.9 87.6 125.5

-13.3 54.2 94.9

-11.7 69.5 107.1 4.3 3.2 3.2 17.3 20.3 19.7 15.6 7.7 8.3 5.6 Off-Normal (Ta) 1 2(2) 103

-50 152.5

-25.1 133.4

-25.0 133.5

-16.8 155.4

-42.7 111.4

-40.9 124.2

-22.0 2.9 17.6 22.0 15.9 9.3 5.2 Accident (Ta) 1 2(2) 103

-50 174.9

-11.6 178.6 9.7 167.6 16.4 159.4

-38.3 173.8 3.2 157.5 13.0 15.5 26.7 4.8 6.5 10.1 3.4 (1)

The maximum concrete temperature for normal operating condition is 129F, which is significantly below the ACI 349 temperature limit of 200F. Similarly, the maximum concrete temperature for accident conditions in 179F, which is also significantly below the ACI 349 temperature limit of 350F.

(2)

Based on maximum gradient at any cross-section.

(3)

The floor outside temperature and maximum gradient are reported for the two-foot thick HSM floor.

AMBIENT TEMPERATURE (4)

(°F)

FRONT WALL GRADIENT (5)

(°F)

BACK WALL GRADIENT (5)

(°F)

COORDINATES FOR THE MAX THERMAL GRADIENT (in.)

103(6) 10.9 (37.5, -34.5, 187.5 to 218) 103(6) 12.9 (37.5, -90, 0 to 36)

-50 8.7 (0, 34.5, 187.5 to 218)

-50 14.8 (0, 34.5, 0 to 36)

(4)

These off-normal cases will bound the normal operating cases. The -50F case will bound the

-20F case, which is the other case without solar insolation, because the outside surface temperature is raised 30F while the inside surface temperature would be raised by a lower amount. The 103F will bound the 87 and 45F cases since the solar insolation is significantly higher (105 vs. 67 Btu/hr-ft2)

(5)

The front and back wall do not include the regions from x = 37.5 to x = 61.5 in. These regions are considered part of the side wall.

(6)

These gradients are from higher outside surface temperatures to lower inside surface temperatures due to the combination of low decay heat and high solar insolation. The maximum gradients for this case from a higher inside surface to a lower outside surface is 6.1 and 11.2 F for the front and back wall respectively.

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SAR-II-8.4 CH 8 Rev. 6 8-33 of 8-98 Table 8.1-8. Thermal Load Case Definitions for NUHOMS-12T HSM Structural Analysis (concluded)

Normal Conditions (1)

Interior module with 45°F ambient temperature, 24 inch side walls open to ambient, solar heat flux of 67.0 Btu/hr-ft2 (2)

Interior module with 87°F ambient temperature, 24 inch side walls open to ambient, solar heat flux of 67.0 Btu/hr-ft2 (3)

Interior module with -20°F ambient temperature, 24 inch side walls open to ambient, solar heat flux neglected Off-Normal Conditions (1)

Interior module with 103°F ambient temperature, 24 inch side walls open to ambient, solar heat flux of 105 Btu/hr-ft2 (2)

Interior module with -50°F ambient temperature, 24 inch side walls open to ambient, neglect solar heat flux Accident Conditions (1)

Interior module with 103°F ambient temperature, the gap between HSMs is blocked, solar heat flux of 105 Btu/hr-ft2 (2)

Interior module with -50°F ambient temperature, the gap between HSMs is blocked, neglect solar heat flux Table 8.1-9. NUHOMS-12T DSC Thermal Analysis Results Summary. (1)

THERMAL CONDITIONS CASE AMBIENT TEMPERATURE

(°F)

MAXIMUM DSC OUTER SURFACE TEMPERATURE

(°F)

BOTTOM SIDE TOP Normal 1

45 94.4 94.7 97.8 2

87 130.3 131.0 134.2 3

-20 29.1 28.2 28.1 Off-Normal 1

103 148.5 149.7 154.4 2

-50 4.8 3.8 3.7 Accident 1

103 180.8 183.8 183.5 2

-50 28.1 26.2 23.3 (1)

See Table 8.1-8 for concrete temperatures and definitions of these cases.

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SAR-II-8.4 CH 8 Rev. 6 8-34 of 8-98 Table 8.1-10 (1, 2). NUHOMS DSC Thermal Analysis Results Summary.

THERMAL CONDITIONS CASE AMBIENT TEMPERATURE

(°F)

MAXIMUM FUEL DEBRIS TEMPERATURE

(°F /°C)

FUEL DEBRIS ACCEPTANCE CRITERIA (3)

(°F /°C)

Normal 1

45

<174/79 724/384 2

87 174/79 724/384 3

-20

<174/79 724/384 Off-Normal 1

103 191/88 1058/570 2

-50

<191/88 1058/570 Accident 1

103 219/104 1058/570 2

-50

<219/104 1058/570 (DSC in cask with internal vacuum) 1 100(3)

<289/143 1058/570 (1)

See Table 8.1-8 for concrete temperatures and definitions of these cases.

(2)

See Table 8.1-9 for DSC Shell temperatures.

(3)

These temperature limits are based on intact fuel. The TMI-2 debris has been subjected to temperatures greater than these and the cladding is degraded. As such, the typical fuel temperature limits are not applicable with respect to cladding damage.

These limits are used here for consistency.

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SAR-II-8.4 CH 8 Rev. 6 8-35 of 8-98 Table 8.1-11. Maximum NUHOMS-12T DSC Stresses for Normal Loads.

DSC COMPONENTS STRESS TYPE STRESS INTENSITY (ksi) (1)

DEAD WEIGHT INTERNAL PRESSURE (2)

THERMAL (3)

NORMAL HANDLING DSC Shell Primary Membrane 2.5 10.5 3.3 Membrane +

Bending 5.3 17.8 9.6 Primary +

Secondary 5.3 17.8 1.6 15.0 Top Shield Plug Primary Membrane 0.6 4.4 0.3 Membrane +

Bending 1.3 6.2 0.3 Primary +

Secondary 1.3 6.2 0.2 0.3 Top Cover Plate Primary Membrane 1.2 8.0 0.3 Membrane +

Bending 1.7 12.0 0.3 Primary +

Secondary 1.7 12.0 0.5 0.3 Inner Bottom Cover Plate Primary Membrane 0.8 0.7 0.3 Membrane +

Bending 1.3 1.0 0.3 Primary +

Secondary 1.3 1.0 0.3 0.7 Outer Bottom Cover Plate Primary Membrane 0.3 3.6 2.0 Membrane +

Bending 0.5 7.1 10.3 Primary +

Secondary 0.5 7.1 0.3 13.4 (1)

Values shown are maximum irrespective of location.

(2)

Even though the normal operating pressure is 0 psig, a bounding 15 psig pressure load is applied.

(3)

All thermal stresses are classified as secondary. All stresses from dead weight and pressure loads are classified as primary membrane or bending.

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SAR-II-8.4 CH 8 Rev. 6 8-36 of 8-98 Table 8.1-12. Maximum NUHOMS-12T DSC Stresses for Off-Normal Loads.

DSC COMPONENTS STRESS TYPE STRESS INTENSITY (ksi) (1)

INTERNAL PRESSURE (2)

THERMAL OFF-NORMAL HANDLING DSC Shell Primary Membrane 10.5 N/A 1.9 Membrane + Bending 17.8 N/A 13.0 Primary + Secondary 17.8 1.6 30.0 Top Shield Plug Primary Membrane 4.4 N/A 0.0 Membrane + Bending 6.2 N/A 0.0 Primary + Secondary 6.2 0.2 0.2 Top Cover Plate Primary Membrane 8.0 N/A 0.1 Membrane + Bending 12.0 N/A 0.1 Primary + Secondary 12.0 0.5 0.6 Inner Bottom Cover Plate Primary Membrane 0.7 N/A 0.0 Membrane + Bending 1.0 N/A 0.3 Primary + Secondary 1.0 0.3 1.3 Outer Bottom Cover Plate Primary Membrane 3.6 N/A 4.0 Membrane + Bending 7.1 N/A 20.6 Primary + Secondary 7.1 0.3 26.8 (1) Values shown are maximum irrespective of location.

(2) Even though the normal operating pressure is 0 psig, a bounding 15 psig pressure load is applied.

Table 8.1-13. Maximum DSC Structure Stresses for Normal and Off-Normal Loads.

COMPONENTS LOAD TYPE CALCULATED STRESS AXIAL (ksi)

SHEAR (ksi)

Support Rail Dead Weight 0.00 0.9 Normal DSC Handling Loads 1.42 3.40 Off-Normal DSC Handling Loads 2.84 4.00 Cross Beam Dead Weight 0.13 0.02 Normal DSC Handling Loads 1.20 0.00 Off-Normal DSC Handling Loads 1.70 0.04

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SAR-II-8.4 CH 8 Rev. 6 8-37 of 8-98 Table 8.1-14. Maximum DSC Support Structure Vertical Displacements for Normal and Off-Normal Loads COMPONENTS LOAD TYPE MAXIMUM VERTICAL DISPLACEMENTS (in.)

Support Rails Dead Weight DWs + DWc 0.004 Normal DSC Handling Loads DWs + HLn 0.047 Off-Normal DSC Handling Loads DWs +DWc + HLo 0.094 Table 8.1-15. Maximum NUHOMS-12T HSM Reinforced Concrete Bending Moments and Shear Forces for Normal and Off-Normal Loads.

STRUCTURAL SECTION FORCE COMPONENT HSM INTERNAL FORCES (kiP/ft., in.-k/ft.) (1)

DEAD WEIGHT LIVE LOADS NORMAL (2,3)

THERMAL OFF-NORMAL (4) THERMAL Floor Slab Shear 0.57 0.1 6.9 6.9 Moment 33.7 3.3 357.2 357.2 Side Wall Shear 0.26 0.1 9.3 9.3 Moment 17.5 2.9 582.4 582.4 Front Wall Shear 9.6 2.9 42.6 42.6 Moment 175.0 36.1 791.5 791.5 Rear Wall Shear 1.1 0.5 33.0 33.0 Moment 128.2 12.1 532.6 532.6 Roof Slab Shear 0.45 0.2 3.2 3.2 Moment 33.3 10.8 430.2 430.2 (1)

Values shown are maximums irrespective of location.

(2)

Maximum moments are based on cracked section properties.

(3)

See Table 8.1-8 for concrete temperatures and definitions of these cases.

(4)

Worst case internal loads are used for normal, off-normal and accident conditions. Normal thermal results are bounded by off-normal thermal results. Therefore, off-normal results are tabulated both for normal and off-normal cases.

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SAR-II-8.4 CH 8 Rev. 6 8-38 of 8-98 Figure 8.1-1. Heating 7 Model of NUHOMS-12T DSC Used in HSM.

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SAR-II-8.4 CH 8 Rev. 6 8-39 of 8-98 Figure 8.1-1. HEATING 7 Model of NUHOMS-12T DSC Used in HSM (Concluded)

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SAR-II-8.4 CH 8 Rev. 6 8-40 of 8-98 TMI-4D7 HSM 45 F AMB,0.86kW/DSC,non-insulated side, solar Thu Aug 8 08:28:01 1996 Steady-State Temperature Distribution at Time 0.0000E+00 Z = 1.1050E+02 64 72.00 l 87.56 87.50 87.33 87.22 87.10 86.95 86.78 86.65 86.60 86.33 86.03 85.72 85.63 85.53 63 71.75 l 87.55 87.49 87.30 87.20 87.07 86.92 86.74 86.60 86.55 86.28 85.97 85.64 85.55 85.45 62 65.31 l 87.33 87.22 86.88 86.68 86.45 86.17 85.86 85.59 85.50 84.99 84.42 83.83 83.66 83.48 61 58.88 l 87.49 87.34 86.87 86.59 86.26 85.86 85.41 85.03 84.90 84.17 83.34 82.48 82.24 81.99 60 52.44 l 88.21 88.02 87.46 87.11 86.70 86.20 85.63 85.14 84.97 84.02 82.92 81.77 81.45 81.11 59 46.00 l 89.58 89.39 88.79 88.42 87.97 87.42 86.77 86.21 86.00 84.85 83.45 81.91 81.46 81.01 58 44.00 l 90.14 89.96 89.38 89.01 88.57 88.03 87.39 86.83 86.62 85.45 83.98 82.22 81.71 81.19 57 42.00 l 90.77 90.60 90.04 89.70 89.27 88.75 88.14 87.59 87.39 86.25 84.77 82.75 82.04 81.40 56 38.25 l 93.76 93.58 92.87 92.41 91.86 91.19 90.40 89.69 89.44 87.96 86.07 83.82 83.15 82.45 55 34.50 l 96.98 96.86 96.06 95.50 94.84 94.09 93.23 92.47 92.20 90.62 88.55 86.02 85.23 84.41 54 33.59 l 97.77 97.72 96.91 96.29 95.62 94.86 93.99 93.25 92.97 91.43 89.37 86.79 85.97 85.11 53 32.97 l 97.79 97.73 97.51 96.85 96.16 95.40 94.54 93.80 93.54 92.02 89.99 87.38 86.55 85.66 52 32.34 l 97.85 97.75 97.52 97.41 96.71 95.95 95.10 94.38 94.12 92.65 90.66 88.03 87.18 86.25 51 31.72 l 97.91 97.81 97.54 97.41 97.27 96.52 95.68 94.96 94.71 93.31 91.38 88.75 87.87 86.91 50 31.09 l 97.98 97.87 97.60 97.42 97.27 97.09 96.27 95.56 95.33 94.00 92.16 89.55 88.63 87.63 49 30.47 l 98.04 97.94 97.66 97.48 97.27 97.08 96.87 96.14 95.97 94.73 93.00 90.43 89.48 88.42 48 30.00 l 98.08 97.98 97.70 97.52 97.32 97.08 96.87 96.55 96.47 95.29 93.67 91.18 90.18 89.07 47 29.84 l 98.09 98.00 97.71 97.54 97.33 97.09 96.87 96.68 96.64 95.48 93.90 91.44 90.43 89.29 46 29.22 l 98.15 98.05 97.77 97.59 97.38 97.14 96.87 96.68 96.61 96.27 94.87 92.58 91.49 90.25 45 28.59 l 98.21 98.11 97.82 97.65 97.44 97.20 96.92 96.71 96.61 96.26 95.91 93.93 92.69 91.28 44 27.97 l 98.26 98.16 97.87 97.70 97.49 97.25 96.98 96.75 96.67 96.27 95.90 95.58 94.05 92.37 43 27.34 l 98.31 98.21 97.92 97.75 97.54 97.30 97.03 96.80 96.73 96.32 95.90 95.58 95.54 93.44 42 24.68 l 98.50 98.41 98.11 97.94 97.73 97.49 97.22 97.00 96.92 96.50 96.04 95.55 95.50 95.46 41 21.94 l 98.67 98.57 98.27 98.10 97.89 97.65 97.38 97.16 97.08 96.65 96.17 95.65 95.50 95.45 40 19.20 l 98.80 98.70 98.40 98.23 98.02 97.78 97.51 97.28 97.20 96.77 96.28 95.76 95.61 95.46 39 16.45 l 98.89 98.79 98.50 98.32 98.11 97.87 97.60 97.37 97.29 96.86 96.37 95.86 95.71 95.56 38 15.00 l 98.93 98.83 98.53 98.36 98.15 97.91 97.63 97.41 97.33 96.90 96.41 95.90 95.76 95.62 37 13.71 l 98.95 98.85 98.55 98.38 98.17 97.93 97.66 97.43 97.35 96.92 96.44 95.93 95.79 95.64 36 10.97 l 98.97 98.88 98.58 98.40 98.20 97.96 97.68 97.46 97.38 96.95 96.47 95.97 95.83 95.69 35 8.22 l 98.96 98.86 98.57 98.39 98.19 97.95 97.68 97.45 97.37 96.95 96.48 95.99 95.85 95.70 34 7.50 l 98.95 98.85 98.56 98.38 98.18 97.94 97.67 97.45 97.37 96.94 96.47 95.98 95.85 95.70 33 5.48 l 98.91 98.81 98.52 98.34 98.14 97.90 97.63 97.41 97.33 96.91 96.45 95.96 95.83 95.69 32 2.74 l 98.83 98.73 98.43 98.26 98.06 97.82 97.55 97.33 97.26 96.84 96.38 95.90 95.77 95.63 31.00 l 98.71 98.61 98.31 98.14 97.94 97.70 97.44 97.22 97.14 96.73 96.27 95.80 95.67 95.54 30 -2.74 l 98.55 98.45 98.16 97.99 97.78 97.55 97.28 97.06 96.99 96.58 96.12 95.65 95.52 95.39 29 -5.48 l 98.35 98.26 97.96 97.79 97.59 97.35 97.09 96.87 96.79 96.38 95.92 95.45 95.32 95.18 28 -7.50 l 98.19 98.09 97.80 97.62 97.42 97.19 96.92 96.70 96.63 96.21 95.75 95.28 95.14 95.01 27 -8.22 l 98.12 98.02 97.73 97.56 97.36 97.12 96.86 96.64 96.56 96.15 95.69 95.21 95.08 94.94 26 -10.97 l 97.85 97.76 97.46 97.29 97.09 96.85 96.59 96.37 96.29 95.87 95.41 94.93 94.79 94.65 25 -13.71 l 97.55 97.45 97.16 96.98 96.78 96.55 96.28 96.06 95.98 95.56 95.09 94.61 94.47 94.33 24 -15.00 l 97.39 97.29 97.00 96.83 96.62 96.39 96.12 95.90 95.82 95.40 94.94 94.45 94.31 94.17 23 -16.45 l 97.21 97.11 96.81 96.64 96.44 96.20 95.93 95.71 95.64 95.22 94.75 94.25 94.11 93.97 22 -19.20 l 96.83 96.73 96.43 96.26 96.05 95.82 95.55 95.33 95.26 94.84 94.37 93.87 93.73 93.58 21 -21.94 l 96.41 96.31 96.01 95.84 95.64 95.40 95.13 94.91 94.84 94.43 93.97 93.47 93.33 93.29 20 -24.68 l 95.96 95.86 95.56 95.38 95.18 94.94 94.67 94.45 94.38 93.98 93.55 93.09 93.05 93.02 19 -27.34 l 95.48 95.38 95.08 94.91 94.69 94.45 94.18 93.96 93.88 93.49 93.11 92.88 92.85 90.23 18 -27.97 l 95.37 95.26 94.97 94.79 94.58 94.33 94.06 93.84 93.75 93.36 93.07 92.86 90.84 88.73 17 -28.59 l 95.25 95.14 94.85 94.67 94.46 94.21 93.93 93.73 93.61 93.33 93.06 90.51 88.92 87.15 16 -28.62 l 95.24 95.14 94.84 94.66 94.45 94.20 93.93 93.72 93.61 93.33 92.98 90.41 88.83 87.08 15 -29.22 l 95.13 95.02 94.73 94.55 94.33 94.08 93.80 93.64 93.59 93.32 91.46 88.54 87.16 85.62 14 -29.84 l 95.01 94.90 94.61 94.43 94.21 93.96 93.77 93.62 93.60 92.05 89.95 86.81 85.56 84.16 13 -30.47 l 94.88 94.77 94.49 94.30 94.08 93.93 93.76 92.73 92.50 90.81 88.52 85.26 84.08 82.79 12 -31.09 l 94.76 94.65 94.37 94.18 94.05 93.91 92.77 91.76 91.44 89.61 87.16 83.84 82.71 81.49

+----------------------------------------------------------------------------------------------------------------

.00 5.00 10.00 12.00 14.00 16.00 18.00 19.50 20.00 22.50 25.00 27.34 27.97 28.59 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Figure 8.1-2. HSM HEATING 7 Results for 45F Ambient.

Idaho Cleanup Project 412.09 (01/01/2022 - Rev. 13)

SAFETY ANALYSIS REPORT FOR TMI-2 ANALYSIS OF DESIGN EVENTS Identifier:

Revision*:

Page:

SAR-II-8.4 CH 8 Rev. 6 8-41 of 8-98 TMI-4D7 HSM 45 F AMB,0.86kW/DSC,non-insulated side, solar Thu Aug 8 08:28:01 1996 Steady-State Temperature Distribution at Time 0.0000E+00 Z = 1.1050E+02 64 72.00 l 87.56 87.50 87.33 87.22 87.10 86.95 86.78 86.65 86.60 86.33 86.03 85.72 85.63 85.53 63 71.75 l 87.55 87.49 87.30 87.20 87.07 86.92 86.74 86.60 86.55 86.28 85.97 85.64 85.55 85.45 62 65.31 l 87.33 87.22 86.88 86.68 86.45 86.17 85.86 85.59 85.50 84.99 84.42 83.83 83.66 83.48 61 58.88 l 87.49 87.34 86.87 86.59 86.26 85.86 85.41 85.03 84.90 84.17 83.34 82.48 82.24 81.99 60 52.44 l 88.21 88.02 87.46 87.11 86.70 86.20 85.63 85.14 84.97 84.02 82.92 81.77 81.45 81.11 59 46.00 l 89.58 89.39 88.79 88.42 87.97 87.42 86.77 86.21 86.00 84.85 83.45 81.91 81.46 81.01 58 44.00 l 90.14 89.96 89.38 89.01 88.57 88.03 87.39 86.83 86.62 85.45 83.98 82.22 81.71 81.19 57 42.00 l 90.77 90.60 90.04 89.70 89.27 88.75 88.14 87.59 87.39 86.25 84.77 82.75 82.04 81.40 56 38.25 l 93.76 93.58 92.87 92.41 91.86 91.19 90.40 89.69 89.44 87.96 86.07 83.82 83.15 82.45 55 34.50 l 96.98 96.86 96.06 95.50 94.84 94.09 93.23 92.47 92.20 90.62 88.55 86.02 85.23 84.41 54 33.59 l 97.77 97.72 96.91 96.29 95.62 94.86 93.99 93.25 92.97 91.43 89.37 86.79 85.97 85.11 53 32.97 l 97.79 97.73 97.51 96.85 96.16 95.40 94.54 93.80 93.54 92.02 89.99 87.38 86.55 85.66 52 32.34 l 97.85 97.75 97.52 97.41 96.71 95.95 95.10 94.38 94.12 92.65 90.66 88.03 87.18 86.25 51 31.72 l 97.91 97.81 97.54 97.41 97.27 96.52 95.68 94.96 94.71 93.31 91.38 88.75 87.87 86.91 50 31.09 l 97.98 97.87 97.60 97.42 97.27 97.09 96.27 95.56 95.33 94.00 92.16 89.55 88.63 87.63 49 30.47 l 98.04 97.94 97.66 97.48 97.27 97.08 96.87 96.14 95.97 94.73 93.00 90.43 89.48 88.42 48 30.00 l 98.08 97.98 97.70 97.52 97.32 97.08 96.87 96.55 96.47 95.29 93.67 91.18 90.18 89.07 47 29.84 l 98.09 98.00 97.71 97.54 97.33 97.09 96.87 96.68 96.64 95.48 93.90 91.44 90.43 89.29 46 29.22 l 98.15 98.05 97.77 97.59 97.38 97.14 96.87 96.68 96.61 96.27 94.87 92.58 91.49 90.25 45 28.59 l 98.21 98.11 97.82 97.65 97.44 97.20 96.92 96.71 96.61 96.26 95.91 93.93 92.69 91.28 44 27.97 l 98.26 98.16 97.87 97.70 97.49 97.25 96.98 96.75 96.67 96.27 95.90 95.58 94.05 92.37 43 27.34 l 98.31 98.21 97.92 97.75 97.54 97.30 97.03 96.80 96.73 96.32 95.90 95.58 95.54 93.44 42 24.68 l 98.50 98.41 98.11 97.94 97.73 97.49 97.22 97.00 96.92 96.50 96.04 95.55 95.50 95.46 41 21.94 l 98.67 98.57 98.27 98.10 97.89 97.65 97.38 97.16 97.08 96.65 96.17 95.65 95.50 95.45 40 19.20 l 98.80 98.70 98.40 98.23 98.02 97.78 97.51 97.28 97.20 96.77 96.28 95.76 95.61 95.46 39 16.45 l 98.89 98.79 98.50 98.32 98.11 97.87 97.60 97.37 97.29 96.86 96.37 95.86 95.71 95.56 38 15.00 l 98.93 98.83 98.53 98.36 98.15 97.91 97.63 97.41 97.33 96.90 96.41 95.90 95.76 95.62 37 13.71 l 98.95 98.85 98.55 98.38 98.17 97.93 97.66 97.43 97.35 96.92 96.44 95.93 95.79 95.64 36 10.97 l 98.97 98.88 98.58 98.40 98.20 97.96 97.68 97.46 97.38 96.95 96.47 95.97 95.83 95.69 35 8.22 l 98.96 98.86 98.57 98.39 98.19 97.95 97.68 97.45 97.37 96.95 96.48 95.99 95.85 95.70 34 7.50 l 98.95 98.85 98.56 98.38 98.18 97.94 97.67 97.45 97.37 96.94 96.47 95.98 95.85 95.70 33 5.48 l 98.91 98.81 98.52 98.34 98.14 97.90 97.63 97.41 97.33 96.91 96.45 95.96 95.83 95.69 32 2.74 l 98.83 98.73 98.43 98.26 98.06 97.82 97.55 97.33 97.26 96.84 96.38 95.90 95.77 95.63 31.00 l 98.71 98.61 98.31 98.14 97.94 97.70 97.44 97.22 97.14 96.73 96.27 95.80 95.67 95.54 30 -2.74 l 98.55 98.45 98.16 97.99 97.78 97.55 97.28 97.06 96.99 96.58 96.12 95.65 95.52 95.39 29 -5.48 l 98.35 98.26 97.96 97.79 97.59 97.35 97.09 96.87 96.79 96.38 95.92 95.45 95.32 95.18 28 -7.50 l 98.19 98.09 97.80 97.62 97.42 97.19 96.92 96.70 96.63 96.21 95.75 95.28 95.14 95.01 27 -8.22 l 98.12 98.02 97.73 97.56 97.36 97.12 96.86 96.64 96.56 96.15 95.69 95.21 95.08 94.94 26 -10.97 l 97.85 97.76 97.46 97.29 97.09 96.85 96.59 96.37 96.29 95.87 95.41 94.93 94.79 94.65 25 -13.71 l 97.55 97.45 97.16 96.98 96.78 96.55 96.28 96.06 95.98 95.56 95.09 94.61 94.47 94.33 24 -15.00 l 97.39 97.29 97.00 96.83 96.62 96.39 96.12 95.90 95.82 95.40 94.94 94.45 94.31 94.17 23 -16.45 l 97.21 97.11 96.81 96.64 96.44 96.20 95.93 95.71 95.64 95.22 94.75 94.25 94.11 93.97 22 -19.20 l 96.83 96.73 96.43 96.26 96.05 95.82 95.55 95.33 95.26 94.84 94.37 93.87 93.73 93.58 21 -21.94 l 96.41 96.31 96.01 95.84 95.64 95.40 95.13 94.91 94.84 94.43 93.97 93.47 93.33 93.29 20 -24.68 l 95.96 95.86 95.56 95.38 95.18 94.94 94.67 94.45 94.38 93.98 93.55 93.09 93.05 93.02 19 -27.34 l 95.48 95.38 95.08 94.91 94.69 94.45 94.18 93.96 93.88 93.49 93.11 92.88 92.85 90.23 18 -27.97 l 95.37 95.26 94.97 94.79 94.58 94.33 94.06 93.84 93.75 93.36 93.07 92.86 90.84 88.73 17 -28.59 l 95.25 95.14 94.85 94.67 94.46 94.21 93.93 93.73 93.61 93.33 93.06 90.51 88.92 87.15 16 -28.62 l 95.24 95.14 94.84 94.66 94.45 94.20 93.93 93.72 93.61 93.33 92.98 90.41 88.83 87.08 15 -29.22 l 95.13 95.02 94.73 94.55 94.33 94.08 93.80 93.64 93.59 93.32 91.46 88.54 87.16 85.62 14 -29.84 l 95.01 94.90 94.61 94.43 94.21 93.96 93.77 93.62 93.60 92.05 89.95 86.81 85.56 84.16 13 -30.47 l 94.88 94.77 94.49 94.30 94.08 93.93 93.76 92.73 92.50 90.81 88.52 85.26 84.08 82.79 12 -31.09 l 94.76 94.65 94.37 94.18 94.05 93.91 92.77 91.76 91.44 89.61 87.16 83.84 82.71 81.49

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.00 5.00 10.00 12.00 14.00 16.00 18.00 19.50 20.00 22.50 25.00 27.34 27.97 28.59 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Figure 8.1-2. HSM HEATING 7 Results for 45F Ambient (Continued)

Idaho Cleanup Project 412.09 (01/01/2022 - Rev. 13)

SAFETY ANALYSIS REPORT FOR TMI-2 ANALYSIS OF DESIGN EVENTS Identifier:

Revision*:

Page:

SAR-II-8.4 CH 8 Rev. 6 8-42 of 8-98 TMI-4D7 HSM 45 F AMB,0.86kW/DSC, non-insulated side, solar Thu Aug 8 08:28:01 1996 Steady-State Temperature Distribution at Time 0.0000E+00 Z = 1.1050E+02 64 72.00 l 69.74 63 71.75 l 69.16 62 65.31 l 61.96 61 58.88 l 58.66 60 52.44 l 56.80 59 46.00 l 55.62 58 44.00 l 55.34 57 42.00 l 55.10 56 38.25 l 54.74 55 34.50 l 54.50 54 33.59 l 54.46 53 32.97 l 54.43 52 32.34 l 54.41 51 31.72 l 54.39 50 31.09 l 54.37 49 30.47 l 54.36 48 30.00 l 54.34 47 29.84 l 54.34 46 29.22 l 54.33 45 28.59 l 54.32 44 27.97 l 54.31 43 27.34 l 54.31 42 24.68 l 54.30 41 21.94 l 54.31 40 19.20 l 54.34 39 16.45 l 54.36 38 15.00 l 54.36 37 13.71 l 54.37 36 10.97 l 54.37 35 8.22 l 54.35 34 7.50 l 54.34 33 5.48 l 54.31 32 2.74 l 54.24 31.00 l 54.15 30 -2.74 l 54.03 29 -5.48 l 53.88 28 -7.50 l 53.75 27 -8.22 l 53.69 26 -10.97 l 53.47 25 -13.71 l 53.21 24 -15.00 l 53.08 23 -16.45 l 52.92 22 -19.20 l 52.59 21 -21.94 l 52.24 20 -24.68 l 51.86 19 -27.34 l 51.48 18 -27.97 l 51.39 17 -28.59 l 51.30 16 -28.62 l 51.29 15 -29.22 l 51.21 14 -29.84 l 51.12 13 -30.47 l 51.03 12 -31.09 l 50.95

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61.50 29 TMI-4D7 HSM 45 F AMB,0.86kW/DSC,non-insulated side, solar Thu Aug 8 08:28:01 1996 Steady-State Temperature Distribution at Time 0.0000E+00 Z = 1.1050E+02 11 -31.72 l 94.63 94.52 94.24 94.14 94.03 93.00 91.81 90.77 90.42 88.44 85.86 82.52 81.43 80.27 10 -32.34 l 94.50 94.38 94.19 94.12 93.17 92.10 90.86 89.79 89.41 87.32 84.63 81.28 80.23 79.10 9 -32.97 l 94.38 94.34 94.17 93.29 92.33 91.22 89.93 88.82 88.43 86.23 83.45 80.11 79.08 78.00 8 -33.59 l 94.35 94.32 93.31 92.47 91.49 90.36 89.02 87.88 87.47 85.17 82.32 78.99 77.99 76.94 7 -34.50 l 93.48 93.25 92.11 91.30 90.32 89.14 87.75 86.55 86.12 83.70 80.76 77.46 76.49 75.48 6 -56.25 l 77.29 76.89 75.64 74.90 74.01 72.98 71.80 70.82 70.47 68.60 66.48 64.30 63.68 63.05 5 -78.00 l 77.18 76.96 76.26 75.81 75.26 74.59 73.76 73.03 72.76 71.21 69.23 66.84 66.13 65.43 4 -90.00 l 69.51 69.29 68.64 68.25 67.78 67.23 66.61 66.09 65.90 64.91 63.81 62.69 62.38 62.07 3 -102.00 l 64.46 64.30 63.82 63.54 63.22 62.84 62.42 62.08 61.96 61.32 60.63 59.94 59.75 59.55 2 -120.00 l 60.56 60.45 60.14 59.96 59.75 59.50 59.23 59.01 58.93 58.51 58.05 57.59 57.46 57.32 1 -204.00 l 65.00 65.00 65.00 65.00 65.00 65.00 65.00 65.00 65.00 65.00 65.00 65.00 65.00 65.00

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.00 5.00 10.00 12.00 14.00 16.00 18.00 19.50 20.00 22.50 25.00 27.34 27.97 28.59 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Figure 8.1-2. HSM HEATING 7 Results for 45F Ambient. (Continued)

Idaho Cleanup Project 412.09 (01/01/2022 - Rev. 13)

SAFETY ANALYSIS REPORT FOR TMI-2 ANALYSIS OF DESIGN EVENTS Identifier:

Revision*:

Page:

SAR-II-8.4 CH 8 Rev. 6 8-43 of 8-98 TMI-4D7 HSM 45 F AMB,0.86kW/DSC, non-insulated side, solar Thu Aug 8 08:28:01 1996 Steady-State Temperature Distribution at Time 0.0000E+00 Z = 1.1050E+02 11 -31.72 l 79.04 77.75 76.42 75.05 73.66 72.24 70.79 69.33 67.17 59.69 58.92 57.63 56.10 53.56 10 -32.34 l 77.93 76.69 75.42 74.11 72.77 71.40 70.01 68.59 66.49 59.20 58.53 57.34 55.89 53.42 9 -32.97 l 76.86 75.68 74.46 73.20 71.91 70.59 69.24 67.86 65.83 58.75 58.16 57.06 55.68 53.29 8 -33.59 l 75.84 74.70 73.53 72.31 71.07 69.79 68.49 67.16 65.20 58.31 57.80 56.79 55.48 53.16 7 -34.50 l 74.43 73.35 72.24 71.09 69.91 68.69 67.45 66.19 64.31 57.72 57.30 56.41 55.19 52.97 6 -56.25 l 62.41 61.76 61.10 60.42 59.73 59.03 58.32 57.60 56.54 52.88 52.67 52.21 51.55 50.25 5 -78.00 l 64.75 64.09 63.44 62.81 62.18 61.56 60.96 60.35 59.49 56.65 55.29 53.98 52.71 50.86 4 -90.00 l 61.76 61.45 61.13 60.82 60.50 60.19 59.88 59.57 59.12 57.70 56.00 54.46 53.03 51.03 3 -102.00 l 59.36 59.16 58.96 58.76 58.56 58.36 58.16 57.95 57.65 56.65 55.32 54.00 52.69 50.78 2 -120.00 l 57.19 57.05 56.91 56.77 56.63 56.49 56.34 56.19 55.97 55.21 54.14 53.00 51.79 49.85 1 -204.00 l 65.00 65.00 65.00 65.00 65.00 65.00 65.00 65.00 65.00 65.00 65.00 65.00 65.00 65.00

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29.22 29.84 30.47 31.09 31.72 32.34 32.97 33.59 34.50 37.50 41.50 45.50 49.50 55.50 15 16 17 18 19 20 21 22 23 24 25 26 27 28 TMI-4D7 HSM 45 F AMB,0.86kW/DSC,non-insulated side, solar Thu Aug 8 08:28:01 1996 Steady-State Temperature Distribution at Time 0.0000E+00 Z = 1.1050E+02 11 -31.72 l 50.86 10 -32.34 l 50.77 9 -32.97 l 50.69 8 -33.59 l 50.61 7 -34.50 l 50.49 6 -56.25 l 48.69 5 -78.00 l 49.04 4 -90.00 l 49.13 3 -102.00 l 48.95 2 -120.00 l 47.74 1 -204.00 l 65.00

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61.50 29 Figure 8.1-2. HSM HEATING 7 Results for 45F Ambient. (Concluded)

Idaho Cleanup Project 412.09 (01/01/2022 - Rev. 13)

SAFETY ANALYSIS REPORT FOR TMI-2 ANALYSIS OF DESIGN EVENTS Identifier:

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SAR-II-8.4 CH 8 Rev. 6 8-44 of 8-98 Figure 8.1-3. NUHOMS-12T HSM Temperature Distribution for 45F Ambient

Idaho Cleanup Project 412.09 (01/01/2022 - Rev. 13)

SAFETY ANALYSIS REPORT FOR TMI-2 ANALYSIS OF DESIGN EVENTS Identifier:

Revision*:

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SAR-II-8.4 CH 8 Rev. 6 8-45 of 8-98 Figure 8.1-4. HEATING 7 Model of NUHOMS-12T DSC.

Idaho Cleanup Project 412.09 (01/01/2022 - Rev. 13)

SAFETY ANALYSIS REPORT FOR TMI-2 ANALYSIS OF DESIGN EVENTS Identifier:

Revision*:

Page:

SAR-II-8.4 CH 8 Rev. 6 8-46 of 8-98 TMI-2 Fuel Debris Can TMI87.inp, normal at 87F case, horiz with air Thu Aug 8 15:19:10 1996 Steady-State Temperature Distribution at Time 0.0000E+00 59 33.62 l 134.20 134.20 134.20 134.20 134.20 134.20 134.20 134.20 134.20 134.20 134.20 58 33.00 l 134.25 134.25 134.24 134.24 134.24 134.24 134.24 134.24 134.24 134.24 134.24 57 31.85 l 143.09 143.32 143.34 143.30 143.27 143.11 143.04 142.18 141.85 135.94 134.20 134.20 134.20 134.20 56 31.23 l 147.75 148.25 148.30 148.24 148.20 148.00 147.91 146.90 146.51 138.74 134.25 134.24 134.24 134.24 55 30.41 l 153.49 154.76 154.85 154.71 154.67 154.46 154.38 153.60 153.35 149.00 148.17 147.16 147.07 146.58 54 29.49 l 159.00 162.54 162.54 162.04 161.98 161.81 161.77 161.56 161.53 161.35 161.34 161.30 161.30 161.36 53 29.24 l 160.00 162.58 162.56 162.06 162.00 161.83 161.79 161.58 161.55 161.37 161.35 161.31 161.31 161.37 52 27.59 l 162.85 163.54 163.55 164.57 164.66 164.78 164.80 164.87 164.87 164.80 164.79 164.64 164.61 164.36 51 27.33 l 163.14 163.68 163.68 164.69 164.71 164.82 164.84 164.91 164.91 164.84 164.83 164.69 164.66 164.42 50 26.48 l 163.84 164.10 164.10 164.99 165.01 165.09 165.11 165.18 165.19 165.12 165.11 164.97 164.94 164.73 49 25.86 l 164.25 164.38 164.38 165.20 165.21 165.28 165.30 165.37 165.37 165.31 165.30 165.16 165.13 164.93 48 21.41 l 165.75 165.77 165.77 166.23 166.24 166.29 166.30 166.33 166.33 166.26 166.25 166.13 166.11 165.96 47 21.30 l 165.78 165.80 165.80 166.25 166.26 166.31 166.32 166.35 166.35 166.28 166.26 166.15 166.12 165.97 46 18.33 l 166.58 166.39 166.39 166.53 166.54 166.56 166.56 166.55 166.55 166.44 166.42 166.29 166.27 166.13 45 18.08 l 166.72 166.43 166.43 166.54 166.54 166.56 166.56 166.55 166.55 166.44 166.42 166.29 166.27 166.12 44 16.43 l 168.04 166.72 166.72 166.85 166.86 166.87 166.86 166.76 166.73 166.35 166.31 166.08 166.04 165.83 43 16.17 l 168.54 166.74 166.73 166.85 166.86 166.87 166.87 166.76 166.74 166.35 166.31 166.07 166.04 165.83 42 16.03 l 168.87 169.02 169.08 169.33 169.36 169.43 169.45 169.51 169.50 165.65 165.64 165.36 165.32 165.00 41 15.78 l 168.89 169.04 169.09 169.34 169.37 169.44 169.45 169.52 169.52 165.64 165.64 165.36 165.32 165.00 40 14.13 l 171.18 171.16 171.16 171.06 171.04 170.93 170.87 169.87 169.87 165.79 165.79 165.44 165.43 165.31 39 13.88 l 171.21 171.19 171.18 171.09 171.06 170.96 170.95 169.92 169.92 165.81 165.81 165.44 165.43 165.31 38 9.44 l 171.81 171.79 171.78 171.67 171.64 171.55 171.53 170.48 170.48 165.97 165.97 165.55 165.55 165.45 37 4.88 l 171.71 171.69 171.68 171.58 171.55 171.46 171.45 170.55 170.54 165.79 165.79 165.39 165.38 165.25 36 4.62 l 171.70 171.67 171.67 171.56 171.54 171.44 171.32 170.54 170.54 165.77 165.76 165.39 165.38 165.24 35 2.97 l 170.71 170.73 170.73 170.71 170.70 170.66 170.64 170.51 170.51 165.58 165.58 165.25 165.20 164.82 34 2.72 l 170.71 170.72 170.73 170.71 170.70 170.66 170.64 170.51 170.50 165.58 165.57 165.25 165.20 164.82 33 1.97 l 169.45 168.36 168.35 168.33 168.31 168.25 168.23 167.99 167.95 167.40 167.32 166.82 166.74 166.19 32 1.72 l 169.16 168.35 168.35 168.33 168.31 168.25 168.22 167.99 167.95 167.40 167.33 166.83 166.74 166.19 31.07 l 168.44 168.32 168.32 168.44 168.45 168.45 168.45 168.36 168.34 168.03 167.98 167.65 167.59 167.21 30.00 l 168.42 168.31 168.31 168.44 168.45 168.45 168.45 168.36 168.34 168.03 167.99 167.65 167.59 167.21 29 -.19 l 168.36 168.31 168.31 168.45 168.45 168.46 168.45 168.37 168.34 168.04 167.99 167.66 167.60 167.23 28 -9.19 l 167.60 167.64 167.64 167.94 167.94 167.96 167.96 167.87 167.85 167.53 167.48 167.12 167.06 166.65 27 -9.44 l 167.53 167.60 167.60 167.90 167.92 167.94 167.94 167.86 167.83 167.51 167.46 167.10 167.04 166.61 26 -11.09 l 166.98 167.30 167.29 167.08 167.05 166.92 166.88 166.60 166.55 165.88 165.79 165.23 165.14 164.63 25 -11.34 l 166.83 167.29 167.29 167.08 167.04 166.92 166.88 166.60 166.55 165.88 165.78 165.22 165.13 164.62 24 -11.40 l 166.79 167.21 167.21 167.00 166.96 166.83 166.78 166.43 166.34 164.92 164.83 164.31 164.23 163.74 23 -11.96 l 166.34 166.52 166.51 166.27 166.21 165.94 165.83 164.76 164.36 155.38 155.36 155.32 155.30 155.05 22 -12.21 l 166.12 166.22 166.21 165.95 165.88 165.56 165.43 164.12 163.66 155.35 155.35 155.32 155.29 155.05 21 -13.86 l 164.54 164.48 164.44 164.04 163.93 163.46 163.28 161.56 161.14 155.06 155.06 154.62 154.58 154.45 20 -14.12 l 164.30 164.23 164.20 163.79 163.68 163.21 163.03 161.33 160.92 155.02 155.02 154.58 154.57 154.45 19 -16.64 l 162.24 162.18 162.15 161.96 161.92 161.76 161.71 161.32 161.30 154.70 154.70 154.36 154.35 154.28 18 -16.89 l 162.24 162.17 162.15 161.95 161.92 161.76 161.71 161.31 161.30 154.67 154.67 154.33 154.33 154.25 17 -18.55 l 161.96 161.93 161.92 161.81 161.79 161.70 161.67 160.95 160.95 154.42 154.42 154.11 154.11 154.05 16 -18.80 l 161.96 161.93 161.92 161.81 161.79 161.70 161.69 160.90 160.90 154.37 154.37 154.07 154.07 154.01 15 -23.11 l 161.60 161.57 161.56 161.43 161.41 161.31 161.29 160.13 160.13 153.14 153.14 153.17 153.17 153.12 14 -23.37 l 161.56 161.53 161.51 161.39 161.37 161.26 161.25 160.08 160.07 153.03 153.03 153.10 153.14 153.09 13 -25.02 l 161.25 161.22 161.20 161.08 161.05 160.94 160.93 159.73 159.72 152.20 152.18 151.49 151.40 150.93 12 -25.27 l 161.19 161.16 161.15 161.02 161.00 160.89 160.87 159.67 159.67 152.18 152.16 151.49 151.39 150.92 11 -25.86 l 161.06 161.03 161.01 160.89 160.86 160.75 160.73 159.53 159.53 152.15 151.76 150.17 149.98 149.04 10 -26.48 l 160.90 160.87 160.86 160.73 160.70 160.59 160.57 159.35 159.35 151.64 151.09 148.69 148.40 147.03 9 -27.80 l 160.53 160.50 160.49 160.36 160.33 160.20 160.18 158.88 158.88 149.42 148.67 145.14 144.71 142.72 8 -28.05 l 160.49 160.46 160.45 160.32 160.29 160.14 159.87 158.77 158.77 148.80 148.02 144.35 143.91 141.87 7 -29.70 l 157.52 157.53 157.53 157.58 157.60 157.67 157.69 157.99 158.00 143.34 142.31 138.46 138.02 136.18

+----------------------------------------------------------------------------------------------------------------

.00 1.46 1.71 3.36 3.62 4.50 4.75 6.41 6.66 8.96 9.21 10.87 11.12 12.62 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Figure 8.1-5. NUHOMS-12T DSC Internal Temperature Distribution for 87F Ambient.

Idaho Cleanup Project 412.09 (01/01/2022 - Rev. 13)

SAFETY ANALYSIS REPORT FOR TMI-2 ANALYSIS OF DESIGN EVENTS Identifier:

Revision*:

Page:

SAR-II-8.4 CH 8 Rev. 6 8-47 of 8-98 TMI-2 Fuel Debris Can TMI87.inp, normal at 87F case,horiz with air Thu Aug 8 15:19:10 1996 Steady-State Temperature Distribution at Time 0.0000E+00 59 33.62 l 58 33.00 l 57 31.85 l 134.20 134.20 134.20 56 31.23 l 134.24 134.22 134.20 134.20 55 30.41 l 146.19 139.87 134.22 134.20 134.20 54 29.49 l 161.38 161.65 161.62 134.20 134.20 53 29.24 l 161.39 161.68 161.69 134.20 134.20 52 27.59 l 164.03 162.74 162.73 134.20 134.20 51 27.33 l 164.39 162.88 162.88 134.20 134.20 50 26.48 l 164.71 163.33 163.32 134.20 134.20 134.20 134.20 134.20 49 25.86 l 164.92 163.62 163.61 134.25 134.24 134.23 134.21 134.18 134.20 48 21.41 l 165.94 165.06 165.06 148.42 147.77 143.76 143.11 132.57 132.54 131.00 47 21.30 l 165.96 165.09 165.08 148.73 148.08 144.04 143.39 132.57 132.54 131.00 131.00 46 18.33 l 166.11 165.49 165.49 155.84 155.35 152.03 151.45 140.66 139.20 131.00 131.00 45 18.08 l 166.09 165.51 165.51 156.47 156.00 152.85 152.28 141.22 139.67 131.00 131.00 44 16.43 l 165.80 165.58 165.58 160.58 160.33 158.68 158.30 144.31 142.34 131.00 131.00 43 16.17 l 165.80 165.58 165.56 161.22 161.00 159.72 159.53 144.70 142.68 131.00 131.00 42 16.03 l 164.93 164.45 164.36 161.58 161.38 160.30 160.27 144.91 142.87 131.00 131.00 131.00 41 15.78 l 164.93 164.45 164.36 161.58 161.39 160.30 160.28 145.24 143.17 131.23 131.02 131.00 40 14.13 l 165.27 164.99 164.94 163.18 163.03 160.06 160.05 146.73 144.68 131.71 131.09 131.00 39 13.88 l 165.28 165.00 164.95 163.20 163.15 160.04 160.02 146.90 144.86 131.76 131.09 131.00 38 9.44 l 165.42 165.17 165.13 163.53 163.48 159.92 159.91 147.80 145.82 131.93 131.11 131.00 131.01 131.01 37 4.88 l 165.22 164.93 164.87 163.08 163.04 159.83 159.82 146.40 144.53 134.60 134.14 131.43 131.02 131.01 36 4.62 l 165.21 164.92 164.86 163.06 162.90 159.84 159.83 146.09 144.24 134.67 134.22 131.45 131.02 131.01 35 2.97 l 164.75 164.20 164.11 161.22 161.04 159.97 159.95 143.40 141.82 135.00 134.58 131.60 131.02 131.01 34 2.72 l 164.75 164.20 164.10 161.21 161.03 159.97 159.92 142.75 141.35 135.08 134.67 131.75 131.02 131.01 33 1.97 l 166.08 165.32 165.30 144.35 144.28 143.23 143.04 140.33 139.81 135.34 135.01 132.65 132.50 131.04 32 1.72 l 166.08 165.31 165.29 144.30 144.28 143.22 143.03 140.33 139.82 135.34 135.01 132.62 132.55 131.04 31.07 l 167.10 164.92 164.91 144.43 144.42 142.17 142.05 140.67 140.37 137.53 137.26 131.99 131.96 131.02 30.00 l 167.13 164.91 164.89 144.44 144.43 142.14 142.05 140.67 140.37 137.53 137.33 131.97 131.94 131.02 29 -.19 l 167.20 164.87 164.86 144.45 144.44 142.07 142.04 140.67 140.38 137.56 137.48 131.92 131.89 131.02 28 -9.19 l 166.62 164.05 164.04 144.03 144.02 141.87 141.84 140.55 140.26 137.51 137.43 131.93 131.91 131.02 27 -9.44 l 166.48 164.03 164.02 143.99 143.98 141.94 141.84 140.54 140.25 137.48 137.23 132.01 131.99 131.02 26 -11.09 l 164.54 163.93 163.92 143.62 143.60 142.58 142.41 140.01 139.54 135.32 135.00 132.76 132.69 131.04 25 -11.34 l 164.53 163.93 163.89 143.64 143.59 142.58 142.41 140.01 139.54 135.31 135.00 132.79 132.70 131.03 24 -11.40 l 163.65 163.07 163.00 144.42 144.34 143.35 143.19 140.07 139.55 135.12 134.67 132.57 131.10 131.01 23 -11.96 l 155.00 154.48 154.38 151.82 151.68 150.86 150.82 140.65 139.70 133.33 131.33 131.29 131.01 131.00 22 -12.21 l 154.99 154.48 154.38 151.83 151.69 150.85 150.84 140.84 139.76 132.79 131.33 131.29 21 -13.86 l 154.42 154.17 154.13 152.70 152.59 150.58 150.57 141.33 140.00 131.84 131.27 131.24 20 -14.12 l 154.42 154.17 154.13 152.71 152.67 150.54 150.53 141.34 140.00 131.75 131.25 131.22 19 -16.64 l 154.25 154.06 154.02 152.76 152.73 150.23 150.22 140.81 139.42 131.10 131.00 131.00 18 -16.89 l 154.23 154.04 154.00 152.75 152.72 150.19 150.18 140.68 139.29 131.17 131.16 17 -18.55 l 154.03 153.85 153.81 152.60 152.56 149.94 149.93 139.53 138.18 131.71 131.70 16 -18.80 l 153.99 153.81 153.78 152.56 152.53 149.90 149.89 139.26 137.93 131.75 131.75 15 -23.11 l 153.10 152.92 152.88 151.56 151.52 148.89 148.88 130.68 130.66 131.00 131.00 14 -23.37 l 153.08 152.88 152.85 151.52 151.36 148.81 148.80 130.67 130.66 131.00 13 -25.02 l 150.85 150.37 150.30 148.70 148.64 148.31 148.30 130.47 130.45 12 -25.27 l 150.84 150.36 150.29 148.69 148.63 148.29 148.25 130.42 130.41 11 -25.86 l 148.89 147.86 147.61 131.39 131.35 131.20 131.15 130.30 130.30 10 -26.48 l 146.80 145.32 145.01 131.30 131.31 131.18 131.14 130.30 9 -27.80 l 142.40 140.41 140.08 130.97 130.97 8 -28.05 l 141.54 139.49 139.15 130.91 130.91 7 -29.70 l 135.86 133.62 133.23 130.48 130.48

+----------------------------------------------------------------------------------------------------------------

12.87 14.52 14.77 20.12 20.38 22.02 22.27 25.86 26.48 31.27 31.53 33.18 33.43 33.44 15 16 17 18 19 20 21 22 23 24 25 26 27 28 Figure 8.1-5. NUHOMS-12T DSC Internal Temperature Distribution for 87F Ambient. (Continued)

Idaho Cleanup Project 412.09 (01/01/2022 - Rev. 13)

SAFETY ANALYSIS REPORT FOR TMI-2 ANALYSIS OF DESIGN EVENTS Identifier:

Revision*:

Page:

SAR-II-8.4 CH 8 Rev. 6 8-48 of 8-98 TMI-2 Fuel Debris Can TMI87.inp, normal at 87F case,horiz with air Thu Aug 8 15:19:10 1996 Steady-State Temperature Distribution at Time 0.0000E+00 59 33.62 l 58 33.00 l 57 31.85 l 134.20 134.20 134.20 56 31.23 l 134.24 134.22 134.20 134.20 55 30.41 l 146.19 139.87 134.22 134.20 134.20 54 29.49 l 161.38 161.65 161.62 134.20 134.20 53 29.24 l 161.39 161.68 161.69 134.20 134.20 52 27.59 l 164.03 162.74 162.73 134.20 134.20 51 27.33 l 164.39 162.88 162.88 134.20 134.20 50 26.48 l 164.71 163.33 163.32 134.20 134.20 134.20 134.20 134.20 49 25.86 l 164.92 163.62 163.61 134.25 134.24 134.23 134.21 134.18 134.20 48 21.41 l 165.94 165.06 165.06 148.42 147.77 143.76 143.11 132.57 132.54 131.00 47 21.30 l 165.96 165.09 165.08 148.73 148.08 144.04 143.39 132.57 132.54 131.00 131.00 46 18.33 l 166.11 165.49 165.49 155.84 155.35 152.03 151.45 140.66 139.20 131.00 131.00 45 18.08 l 166.09 165.51 165.51 156.47 156.00 152.85 152.28 141.22 139.67 131.00 131.00 44 16.43 l 165.80 165.58 165.58 160.58 160.33 158.68 158.30 144.31 142.34 131.00 131.00 43 16.17 l 165.80 165.58 165.56 161.22 161.00 159.72 159.53 144.70 142.68 131.00 131.00 42 16.03 l 164.93 164.45 164.36 161.58 161.38 160.30 160.27 144.91 142.87 131.00 131.00 131.00 41 15.78 l 164.93 164.45 164.36 161.58 161.39 160.30 160.28 145.24 143.17 131.23 131.02 131.00 40 14.13 l 165.27 164.99 164.94 163.18 163.03 160.06 160.05 146.73 144.68 131.71 131.09 131.00 39 13.88 l 165.28 165.00 164.95 163.20 163.15 160.04 160.02 146.90 144.86 131.76 131.09 131.00 38 9.44 l 165.42 165.17 165.13 163.53 163.48 159.92 159.91 147.80 145.82 131.93 131.11 131.00 131.01 131.01 37 4.88 l 165.22 164.93 164.87 163.08 163.04 159.83 159.82 146.40 144.53 134.60 134.14 131.43 131.02 131.01 36 4.62 l 165.21 164.92 164.86 163.06 162.90 159.84 159.83 146.09 144.24 134.67 134.22 131.45 131.02 131.01 35 2.97 l 164.75 164.20 164.11 161.22 161.04 159.97 159.95 143.40 141.82 135.00 134.58 131.60 131.02 131.01 34 2.72 l 164.75 164.20 164.10 161.21 161.03 159.97 159.92 142.75 141.35 135.08 134.67 131.75 131.02 131.01 33 1.97 l 166.08 165.32 165.30 144.35 144.28 143.23 143.04 140.33 139.81 135.34 135.01 132.65 132.50 131.04 32 1.72 l 166.08 165.31 165.29 144.30 144.28 143.22 143.03 140.33 139.82 135.34 135.01 132.62 132.55 131.04 31.07 l 167.10 164.92 164.91 144.43 144.42 142.17 142.05 140.67 140.37 137.53 137.26 131.99 131.96 131.02 30.00 l 167.13 164.91 164.89 144.44 144.43 142.14 142.05 140.67 140.37 137.53 137.33 131.97 131.94 131.02 29 -.19 l 167.20 164.87 164.86 144.45 144.44 142.07 142.04 140.67 140.38 137.56 137.48 131.92 131.89 131.02 28 -9.19 l 166.62 164.05 164.04 144.03 144.02 141.87 141.84 140.55 140.26 137.51 137.43 131.93 131.91 131.02 27 -9.44 l 166.48 164.03 164.02 143.99 143.98 141.94 141.84 140.54 140.25 137.48 137.23 132.01 131.99 131.02 26 -11.09 l 164.54 163.93 163.92 143.62 143.60 142.58 142.41 140.01 139.54 135.32 135.00 132.76 132.69 131.04 25 -11.34 l 164.53 163.93 163.89 143.64 143.59 142.58 142.41 140.01 139.54 135.31 135.00 132.79 132.70 131.03 24 -11.40 l 163.65 163.07 163.00 144.42 144.34 143.35 143.19 140.07 139.55 135.12 134.67 132.57 131.10 131.01 23 -11.96 l 155.00 154.48 154.38 151.82 151.68 150.86 150.82 140.65 139.70 133.33 131.33 131.29 131.01 131.00 22 -12.21 l 154.99 154.48 154.38 151.83 151.69 150.85 150.84 140.84 139.76 132.79 131.33 131.29 21 -13.86 l 154.42 154.17 154.13 152.70 152.59 150.58 150.57 141.33 140.00 131.84 131.27 131.24 20 -14.12 l 154.42 154.17 154.13 152.71 152.67 150.54 150.53 141.34 140.00 131.75 131.25 131.22 19 -16.64 l 154.25 154.06 154.02 152.76 152.73 150.23 150.22 140.81 139.42 131.10 131.00 131.00 18 -16.89 l 154.23 154.04 154.00 152.75 152.72 150.19 150.18 140.68 139.29 131.17 131.16 17 -18.55 l 154.03 153.85 153.81 152.60 152.56 149.94 149.93 139.53 138.18 131.71 131.70 16 -18.80 l 153.99 153.81 153.78 152.56 152.53 149.90 149.89 139.26 137.93 131.75 131.75 15 -23.11 l 153.10 152.92 152.88 151.56 151.52 148.89 148.88 130.68 130.66 131.00 131.00 14 -23.37 l 153.08 152.88 152.85 151.52 151.36 148.81 148.80 130.67 130.66 131.00 13 -25.02 l 150.85 150.37 150.30 148.70 148.64 148.31 148.30 130.47 130.45 12 -25.27 l 150.84 150.36 150.29 148.69 148.63 148.29 148.25 130.42 130.41 11 -25.86 l 148.89 147.86 147.61 131.39 131.35 131.20 131.15 130.30 130.30 10 -26.48 l 146.80 145.32 145.01 131.30 131.31 131.18 131.14 130.30 9 -27.80 l 142.40 140.41 140.08 130.97 130.97 8 -28.05 l 141.54 139.49 139.15 130.91 130.91 7 -29.70 l 135.86 133.62 133.23 130.48 130.48

+----------------------------------------------------------------------------------------------------------------

12.87 14.52 14.77 20.12 20.38 22.02 22.27 25.86 26.48 31.27 31.53 33.18 33.43 33.44 15 16 17 18 19 20 21 22 23 24 25 26 27 28 Figure 8.1-5. NUHOMS-12T DSC Internal Temperature Distribution for 87F Ambient. (Continued)

Idaho Cleanup Project 412.09 (01/01/2022 - Rev. 13)

SAFETY ANALYSIS REPORT FOR TMI-2 ANALYSIS OF DESIGN EVENTS Identifier:

Revision*:

Page:

SAR-II-8.4 CH 8 Rev. 6 8-49 of 8-98 TMI-2 Fuel Debris Can TMI87.inp, normal at 87F case,horiz with air Thu Aug 8 15:19:10 1996 Steady-State Temperature Distribution at Time 0.0000E+00 6 -29.95 l 157.51 157.52 157.52 157.57 157.58 157.65 157.68 157.96 157.96 142.08 141.00 137.40 137.00 135.32 5 -30.41 l 154.01 154.00 154.00 153.97 153.95 153.89 153.86 153.31 152.85 139.45 138.23 135.36 135.06 133.77 4 -31.23 l 147.72 147.68 147.67 147.49 147.43 147.19 147.07 145.76 145.20 134.33 131.54 131.41 131.37 131.05 3 -31.85 l 142.88 142.82 142.80 142.53 142.45 142.13 142.00 140.60 140.15 132.64 131.52 131.39 131.36 131.03 2 -33.00 l 133.96 133.87 133.83 133.45 133.37 133.04 132.93 132.08 131.92 130.46 130.30 1 -33.62 l 133.95 133.85 133.81 133.43 133.35 133.02 132.91 132.06 131.91 130.46 130.30

+----------------------------------------------------------------------------------------------------------------

.00 1.46 1.71 3.36 3.62 4.50 4.75 6.41 6.66 8.96 9.21 10.87 11.12 12.62 1 2 3 4 5 6 7 8 9 10 11 12 13 14 TMI-2 Fuel Debris Can TMI87.inp, normal at 87F case,horiz with air Thu Aug 8 15:19:10 1996 Steady-State Temperature Distribution at Time 0.0000E+00 6 -29.95 l 135.02 132.78 132.24 130.42 130.42 5 -30.41 l 133.51 131.37 130.34 130.30 130.30 4 -31.23 l 130.98 130.41 130.30 130.30 3 -31.85 l 130.96 130.40 130.30 2 -33.00 l 1 -33.62 l

+----------------------------------------------------------------------------------------------------------------

12.87 14.52 14.77 20.12 20.38 22.02 22.27 25.86 26.48 31.27 31.53 33.18 33.43 33.44 15 16 17 18 19 20 21 22 23 24 25 26 27 28 TMI-2 Fuel Debris Can TMI87.inp, normal at 87F case,horiz with air Thu Aug 8 15:19:10 1996 Steady-State Temperature Distribution at Time 0.0000E+00 6 -29.95 l 5 -30.41 l 4 -31.23 l 3 -31.85 l 2 -33.00 l 1 -33.62 l

+--------

33.62 29 Figure 8.1-5. NUHOMS-12T DSC Internal Temperature Distribution for 87F Ambient. (Concluded)

Idaho Cleanup Project 412.09 (01/01/2022 - Rev. 13)

SAFETY ANALYSIS REPORT FOR TMI-2 ANALYSIS OF DESIGN EVENTS Identifier:

Revision*:

Page:

SAR-II-8.4 CH 8 Rev. 6 8-50 of 8-98 Figure 8.1-6. DSC Shell Stress Analysis Diagram.

Idaho Cleanup Project 412.09 (01/01/2022 - Rev. 13)

SAFETY ANALYSIS REPORT FOR TMI-2 ANALYSIS OF DESIGN EVENTS Identifier:

Revision*:

Page:

SAR-II-8.4 CH 8 Rev. 6 8-51 of 8-98 Figure 8.1-7. DSC Shell Axisymmetric Analytical Model.

Idaho Cleanup Project 412.09 (01/01/2022 - Rev. 13)

SAFETY ANALYSIS REPORT FOR TMI-2 ANALYSIS OF DESIGN EVENTS Identifier:

Revision*:

Page:

SAR-II-8.4 CH 8 Rev. 6 8-52 of 8-98 Figure 8.1-8. DSC Bottom Cover Plate/Grapple Ring Analytical Model.

Idaho Cleanup Project 412.09 (01/01/2022 - Rev. 13)

SAFETY ANALYSIS REPORT FOR TMI-2 ANALYSIS OF DESIGN EVENTS Identifier:

Revision*:

Page:

SAR-II-8.4 CH 8 Rev. 6 8-53 of 8-98 Figure 8.1-9. NUHOMS-12T DSC Spacer Disc Analytical Model.

Idaho Cleanup Project 412.09 (01/01/2022 - Rev. 13)

SAFETY ANALYSIS REPORT FOR TMI-2 ANALYSIS OF DESIGN EVENTS Identifier:

Revision*:

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SAR-II-8.4 CH 8 Rev. 6 8-54 of 8-98 COMPONENT DEAD LOAD 8g LOAD(2)

FP, Spacer Disk 495 lb 3.96 k FF, TMI-2 Canisters 21.78 lb/in 5096 lb (1) 40.77 k FRGS, Support Rod 1.29 lb/in 151 lb 1.21 k FApplied, Total Applied Load 45.94 k FReaction, ANSYS Reaction Load 44.30 k Ratio (FPApplied / FReaction) 1.04 (1)

Dead load is calculated for 12 heaviest TMI-2 canisters, maximum spacer disc spacing of 19.5 inches.

(2)

Represents worst case transportation/transfer loads applied in 0-180 degree plane.

Figure 8.1-10. NUHOMS-12T DSC Spacer Disc Applied 8g Loading.

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SAR-II-8.4 CH 8 Rev. 6 8-55 of 8-98 Figure 8.1-11. DSC Axial Jam Condition.

Figure 8.1-12. DSC Binding (Pinching) Condition.

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SAR-II-8.4 CH 8 Rev. 6 8-56 of 8-98 Figure 8.1-13. Prefabricated HSM Analytical Model.

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SAR-II-8.4 CH 8 Rev. 6 8-57 of 8-98 Figure 8.1-14. Analytical Model for DSC Support Structure.

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SAR-II-8.4 CH 8 Rev. 6 8-58 of 8-98 Figure 8.1-15. Analytical Model for HSM Base Unit Thermal Conditions.

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SAR-II-8.4 CH 8 Rev. 6 8-60 of 8-98 Figure 8.1-17. MP187 Cask Handling Loads.

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SAR-II-8.4 CH 8 Rev. 6 8-61 of 8-98 8.2 Accident Analyses for the ISFSI The design basis accident events specified by ANSI/ANS 57.9-1984, and other credible and non-credible accidents postulated to affect the normal safe operation of the TMI-2 ISFSI are addressed in this section of the SAR. In accordance with 10 CFR Part 72, analyses are provided for a range of hypothetical accidents, including those with the potential to result in an annual dose greater than 25 mrem outside the owner-controlled area. The postulated accidents considered in the analyses and the associated NUHOMS-12T components affected by each accident condition are shown in Table 8.2-1.

Based on a review of the Safety Analysis Reports of other facility activities within the INTEC site, it has been determined that no credible explosion or fire associated with a co-located INTEC facility could occur that would pose a threat to the ISFSI which either exceeds a vehicular fire related to an ISFSI service vehicle (Section 8.2.9), or exceed the potential impacts of either the wind loading, or airborne missile impact of a tornado scenario (Section 8.2.2). Thus, the impacts of any credible accidents involving fire or explosion at co-located INTEC facilities are bounded by the analysis of the design basis tornado and the combustion of fuel from an ISFSI service vehicle.

In the following sections, each accident condition is analyzed to demonstrate that the applicable requirements of 10 CFR 72.122 are met and that adequate safety margins exist for the TMI-2 ISFSI design.

Radiological calculations are performed to confirm that on-site and off-site dose rates are within acceptable limits. The resulting accident condition stresses in the NUHOMS-12T system components are evaluated and compared with the applicable code limits set forth in Section 3.2. Where appropriate, these accident condition stresses are combined with those of normal operating loads in accordance with the load combination definitions in Tables 3.2-4, 3.2-5, and 3.2-6.

The postulated accident conditions addressed in this SAR section include:

1.

Reduced HSM self-shielding

2.

Tornado winds and tornado generated missiles

3.

Design basis earthquake

4.

Design basis flood

5.

Accidental cask drop

6.

Lightning effects

7.

Postulated DSC leakage

8.

DSC pressurization

9.

On-site fire and explosion hazards

10.

Blockage of gap between adjacent modules.

For each postulated accident condition, the cause, the consequences (structural, thermal, radiological),

and the recovery measures are presented.

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SAR-II-8.4 CH 8 Rev. 6 8-62 of 8-98 8.2.1 Reduced HSM Self-Shielding This postulated accident is the partial loss of self-shielding for the HSM side walls provided by the adjacent HSMs. All other components of the NUHOMS-12T system are assumed to be functioning normally.

8.2.1.1 Cause of Accident. An array of free-standing prefabricated HSMs is designed to remain intact for all postulated events. In the unlikely event that large settlements of the ISFSI foundation occur, or a seismic event displaces individual HSMs, the resultant shifting of adjacent HSMs may cause the HSMs to separate. Although adequate means are provided to ensure that the spacing of adjacent HSMs is maintained, it is postulated for the sake of this conservative analysis that an HSM in the middle of the array is separated.

The HSM now rests against the adjacent HSM side wall, moves forward/backward, and/or projects out from the face of adjacent HSMs. The arrangement will increase the distance between the HSM on the opposite side from the nominal spacing of six inches between HSMs, or result in HSMs staggered in plan from adjacent modules.

8.2.1.2 Accident Analysis. A separation of the HSMs will not result in any structural consequences that will affect the safe operation of the NUHOMS 12T system. Potential thermal effects of this accident result from the HSM side walls coming in contact with each other or exposure of additional side walls due to the longitudinal displacement. For the thermal accident case considered, the HSMs were analyzed assuming fully insulated boundaries on both side walls, and the accident calculations presented in Section 8.1.3 bound the thermal consequences of this accident for the HSM, DSC, and TMI-2 canister maximum temperatures.

The radiological consequences of this accident are described in the paragraph below.

8.2.1.3 Accident Dose Calculations. The off-site radiological effects resulting from a partial loss of adjacent HSM shielding are a very small increase in the air scattered (skyshine), and direct doses between separated HSMs. On-site radiological effects result from an increase in direct radiation during recovery operations and increased skyshine radiation. The calculation of these doses during normal conditions is described in Section 7.3.2.2.

The peak dose rate expected for any surface location of an uncovered interior wall is about 270 mrem per hour. This represents an external dose only, as no radionuclide releases will occur as a result of this event. The actual exposure received by workers during the recovery from this event is dependent on the final configuration of the HSMs. Conservatively assuming two workers spend an hour at the highest dose location, the resultant dose will be 0.54 person-rem. Exposure increases due to the reduced shielding at locations outside the INTEC will be negligible and no increase in exposure at locations beyond the INL controlled area boundary will be observed.

8.2.1.4 Recovery. Recovery from an accident resulting in a partial loss of adjacent HSM shielding requires repositioning of the HSMs. This can be accomplished by using hydraulic jacks or by reinstalling the HSM lifting eyes and repositioning the affected HSMs with a suitable crane. As stated above, an exposure of 0.54 person-rem is conservatively estimated for workers during recovery from this event. Severe foundation settlement requires that the affected HSMs be unloaded and removed from service until foundation repairs are made.

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SAR-II-8.4 CH 8 Rev. 6 8-63 of 8-98 8.2.2 Tornado Winds/Tornado Missile 8.2.2.1 Cause of Accident. In accordance with ANSI-57.9 and 10 CFR 72.122, the TMI-2 ISFSI NUHOMS-12T HSM is designed for tornado effects, including tornado wind loads. Although not specifically required by ANSI-57.9 and 10 CFR 72.122, the HSM is also designed for tornado missile effects. For this evaluation, the NUHOMS-12T system is designed to be located on the INL in the State of Idaho. As described in Section 3.2.1, the design basis for this postulated accident is provided by the Region III tornado wind loadings specified by NUREG-0800 [8.4] and NRC Regulatory Guide 1.76 [8.22]

modified in accordance with References 8.23 and 8.24.

8.2.2.2 Accident Analysis. The applicable design parameters for the design basis tornado (DBT) are specified in Section 3.2.1. The tornado wind and tornado missile loads acting on the HSM are detailed in Section 3.2.1.4. The front and rear faces of all modules and the side walls of the end modules of an array resist the tornado wind and missile loads. The pressures used for the stability evaluations of the HSM conservatively used a value of 240 mph rather than apply the Region III tornado wind speed reduction of 240 mph to 200 mph permitted by References 8.25 and 8.26. The missile impact analysis in Section 8.2.2.8 uses the 200 mph allowed by References 8.23 and 2.24. The tornado loads are assumed to act on a single free-standing HSM. This case conservatively envelopes the effects on an HSM array.

A.

Effect of DBT Wind Pressure Loads on HSM The maximum DBT generated wind loads for a 200 mph maximum wind speed are 99 lb/ft2 and 62 lb/ft2 on the windward and leeward walls of an HSM. The exterior side wall and roof are also subjected to a suction load of 87 lb/ft2. The leeward side of the windward end HSM has no appreciable suction load in an array of HSMs due to the close proximity of the adjacent module. The 62 lb/ft2 suction load is applicable to the exposed side wall of the leeward end module in the array. In addition, the HSM is subjected to a differential wind pressure load of 1.5 psi (216 psf) suction on the walls and the roof.

The DBT pressures are applied to the HSM as uniformly distributed loads on the walls and roof. The rigidity of the HSM in the transverse direction (frame and shear wall action of a single HSM) is the primary load transfer mechanism assumed in the analysis. The bending moments and shears at critical locations in the HSM walls and roof are calculated by performing an analysis using the ANSYS analytical model of the prefabricated HSM shown in Figure 8.1-13 and Table 8.1-15. The resulting moments and shears are tabulated in Table 8.2-2 and are included in the formulation of HSM load combination results reported in Section 8.3.5.

An analysis is also performed to evaluate the effects of overturning and sliding of a single, free-standing HSM due to a postulated DBT. For the DBT wind overturning analysis, the overturning moment and the resulting stabilizing moments are calculated.

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SAR-II-8.4 CH 8 Rev. 6 8-64 of 8-98 (i)

HSM Overturning Analysis The stabilizing moment (Mst) for the windward module plus shield wall in an array is:

Mst

=

Wd Where:

W

=

302 kips, Weight of HSM plus minimum weight of DSC d

=

61.5 in. (5.13 ft), Horizontal distance between center of gravity of HSM to the outer edge of the side wall.

Therefore:

Mst

=

18,570 K-in and the overturning moment (Mto) for the windward module due to DBT wind pressure is:

Mto

=

w1Awh/2 + w3Ard Where:

w1

=

(99 - 216 + 62+216) = 161 lb/ft2, wind load, windward wall h

=

14.5 ft wall height w3

=

(87 + 216) = 303 lb/ft2, wind uplift on roof Ar

=

186 ft2, Roof area Aw

=

263 ft2, wall area Therefore:

Mto

=

7,159 K-in Since the overturning moment is significantly smaller than the stabilizing moment, the free-standing HSM will not overturn. The resulting factor of safety against overturning effects for DBT wind loads is 2.6.

The (87 + 216) = 303 lb/ft2 DBT negative pressure acting on the HSM door results in a total load of 12.8 kips, which is reacted by the HSM door embedded anchors. The door embedded anchors have a tensile load capacity that provides a large factor of safety based on this total load.

(ii)

HSM Sliding Analysis To evaluate the potential for sliding of a single, free-standing HSM, the sliding force generated by the postulated DBT wind pressure is compared to the sliding resistance provided by friction between the base of the HSM and the ISFSI basemat.

The force (Fsl) required to slide the end module in an array is:

Fsl =

[W - w3Ar]

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SAR-II-8.4 CH 8 Rev. 6 8-65 of 8-98 Where:

=

0.6, coefficient of friction (ACI 349) [8.5]

W, w3 and Ar are defined above.

Substituting gives:

Fsl =

147 kips The sliding force (Fhw) generated by DBT wind pressure for a single HSM is:

Fhw =

w1Aw Where: w1 and Aw are as defined above. Substituting gives:

Fhw =

43 kips Since the horizontal force generated by the postulated DBT is smaller than the force required to slide the end module in an HSM array, the HSM will not slide. The factor of safety against sliding of the HSM due to DBT wind loads is 3.4.

B.

HSM Missile Impact Analysis The side walls and roof slab of the reinforced concrete HSM are 24 and 30 inches thick, respectively.

The walls and roof are designed to provide adequate biological shielding and easily meet the minimum acceptable barrier thickness requirements for local damage against tornado generated missiles, as specified for Region III in NUREG-0800, Section 3.5.3, Table 1. To demonstrate the adequacy of the HSM design for tornado missiles, a bounding analysis of the end module in an array is performed. The items evaluated include the resistance to penetration, spalling, scabbing, and perforation for a postulated missile impact. In addition, the tornado missile loads are individually applied to the roof and walls of the HSM analytical model shown in Figure 8.1-13. The resultant tornado missile forces and moments are combined with other loads as required by Table 3.2-4 and the results are reported in Table 8.3-4. For these analyses, a rigid, penetration-resistant missile consisting of a 125 Kg (276 lb), 8-in.-diameter blunt nosed hardened steel object traveling at 35% of the maximum wind speed (103 fps) is conservatively postulated. The method of analysis is based on the modified National Defense Research Committee (NDRC) formula as recommended in Section 3.5.3 of NUREG-0800.

The doors covering the access openings of the HSM are also evaluated for DBT missile penetration resistance. The HSM front wall access door is constructed of steel plate, and is filled with concrete.

The rear door covering the HEPA filter DSC vent area is fabricated from steel plate. Missile impact is resisted by the steel plate. The results of these evaluations indicate that the HSM access and HEPA filter enclosure doors provide sufficient capacity to preclude penetration.

The DBT missile penetration resistance analyses for the HSM are presented in the following paragraphs.

(i)

Missile Impact Penetration Resistance Analysis

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SAR-II-8.4 CH 8 Rev. 6 8-66 of 8-98 The modified NDRC formula from Kennedy, Holmes and Narver [8.25] is used to predict the HSM wall penetration depth for a postulated DBT missile.

When:

x d 2

2.0 Where

x

=

Total penetration depth (in.)

d

=

8 in., Projectile diameter K

=

180/

fc 3, Concrete penetrability factor

=

2.55 for

fc 4= 5000 N

=

0.84 (blunt nosed), Projectile shape factor

fc

=

5000 psi, Concrete compressive design strength at 150 F W

=

Projectile weight = 276 lb V

=

Striking velocity = 103 ft/s Therefore:

x

=

2.7 inches and x d = 0 34 The perforating thickness, or maximum thickness that the postulated DBT missile will completely penetrate, is calculated using the correlation:

e d = 3.19 x d

0.718 x d

2 135 for x d

Substituting yields:

e

=

8.07 in Therefore, e, the maximum perforation thickness, is conservative.

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SAR-II-8.4 CH 8 Rev. 6 8-67 of 8-98 The minimum thickness necessary to prevent scabbing of material from the rear face of the target is calculated using:

s d = 7.91 x d

5.06 x d

2 065 for x d

Substituting yields:

s

=

16.9 in.

Where:

s

=

Scabbing thickness (in.)

Scabbing effects control the minimum required wall thickness. ACI 349 requirements for nuclear safety related concrete structures require a minimum of 20% additional wall thickness to prevent perforation and scabbing. Therefore, the minimum wall thickness required to provide complete protection for the enveloping DBT missile is:

1.2s

=

20.3 in.

The specified minimum wall thickness for the HSM side walls is 24 inches. To provide complete protection for the DSC shell, 12-inch thick shield walls are added to the end module exposed side walls in each row (a total of four walls). The addition of these walls provides a minimum concrete thickness of 27 inches behind the DSC (rear wall), 30 inches (front wall), and 36 inches for exposed side walls. Consequently, there is adequate protection against local DBT missile impact damage.

(ii)

Local Barrier Impingement Analysis A composite door comprising a steel plate and concrete is used to cover the HSM front wall access opening. The HSM front wall access and HEPA filter vent doors are analyzed to verify their adequacy for local barrier impingement of a DBT missile. The 276-pound, 8-in. diameter missile is used for this calculation as it envelopes effects caused by all other missiles. The minimum thickness of a steel plate that can be perforated by the postulated DBT missile is given in the McDonalds, Mehta, and Minor paper [8.26] as:

T

=

(0.5M V )

672d 2/3 m

s2 m

5 = 0.24 in.

Where:

T

=

Perforation thickness (in.)

Mm

=

Mass of missile = W g 6= 8.57 slugs W

=

Weight = 276 lb g

=

32.2 ft/s2 Vs

=

Missile strike velocity = 103 ft/s dm

=

Diameter of missile = 8 in.

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SAR-II-8.4 CH 8 Rev. 6 8-68 of 8-98 The specified 1-1/2-in. thick steel HSM access and HEPA filter access rear doors easily exceed the minimum required perforation thickness of (1.25 x 0.24) = 0.3 inches.

(iii)

Massive Missile Impact Analysis The HSM stability and potential damage due to impact of the postulated DBT massive missile consisting of a 3,990 lb. automobile, 20 sq. ft. frontal area traveling at 103 ft/sec., is evaluated. The massive missile is assumed to impact the side wall of an end module in an array. The shield wall is conservatively ignored. Using the principles of conservation of momentum with a coefficient of restitution of zero, the analysis presented below demonstrates that the end module remains stable and the missile energy is dissipated by sliding or slight tipping of the module.

Using conservation of momentum, the missile impact force equals the change in linear (sliding) or angular (overturning) momentum of the HSM. The HSM velocities immediately after impact are:

Sliding:

V =

m v M + m i

Overturning:

A A

=

md md + I i

2

v Where:

V

=

Initial linear velocity of module after impact vi

=

103 ft/sec., Initial velocity of missile A

=

Initial rotational velocity about point A (Figure 8.2-2) m

=

124 slugs, Mass of missile M

=

10,310 slugs, Mass of loaded HSM (single freestanding HSM)

D

=

12.5 ft, height of missile center of gravity above basemat IA

=

1,225,000 slug-ft2, mass moment of inertia of loaded HSM about point A Substituting and solving for V and A produces an initial linear velocity of 1.23 ft/sec. and an angular velocity of 0.13 radians/sec.

The actual ratio between HSM sliding and rotation depends on where the missile impacts the shield wall. A low elevation impact produces mainly sliding, while a high elevation impact produces mainly rotation.

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SAR-II-8.4 CH 8 Rev. 6 8-69 of 8-98 For an impact at the bottom of the HSM wall, the kinetic energy imparted to the HSM is absorbed by sliding friction between the concrete of the HSM and the basemat. ACI 349 [8.5] recommends a coefficient of friction of 0.6. Assuming that the missile impulse load results in sliding of the HSM and equating the kinetic energy to the sliding friction gives:

(M + m) g

= 1 2 (M + m) V2

Where:

=

0.6, Coefficient of friction

=

Linear distance module slides M, m, g, and V are as defined above Substituting gives = 0.47 inches Therefore, a single free-standing module slides a maximum distance of 0.47 inches due to a low elevation tornado missile impact. At the opposite extreme, when the massive missile impacts at the top of the side wall, most of the missile energy is absorbed in rotation of the HSM. Equating the initial kinetic energy of the HSM to the increase in potential energy as the HSM center of gravity rises due to rotation gives:

Loss of Kinetic Energy =

Increase in Potential Energy 1

2 I

= Mgd [cos(

+

- 2 ) - cos ]

2 A

A

Where:

and are defined in Figure 8.2-2 M, g, d, IA, and A are as defined above Substituting and solving for shows that the HSM rotates a maximum of 0.41 degrees about the bottom edge opposite the point of impact. Therefore, the HSM provides a stable body as tip over will not occur until the center of gravity rotates past the edge (point A in Figure 8.2-2) to an angle of more than 34.6 degrees 8.2.2.3 Accident Dose Calculations. Each exposed component of the NUHOMS-12T system is specifically designed to withstand tornado generated missiles as discussed in the preceding paragraphs. The consequence of reduced shielding effects of adjacent HSMs is presented in Section 8.2.1.

8.2.2.4 Recovery. Following a tornado missile strike, all damaged HSM components will be evaluated for structural adequacy and measured dose rates, and the appropriate repair/replacement program initiated.

For a missile strike on an end shield wall, this plan would include the use of temporary shielding until the wall can be repaired or replaced. As noted above, the dose consequences to the INL workers and public are bounded by the reduced shielding analysis presented in Section 8.2.1.

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SAR-II-8.4 CH 8 Rev. 6 8-70 of 8-98 8.2.3 Earthquake 8.2.3.1 Cause of Accident. The design basis seismic acceleration described in Section 3.2.3.1 are assumed to act on the TMI-2 ISFSI system components. For this evaluation, the TMI-2 ISFSI free field site accelerations (zpas) with design response spectra of NRC Regulatory Guide 1.60 [8.27] are used for the seismic analysis of the NUHOMS-12T system components. The horizontal and vertical design response spectra are shown in Figure 8.2-3.

8.2.3.2 Accident Analysis. As discussed in Section 3.2.3.2, the peak free field horizontal ground acceleration of 0.36g and the peak vertical ground acceleration of 0.24g are utilized for the seismic analysis of the NUHOMS-12T components. Based on NRC Regulatory Guide 1.61 [8.28], a damping value of three percent is used for the DSC seismic analysis. Similarly, a damping value of seven percent for miscellaneous steel and concrete is utilized for the HSM. An evaluation of the frequency content of the loaded HSM is performed to determine the dynamic amplification factors associated with the design basis seismic response spectra for the NUHOMS-12T HSM and DSC. The dominant structural frequencies calculated are 23.8 Hz in the horizontal direction and 67.8 Hz for the DSC and DSC support steel in the vertical direction, and 36.3 Hz and 67.8 Hz for the HSM in the horizontal and vertical directions, respectively. Table 1 of NRC Regulatory Guide 1.60 requires amplification factors for structural frequencies below 33 Hz, which result in horizontal accelerations of 0.49g for the DSC and 0.36g for the HSM. The dominant vertical frequencies of the DSC and HSM exceed 33 Hz, corresponding to the zero period acceleration of 0.24g in the vertical direction.

A.

DSC Seismic Evaluation As discussed above, the maximum calculated seismic accelerations for the DSC inside the HSM are 0.49g horizontally and 0.24g vertically. A simplified rigid body static analysis using these accelerations shows that the DSC may lift off the support rails inside the HSM. The resulting stresses in the DSC shell due to vertical and horizontal seismic loads are determined and included in the appropriate load combinations. The seismic evaluation of the DSC is described in the paragraphs that follow. The DSC support structure is subjected to the calculated DSC seismic reaction loads as discussed in Paragraph C below.

(i)

DSC Natural Frequency Calculation Two natural frequencies, each associated with a distinct mode of vibration of the DSC, are evaluated. These two modes are the DSC shell cross-sectional ovalling mode, and the mode with the DSC shell bending as a beam.

(a)

DSC Shell Ovalling Mode The natural frequency of the DSC shell ovalling mode is determined from the Blevins [8.29]

correlation as follows:

f

=

i R

E 2

1 2

(

)

(Table 12-1, Case 3) [8.29]

R

=

33.31 in., DSC mean radius E

=

28.8E6 psi, Youngs Modulus

=

0.3, Poissons ratio

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SAR-II-8.4 CH 8 Rev. 6 8-71 of 8-98 i

=

0.289 t R

i i i

(

)

2 2

1 1

+

t

=

0.55 in., minimum thickness of DSC shell

=

0.284/g, steel mass density The lowest natural frequency corresponds to the case when i = 2.

Hence:

2

=

0.0128 sec.

Substituting gives:

f

=

12.7 Hertz For conservatism it is assumed that the entire DSC mass is excited by the DSC ovalling mode and the static analysis is performed using this frequency.

(b)

DSC Beam Bending Mode The DSC shell is conservatively assumed to be simply supported at the two ends of the DSC.

The beam bending mode natural frequency of the DSC was calculated using the following equation:

fi

=

i 2

2 2 L EI m

(Table 8.1, Case 5) [8.29]

E

=

28.8E6 psi, Young's Modulus I

=

65,700 in4, DSC moment of inertia L

=

163.5 in, Total length of DSC m

=

70,000/163.5g = 428/g lb/in 7

=

i ;

8 for lowest natural frequency, i = 1 Substituting yields:

f1

=

76.6 Hertz.

The DSC spectral accelerations at this frequency correspond to the zero period acceleration.

(ii)

DSC Seismic Stress Analysis With the DSC resting on the support rails inside the HSM, the stresses induced in the DSC shell are calculated due to the 0.62g horizontal and 0.24g vertical seismic accelerations applied as equivalent static loads. The DSC stresses due to the resulting vertical acceleration are calculated by factoring the dead load analysis results reported in Section 8.1. For the stress evaluation of the DSC shell due to seismic accelerations in the lateral direction, the

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SAR-II-8.4 CH 8 Rev. 6 8-72 of 8-98 resulting equivalent static acceleration is conservatively assumed to be resisted by one of the two support rails inside the HSM. The DSC shell stresses obtained from the analyses of vertical and horizontal seismic loads are combined by sum of the squares (SRSS).

As stated in Section 4.2.5.1, an axial retainer is included in the design of the DSC support system inside the HSM to prevent sliding of the DSC in the axial direction during a postulated seismic event. The stresses induced in the DSC shell and bottom cover plate due to the restraining action of this retainer for a horizontal seismic load, applied along the axis of the DSC, are evaluated and found to be negligible.

The stability of the DSC against lifting off one of the support rails during a seismic event is evaluated by performing a rigid body analysis, using the free field 0.36g horizontal and 0.24g vertical accelerations. The horizontal equivalent static acceleration of 0.49g is applied laterally to the center of gravity of the DSC. The point of rigid body rotation of the DSC is assumed to be the center of the support rail. The applied moment acting on the DSC is calculated by summing the overturning moments. The stabilizing moment, acting to oppose the applied moment, is calculated by subtracting the effects of the upward vertical seismic acceleration of 0.24g from the total weight of the DSC and summing moments at the support rail. Since the rigid body applied moment is greater than the stabilizing moment, the DSC may lift off the DSC support structure inside the HSM. However, as shown in the Appendix A drawings, the DSC is embedded within the HSM front and rear walls to ensure that the maximum possible lift off is limited to one inch. Therefore, the DSC will remain seated on the support rails in the HSM at the conclusion of a seismic event. The DSC vent system, including HEPAs, will remain functional during and after the seismic event.

B.

HSM Seismic Evaluation The seismic response of the HSM is performed using equivalent static procedure. The dominant frequency of the HSM concrete structure in the lateral direction is 36.3 Hz. The dominant frequency of the concrete structure in the vertical direction is 70.8 Hz. All these frequencies are in the rigid range. Therefore, the seismic evaluation of the HSM concrete structure is performed in the rigid range by applying the rigid range horizontal (0.36g) and vertical (0.24g) accelerations. Seismic loads in the horizontal direction are assumed to be resisted by frame and shear wall action of the HSM.

Accordingly, the HSM is modeled with three-dimensional brick elements and the horizontal rigid range accelerations are applied to the model in both horizontal directions. Similarly, the vertical rigid range acceleration is applied to account for the vertical seismic effect. The results are included in the load combinations with the appropriate strength reduction factors. The factors used for the HSM are presented in Section 3.2.5. The load combination results for normal, off-normal, and accident conditions are presented in Section 8.3.5.

An analysis is also performed to establish a conservative rigid body factor of safety against overturning and sliding for a single, free-standing module. This analysis consists of comparing the stabilizing moment produced by the weight of the HSM and DSC, reduced by 24% to account for the upward vertical seismic acceleration, against the overturning moment produced by applying the 0.36g load at the centroid of the HSM and 0.49g at the centroid of the DSC. For sliding of the HSM, the horizontal force of 0.36g acceleration for the HSM and 0.49g for the DSC are compared against the frictional resisting force of the foundation slab. In this manner, the factor of safety against sliding is established. The concrete coefficient of friction is taken as 0.6 as defined in Section 11.7.4.3 of ACI 318-95 [8.30].

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SAR-II-8.4 CH 8 Rev. 6 8-73 of 8-98 The details of the seismic evaluation of the HSM are described in the paragraphs that follow.

(i)

HSM Frequency Analysis To determine the loaded HSM frequency content, the ANSYS [8.31] finite element model shown in Figure 8.1-13 is utilized. The lowest significant horizontal structural frequencies calculated are 23.8 Hz for the DSC with the support structure, and 36.3 Hz for the HSM. The lowest significant vertical mode is 67.9 Hz for both DSC with the support structure HSM.

The corresponding spectral accelerations are 0.36g for the HSM, 0.49 for the DSC and DSC support structure in the horizontal direction, and 0.24g for the HSM and DSC in the vertical direction.

(ii)

HSM Seismic Response Spectra Analysis The rigid range accelerations in horizontal and vertical directions are applied to the finite element model in three orthogonal directions and static analysis is performed. The model of the DSC and the DSC support structure is also included in the three-dimensional model of the HSM. The resulting forces and moments in the HSM walls, roof and floor of a single HSM are calculated using the linear finite element model shown in Figure 8.1-13 and the computer program ANSYS [8.31]. The three directional rigid mode responses are combined by the square root of the sum of the squares (SRSS) method. The combined maximum moments and shear forces are reported in Table 8.2-2.

(iii)

HSM Overturning Due to Seismic The following conservative analysis is performed to show that a single, free-standing HSM will not overturn due to seismic loads. The HSM stabilizing moment (Mst) is:

Mst

=

(Whsm + Wdsc)d = 1703K-ft Where:

Whsm

=

Weight of HSM = 272 kips Wdsc

=

Maximum weight of DSC = 60 kips d

=

5.13 ft, Horizontal distance from center line to corner d1

=

Height of HSM center of gravity = 7.4 ft d2

=

Height of DSC center line = 8.5 ft

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SAR-II-8.4 CH 8 Rev. 6 8-74 of 8-98 The seismic overturning moment is:

Mot

=

(Whsmav1 + Wdscav2)d + Whsmd1ah1 +Wdscd2ah2 = 1,423 K-ft Where:

Mot

=

Overturning moment av1

=

0.24g, HSM vertical seismic acceleration av2

=

0.24g, DSC vertical seismic acceleration ah1

=

0.36g, HSM horizontal seismic acceleration ah2

=

0.49g, DSC horizontal seismic acceleration The result of this analysis indicates that a single loaded free-standing HSM will not overturn during a seismic event. The margin of safety against overturning is 1.2.

(iv)

HSM Sliding Due to Seismic To show that a single free-standing HSM will not slide due to the postulated horizontal and vertical seismic accelerations, the following conservative analysis is performed. The friction force resisting sliding (Fs1) is:

Fs1

=

[Whsm(1-av1) + Wdsc(1-av2)]

= 151 kips

=

Coefficient of friction between the HSM concrete walls and the floor slab foundation = 0.6 The applied horizontal seismic force (Fhs) is:

Fhs

=

Whsmah1 +Wdscah2 = 127 kips The force required to slide the HSM is larger than the resulting lateral seismic force and therefore, the HSM will not slide. The factor of safety against sliding is 1.19.

C.

DSC Support Structure Seismic Evaluation (i)

DSC Support Structure Natural Frequency The lowest structural frequency of the DSC and the DSC support structure inside the HSM is dominated by the mass of the DSC. The DSC and support structure are included in the HSM analytical model. The lowest significant horizontal and vertical frequencies of the DSC/DSC support structure are 23.8 Hz and 67.9 Hz, respectively.

(ii)

DSC Support Structure, Seismic Response Spectra Analysis The DSC support structure has two flexible modes; 15.22 Hz and 24.13 Hz in two orthogonal horizontal directions. The flexible mode response of the support structure is computed for these two modes and combined with the rigid mode response. For the support frame cross members, the maximum axial stress is 2 ksi and the maximum shear stress is

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SAR-II-8.4 CH 8 Rev. 6 8-75 of 8-98 0.03 ksi. Similarly, the maximum axial stress in the support rail is 2.9 ksi and the maximum shear strerss is 5.7 ksi. These compare with Code allowables of 13.1 ksi for shear stress and 19.7 ksi axial stress and, as a result, have a considerable design margin.

The effect of concentrated anchor bolt forces is included in the design of the DSC support structure connection details. Similarly, each connection of the support rails is designed for the resulting seismic loads. This condition envelopes all other loading conditions for the individual bolts or structural elements of the DSC support structure.

The stresses in the DSC support structure members due to seismic accelerations are included in the load combination results reported in Section 8.3.7.

(iii)

DSC Seismic Retainer Analysis The DSC seismic retainer detail, located inside the HSM access opening, is shown on the Appendix A drawings. The retainer bears on the end of the DSC and transfers axial seismic loads to the support structure.

The clearance between the DSC seismic retainer and the DSC is designed for the maximum DSC thermal expansion. During normal storage, there is a small (1/4 to 1/2 inch) gap that allows movement of the DSC relative to the HSM. This motion produces a small increase in the DSC axial force due to seismic loads and has been included in the design of the DSC.

The HSM ANSYS finite element model includes the seismic retainer and the DSC rear stop.

The resulting axial forces in the retainer calculated in seismic analysis are combined and applied to the seismic retainer. The total resulting force excluding the force to overcome friction is 49.1 kips.

This load is transferred to the HSM front and rear walls and is included in the design basis forces and moments. The maximum bending and shear stresses in the seismic retainer are 12.3 ksi and 16.1 ksi which are less than the allowable shear and bending stresses of 18.4 ksi and 39.4 ksi respectively.

8.2.3.3 Accident Dose Calculations. The NUHOMS-12T system components are conservatively designed and analyzed to withstand the forces generated by a postulated design basis earthquake accident.

Hence, there are no dose consequences resulting from an earthquake.

8.2.4 Flood 8.2.4.1 Cause of Accident. As described in Section 3.2.2, the ISFSI base slab is located at, or above, the maximum predicted flood plain elevation. The maximum flood water elevation is well below the bottom of the DSC, which is some 5-9 above the slab. Therefore, there are no credible flood loads that act on the TMI-2 ISFSI components.

8.2.5 Accidental Cask Drop This section addresses the structural integrity of the DSC when subjected to postulated cask drop accident conditions. As described in Chapter 3, the DSC basket assembly is not required to maintain subcriticality of the TMI-2 canisters for this application. Therefore, no evaluation of the basket is performed.

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SAR-II-8.4 CH 8 Rev. 6 8-76 of 8-98 The structural integrity of the MP187 cask and DSC confinement boundary for all postulated on-site transfer accident drops have been evaluated for bounding loads and the results are described in the SMUD SAR

[8.12]. The transportation hypothetical accident conditions are described in the 10 CFR Part 71 SAR [8.18].

This section demonstrates the adequacy of the DSC confinement boundary for postulated 10 CFR Part 72 transfer operation drop accident events.

8.2.5.1 Cause of Accident A.

Cask Handling and Transfer Operation As described in Section 5.0, all handling operations involving hoisting and movement of the cask and DSC are performed inside the TAN facility. These include utilizing the crane for placement of the DSC into the cavity of the MP187 cask, and placement of the MP187 cask/DSC onto/off of the turning skid/transport skid/trailer.

Once the MP187 cask is loaded onto the transport skid/trailer, and the impact limiters are installed and secured, it is pulled to the TMI-2 ISFSI site by a tractor vehicle using 10 CFR Part 71 transportation regulations. A predetermined route is chosen to minimize the potential hazards that could occur during transport. This movement is performed at low speeds. Safeguards provided in system operating procedures ensure that the system design margins are not compromised. As a result, it is highly unlikely that any plausible incidents leading to a cask drop accident could occur.

Similarly, at the ISFSI site, after removing the impact limiters, the transport skid/trailer is aligned with the HSM using hydraulic positioning equipment. The cask is docked with, and secured to, the HSM access opening. The loaded DSC is transferred to/from the HSM using a hydraulic ram system.

The hold down mechanisms that secure the cask to the transport skid/trailer remain in place at all times during the DSC transport. As a result, there is no mechanistic way during these operations for a cask drop accident to occur.

B.

Cask Drop Accident Scenarios In spite of the incredible nature of any scenario that could lead to a cask drop accident, a conservative range of drop scenarios during the 10 CFR Part 72 transfer operations are developed and evaluated. These bounding scenarios assure that the integrity of the DSC and TMI-2 canisters are not compromised. Comparison of these scenarios to the 10 CFR Part 71 postulated drop scenarios demonstrate that the MP187 cask maintains structural integrity and the DSC confinement boundary will remain intact. Therefore, there is no potential for a release of radioactive materials to the environment due to a cask drop. The range of 10 CFR Part 72 transfer drop scenarios conservatively selected for design are illustrated in Figure 8.2-1 and include the following cases:

A horizontal side drop from a height of 80 inches after the impact limiters have been removed.

An oblique corner drop from a height of 80 inches at an angle of 30° to the horizontal, onto the top or bottom corner of the cask after the impact limiters have been removed (two cases).

The height of 80 inches is chosen as this envelopes the maximum vertical height of the cask when secured to the transport skid/trailer assembly. The angle of inclination for the corner drop of 30° represents the maximum possible impact angle that the cask can rotate downward with one end supported horizontally on the transport skid trailer at a height of 80 inches.

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SAR-II-8.4 CH 8 Rev. 6 8-77 of 8-98 C.

Cask Drop Accident Load Definitions Transportation of the MP187 cask and contents from the TAN facility to the TMI-2 ISFSI site will be performed under the rules of 10 CFR Part 71 with the cask impact limiters installed as described in the MP187 cask SAR [8.18]. The impact limiters will be removed once the transportation trailer reaches the TMI-2 ISFSI site. From the time the impact limiters are removed until the DSC is transferred into the HSM, the trailer will be moved a distance of less than 500 feet over compacted gravel, asphalt, and 12-to 30-inch-thick concrete surfaces. There will be no vertical lifts of the cask and the only large horizontal motions are controlled by movement of the tractor/trailer at speeds of 5 mph or less. Once the cask and DSC are located at the chosen HSM, final alignment will be achieved by use of the hydraulically controlled skid positioning system. Neither of these movements create a potential for a cask drop accident.

As described in the SMUD Part 72 SAR [8.12], it is assumed that the cask can slide off the end (corner impact), or side (side impact) by an undefined nonmechanistic failure. These drops would result in a maximum drop height of 80 inches onto an under-reinforced concrete slab. As there are no vertical cask movements considered at the TMI-2 ISFSI there are no potential end impact scenarios.

EPRI Report NP-4830, The Effects of Target Hardness on the Structural Design of Concrete Storage Pads for Spent Fuel Casks, [8.33] provides expected decelerations for postulated cask side and end drops for typical ISFSIs licensed to 10 CFR Part 72 requirements. The report establishes the maximum expected decelerations for a range of surface conditions and drop heights up to 80 inches.

For the MP187 cask weight and dimensions, the expected maximum decelerations for drops onto a 30-in.-thick under-reinforced concrete slab are less than 75g for an end drop or a side drop. Corner drops are not explicitly covered in the EPRI report. However, the maximum decelerations for a corner drop are determined to be significantly lower than those for side drops [8.33].

8.2.5.2 Accident Analyses. The 10 CFR Part 72 and 10 CFR Part 71 hypothetical accident condition drop analyses presented in the SMUD Part 72 SAR [8.12] and the MP187 cask SAR [8.18], respectively, provide assurance that the MP187 cask can withstand all postulated impacts without affecting the DSC confinement boundary integrity.

A.

NUHOMS-12T DSC Drop Analyses The DSC confinement boundary stresses for the postulated Part 72 drops are similar to those reported in the generic NUHOMS SAR [8.9]. The finite element model shown in Figure 8.1-7 was used to develop the DSC confinement boundary stress intensities reported in Table 8.2-3 for the design basis 75g side (horizontal) and end (vertical) drops.

The basic design of the TMI-2 DSC basket is similar to the NUHOMS generic DSC basket which has been demonstrated to withstand the design basis 10 CFR Part 72 drop loads without significant deformations [8.9]. The TMI-2 ISFSI basket assembly is not required to maintain geometrical control to ensure subcriticality of the TMI-2 canisters following a drop accident and, therefore, is not analyzed for the hypothetical drop accident decelerations. It is expected that the TMI-2 DSC basket will behave in a similar manner to that described for the NUHOMS generic basket in the Reference 8.9 postulated drop decelerations. However, in the worst case, the TMI-2 basket structure may yield and the TMI-2 canisters could bear on the DSC shell. The TMI-2 canisters were drop tested to 190g during their design program, as described in Reference 8.34, and can support the weight of all TMI-2 canister assemblies without affecting any of the criticality assumptions described in Section 3.3. The

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SAR-II-8.4 CH 8 Rev. 6 8-78 of 8-98 original drop tests were performed by placing the TMI-2 canisters into pipe sections to simulate the 125-B inner containment vessel. The TMI-2 DSC spacer disc simulates the tube design of the 125-B and will have little effect on the maximum TMI-2 canister stresses. The TMI-2 canister maximum bending stress in the 14-inch diameter, 1/4-in. wall tube for the 75g horizontal drop is less than 2 ksi.

This is a very small fraction of the material yield of 30 ksi and the ASME Code Service Level D allowable of 66 ksi at 300°F.

B.

MP187 Transfer Cask Drop Analyses Results for the 10 CFR Part 72 drop analyses for the MP187 cask are reported in Reference 8.12.

These analyses envelope those resulting from the drop accident scenarios described above, and clearly demonstrate the acceptability of the MP187 cask and DSC confinement boundary to meet ASME Code requirements.

8.2.5.3 Loss of Neutron Shield. This accident conservatively postulates loss of neutron shield on the MP187 cask.

8.2.5.3.1 Cause of AccidentThe reference SMUD Part 72 SAR [8.12] design basis cask drop analyses show that all components of the MP187 maintain their structural integrity for the postulated drop accident events. Complete loss of the neutron shield material is not a credible event. For this conservative analysis, it is assumed that the cask neutron shield is breached as a result of a postulated drop accident, and the shielding effects are lost. The effect on the cask, DSC, and TMI-2 canister temperatures is bounded by the results of the 103°F ambient case described in Section 8.1.3.3.

8.2.5.3.2 Accident Dose for Loss of Neutron ShieldFor an MP187 cask containing a design basis NUHOMS-12T DSC, complete loss of the neutron shield would increase the peak cask surface contact dose from 3.2 mrem/hour to 4.9 mrem/hour. The only potential off-site dose consequences would be additional direct and air scattered radiation. This accident dose rate is well below typical fuel storage and transportation cask dose rates for normal operations. Because the accident dose rate is so low, no significant exposure will be received by workers or members of the public. Following a loss-of-neutron shield event, the canister should be removed from the cask as discussed in Section 8.2.5.4. Conservatively assuming two workers are exposed to the peak cask surface dose rate for eight hours, an additional exposure of 27 person-mrem may be received. No release of radionuclides will occur due to this event.

This does not preclude handling operations for recovery of the cask and its contents. Water bags, or other neutron absorbing material could be wrapped around the cask to reduce the surface dose to an acceptable limit for recovery operations, thus minimizing exposure of personnel in the vicinity. The actual local dose rate, recovery time, operations needed to retrieve the cask, and the required actions to be performed following the event, depend upon the severity of the event and the resultant cask and trailer/skid damage.

The INL site boundary is approximately eight miles from the ISFSI and there would be no measurable increase above background levels regardless of how long it took to recover from the accidental loss of the neutron shield.

8.2.5.4 Recovery. For drop heights of less than 15 inches, the cask will be loaded back onto the transfer skid/trailer and moved to the HSM. The DSC will then be transferred to the HSM in the normal manner described previously. For drop heights greater than 15 inches, the cask and contents will be returned to the TAN Hot Shop, or other appropriate facility. There, the DSC will be inspected for damage, the DSC

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SAR-II-8.4 CH 8 Rev. 6 8-79 of 8-98 opened, and the TMI-2 canisters removed for inspection, as necessary. Removal of the DSC cover plate and shield plug assembly are described in Section 5.0.

Following recovery of the cask and unloading of the DSC, the cask will be inspected, repaired, and tested, as appropriate, prior to reuse.

8.2.6 Lightning 8.2.6.1 Postulated Cause of Event. As described in Reference 8.35, a lightning risk assessment has been conducted for the ISFSI in accordance with the Lightning Protection Code, NFPA 780. This risk assessment calculated a moderate risk factor for the TMI-2 ISFSI site. Although the effects of a lightning strike are not significant, a lightning protection system is provided to further reduce the risk.

8.2.6.2 Analysis of Effects and Consequences. Should lightning strike in the vicinity of the HSM, the normal storage operations of the HSM will not be affected. The current discharged by the lightning will follow the low impedance path offered by an early streamer emission system attached to the light support poles surrounding the ISFSI. As an added safety factor, the HSMs would not be damaged by the heat or mechanical forces generated by any current passing through the higher impedance concrete. Since the HSMs require no equipment for their continued operation, any resulting current surge from a lightning strike would not affect the normal operation of the HSM.

Since no off-normal condition will develop as a result of a lightning strike in the vicinity of the HSM, no corrective action would be necessary. Also, there are no radiological consequences of a lightning strike at the ISFSI.

8.2.7 DSC Leakage The DSC shell is designed as a confinement boundary to prevent leakage of contaminated materials.

The analyses of normal, off-normal, and accident conditions have shown that no credible conditions can breach the DSC shell or fail the double seal welds at each end of the DSC. For the TMI-2 ISFSI, the DSC vent and purge ports are each connected to HEPA filters to allow the release of hydrogen gas created by potential radiolysis within the TMI-2 canisters. The vents in the DSC and the TMI-2 Debris Canisters ensure that this equipment is maintained in an unpressurized state under all conditions.

As discussed below, this event postulates a direct release to the environment of 48.3 Ci (over a one-month period) of fission product gases and particulates contained in the TMI-2 canisters. This nonmechanistic accident conservatively assumes that both HEPA filter trains are instantaneously ruptured and no longer provide any filtration. All other components of the NUHOMS-12T system remain intact.

8.2.7.1 Cause of Accident. There is no credible event that could result in the rupture of the DSC HEPA filter trains. The passive nature of the NUHOMS-12T system and the various design features ensure that the integrity of the DSC shell confinement boundary and HEPA filter trains are maintained.

Nevertheless, this evaluation assumes that the HEPA filter trains are ruptured due to an event of unspecified origin.

8.2.7.2 Accident Analysis. There are no structural or thermal consequences resulting from the DSC leakage accident described above. The radiological consequences of this accident are described in Section 8.2.7.3.

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SAR-II-8.4 CH 8 Rev. 6 8-80 of 8-98 8.2.7.3 Accident Dose Calculations. The postulated accident assumes that the HEPA filter trains from one DSC have failed such that 48.3 Ci of the fission product gases and particulates in the TMI-2 canisters are released to the atmosphere over a one-month period. The nuclide composition of the postulated release is provided in Chapter 7, Table 7.2-3. The total effective dose equivalent received by an individual at or beyond the INL controlled area boundary under the worst meteorological conditions was calculated using the RSAC code. Contributions to this exposure include the 50-year committed effective dose equivalents from the inhalation and ingestion pathways and the effective dose equivalents from the ground surface and cloud gamma pathways. Table 7.6-3, Accident condition provides the thyroid organ doses and the whole-body doses for the accident condition of HEPA filter failure. The calculations for this accident condition conservatively assume that all 62 TMI-2 Filter Canisters are contained in the DSC experiencing the accident condition. The resultant accident dose of 0.6 mrem is well within the 10 CFR 72.106 limit, which restricts the maximum whole body or organ dose beyond the owner-controlled area from any design basis accident to be less than 5 rem.. A worker spending 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week for one month at the INTEC fence would receive a total effective dose equivalent of 151 mrem from the effluent gases and particulates, as well doses received from external radiation, (Table 7.4-2) as a result of this event.

8.2.7.4 Recovery. Recovery from this accident condition may be achieved by returning the challenged DSC to the TAN Hot Shop, or other appropriate facility, offloading the TMI-2 canister, repairing the canisters and/or DSC as appropriate, reloading and returning it to the ISFSI. Alternatively, a dry transfer cask-to-cask system may be used to replace the challenged DSC. An HSM with an overpack is provided to ensure that the challenged DSC can be safely stored at the ISFSI without the need to return it to the TAN Hot Shop, or other appropriate facility.

As shown on the Appendix A drawings, the overpack is similar to the DSC confinement boundary.

The 5/8 steel shell has an inner diameter of 68.25 inches and a length sufficiently long to accommodate the DSC, HEPA filter blocks, and grapple ring. The inner end of the overpack is sealed with a shop-installed 1-1/2-in.-thick plate. The method of construction, codes, and inspections will be the same as those invoked for the DSC confinement boundary to provide assurance of a leak-tight system. The steel overpack is installed onto the HSM rails and is welded to the front HSM access collar to resist longitudinal load and maintain the circularity of the front end over a long period of time.

After insertion of a challenged DSC, the front (outer) end of the overpack is sealed by welding a 1-1/2-in.-thick closure plate across the shell opening. The rear plate of the opening contains HEPA filters similar to those described for the DSCs installed in the other HSMs. Those filters have the same design, operating, and maintenance criteria as the HEPA filter trains described in Chapter 6.

The changes required for the HSM to accommodate the overpack are: a slightly larger, steel lined, front access sleeve; support rails lowered by approximately 1 inch to maintain the same cask center line height during transfer; and an increase in the diameter of the rear wall recess to clear the overpack. These modifications do not alter the calculated behavior or reinforcement requirements described in Section 8.1.

The HSM front access door, seismic restraint, and rear HEPA filter access door are similar to those described for the NUHOMS-12T HSM and do not require re-analysis for this scenario.

The DSC overpack design loads for pressure, dead load, and seismic forces are of the same magnitude as those of a DSC loaded into an HSM. This assures that the reported maximum stresses for the DSC envelope the calculated stresses for the overpack.

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SAR-II-8.4 CH 8 Rev. 6 8-81 of 8-98 8.2.8 Accident Pressurization of DSC The analysis documented in Appendix C shows that the maximum hydrogen concentration under worst case conditions will be 4.8% in the TMI-2 canister, and less than 1.2% in the DSC. These concentrations are less than the lower flammability limit of 5%. With a vented system there will be no pressure build-up within the DSC.

8.2.8.1 Cause of Accident. The TMI-2 canisters and DSC are configured to prevent hydrogen concentrations from reaching the flammability limit. However, hydrogen deflagration within one of the TMI-2 canisters is postulated as an accident scenario. For this event to occur, not only must the hydrogen concentration reach the 5% minimum required for flammable levels, but an ignition source must also occur within the TMI-2 canister to initiate combustion.

8.2.8.2 Accident Analysis. In the unlikely event that combustion occurs inside a TMI-2 canister, the hydrogen-air combustion can take the form of either a deflagration or a detonation. Hydrogen deflagrations are subsonic combustion processes that produce peak to initial pressure ratios of approximately 8:1.

Hydrogen detonations are supersonic combustion processes that can produce shock waves with momentary face-on pressure ratios of up to 40:1 [8.36]. Detonations require hydrogen concentrations in excess of 18% in air, and certain types of geometry which allow the flame to accelerate up to its detonation velocity. As the maximum TMI-2 canister hydrogen concentration is less than 5%, the likelihood of both detonation conditions being met in the TMI-2 storage canisters is extremely remote. Therefore, only the deflagration process is considered. Since the hydrogen concentration buildup will always be higher in the TMI-2 canisters than in the DSCs, only the deflagration in one TMI-2 canister is considered.

The maximum calculated pressure in the TMI-2 canister and DSC during normal, off-normal, or accident conditions from Section 8.1.1.1.B is 11.4 psig (26.1 psia). Therefore, using a pressure ratio of 8:1, the maximum overpressure due to hydrogen deflagration in the canister will be approximately 209 psia (194 psig).

The specified design pressure of the TMI-2 canisters is 150 psig. The original Babcock and Wilcox ASME Stress Analysis [8.37] was reviewed to determine whether an internal pressure of 194 psig would overstress the TMI-2 canisters. The report indicates that the canisters were sufficiently over-designed to accommodate pressures in excess of 194 psig and still meet ASME Section VIII allowable stress values.

Therefore, no degradation of the TMI-2 canister will occur in the event of a hydrogen deflagration.

Because of the very small ports in the TMI-2 canisters and the momentary nature of the over pressure condition, the amount of gas which would escape from a TMI-2 canister into the DSC cavity during a deflagration is too small to affect the DSC cavity pressure. No degradation of the DSC confinement boundary will occur in the event of a hydrogen deflagration inside a TMI-2 canister.

8.2.8.3 Accident Dose Calculations. There are no dose consequences as a result of accidental pressurization of the DSC. Any release that may occur is bounded by the evaluation in Section 8.2.7.

8.2.9 Fire and Explosion 8.2.9.1 Cause of Accident. A fire hazards analysis has been performed for the TMI-2 ISFSI. Because it is constructed of non-combustible materials, the ISFSI does not impose a fire or explosion hazard to the surrounding area or facilities. The buildings, storage yards, fuel storage tanks, and access roads nearest the ISFSI have been evaluated for potential impacts to the ISFSI due to fire and explosions. Because of the

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SAR-II-8.4 CH 8 Rev. 6 8-82 of 8-98 limited fuel quantities present and the substantial distances to the ISFSI, it has been determined that they pose no threat to the ISFSI structures or components. The most severe source of fire comes from burning fuel in the support vehicles used for loading and unloading the facility, and sampling and testing of the DSCs in storage. The consequences of a fire from these sources are evaluated below.

8.2.9.2 Accident Analysis. Any fire from ISFSI service vehicles would be very limited due to the small amount of combustible material available. At any given time, the total amount of fuel in service vehicles and other containers within the ISFSI boundaries will be limited to less than 300 gallons. This insures that any fire will be relatively small and slow burning since the fuel would be discharging at a low rate, or that the fire would be over in a relatively short time. Based on burn test data for gasoline, if all 300 gallons of fuel was pooled and available for burning, the fire would last less than 30 minutes. Since there is no place for the fuel to pool on the pad to support such a fire, the actual burn time of any ISFSI fire would be considerably less than 30 minutes. An infinitely large fire that burned for two hours around an HSM was evaluated in Reference 8.38. This evaluation showed that only the outer three inches of the HSM concrete would reach the ACI 349 accident temperature limit of 650°F. Some minor concrete cracking may occur due to the differential heating of the HSM but will be controlled by the steel reinforcement. The inside temperature of the HSMs will not vary significantly from normal operating limits due to the massive nature of the HSM concrete thickness. Cracks, or surface spalling that may occur as a result of the worst case postulated fire would have no significant effect on the structural strength or the radiation protection provided by the HSM.

8.2.9.3 Accident Dose. Although local radiation levels may increase slightly, there will be no effect to the public. Any dose increase due to fire damage is bounded by the loss of shielding discussed in Section 8.2.1. There will be no increases in airborne releases due to this event since the DSC and vent are protected inside the HSM and are well above potential fuel sources for the fire.

8.2.9.4 Recovery. Recovery from the fire would involve a radiological and physical survey of the HSM and surrounding area to determine the extent of potential damage. As stated above, it is expected that there will be no significant structural damage to the HSMs or increases in radiation exposures due to this accident. Any damage to the HSMs can be repaired or replaced. Replacement would require that the loaded DSC be removed and a new HSM installed. Alternatively, surface spalling can be repaired, in place, and the HSM returned to service without the need to remove the loaded DSC.

8.2.10 Blockage of Space Between Adjacent HSMs This accident conservatively postulates the complete blockage of the gap between the adjacent HSMs.

The debris is conservatively assumed to remain in place for an indefinite period of time.

8.2.10.1 Cause of Accident. Since the HSMs are located outdoors, there is a small probability that the gap between HSMs could become blocked by debris from such events as wind storms, floods, and tornadoes.

Even though complete blockage of the entire gap is unlikely, such an accident is postulated to occur for this conservative analysis.

8.2.10.2 Accident Analysis. The structural consequences due to the weight of the debris blocking the gap between the modules are negligible and are bounded by the HSM loads induced for a postulated tornado (Section 8.2.2) or earthquake (Section 8.2.3).

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SAR-II-8.4 CH 8 Rev. 6 8-83 of 8-98 The thermal effects of this accident result from the increased temperatures of the DSC and the HSM due to blockage of the gap between HSMs. Even though any blockage will be cleared by site personnel, it is conservatively assumed for this analysis that the blockage is there for an indefinite amount of time.

The thermal analysis of the HSM for the blocked gap condition is performed using the same model as the normal conditions described in Section 8.1.3. The only change is the use of an insulated boundary condition in the gap region simulating blocked gap conditions. A steady state analysis is carried out to determine the worst temperature distribution in the DSC and the HSM.

The results of the blocked gap case are included in Tables 8.1-8, through 8.1-10. The results show that the maximum concrete temperature is 179°F and the maximum core debris temperature is below 219°F.

Therefore, the concrete, fuel, and all the material temperatures are well within their limits for accident conditions.

These temperatures are below the levels at which safety-impairing damage would occur to the HSM or DSC. The maximum DSC internal pressure during this event is less than 15 psig as shown in Section 8.1.1.1.B.

8.2.10.3 Accident Dose Calculations. There are no off-site dose consequences as a result of this accident. The only significant dose increase is related to the recovery operation. It is conservatively estimated that the on-site workers will receive an additional dose of less than one man-rem during the estimated eight-hour period that may be required for debris removal from the gap between the HSMs.

8.2.10.4 Recovery. Debris removal is all that is required to recover from a postulated blockage of the gap between the HSMs. Cooling will begin immediately following removal of the debris. The amount and nature of debris can vary, but even in the most extreme case, manual means or readily available equipment can be used to remove debris.

8.2.11 Basaltic Lava Flow The analysis of normal, off-normal, and accident conditions have shown that a future basaltic lava flow represents an extremely unlikely (5 x 10-6) but credible event that, without mitigation could have a potential to adversely affect facility performance.

8.2.11.1 Cause of Accident. The general topography and vent locations within the INL area suggest that any future lava will most likely erupt from vents along the axial volcanic zone or at the intersections of that zone with the volcanic-rift zones. The volcanic event could cause lava to flow from the south toward the central INL and INTEC, which includes the ISFSI.

8.2.11.2 Mitigation Actions and Design Features 8.2.11.2.1 Warning TimeTotal warning time from time of identification of magma-induced seismicity to arrival of lava in the vicinity of INTEC will range from 1 week to greater that 1 month. An alert will be declared when the INL Seismic Monitoring Program reports to the Warning Communications Center that seismic activity recorded by seismographs indicates that a volcanic eruption is possible in the near future. Equipment and workforce available at INL could construct a barrier to protect endangered facilities in as little time as one week.

8.2.11.2.2 Arrival TimeThe distance from the volcanic event will be at least 12 km (55th percentile length) and more likely in the range of 16 km (70th percentile length), and the effusion rate is likely to be waning by the time it reaches the ISFSI area. Analogy to flow velocities in other areas of the

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SAR-II-8.4 CH 8 Rev. 6 8-84 of 8-98 world with similar terrain indicates that velocities of about 2 km/day are most likely, and thus it would take several days for lava from most of the critical volcanic source area to reach the site.

8.2.11.2.3 Mitigation Actions and Design FeaturesEither of the two following mitigation actions would protect facilities from the advance of a lava flow. The objective of these actions is not to stop the advancing lava but to divert or direct the flow of lava around INTEC, and on down the Big Lost River Valley.

8.2.11.2.4 Earthen BarrierBig Lost River alluvium in the vicinity of INTEC consists almost entirely of sandy and silty limestone gravel (pebbles and cobbles). This material could be used to construct a barrier to divert or direct the lava flow around the ISFSI. The high-density limestone gravel would not be subject to breaching, as has happened in cases where low-density volcanic ash barriers were used. To protect the ISFSI from an approaching lava flow, an earthen barrier could be constructed within the INTEC and immediately surrounding the ISFSI. The earthen barrier or berm (as shown in Figure 8.2-4) would be approximately 20 feet in height and 50 feet in width at the base. Such a structure would be adequate to divert the design basis lava flow reaching the ISFSI sight and could erected within a week using excavation equipment that is available at the INL and surrounding communities.

8.2.11.2.5 Water SprayDiversion of lava flows by cooling with water sprays has been used successfully in Iceland, Japan, and Italy [8.40, 8.41, 8.42]. The combined pumping capacity of wells at INTEC, TRA, CFA, and RWMC (south-central INL facilities which have large pumping capacity) is over 14 million gallons per day. This capacity surpasses the pumping rate used on Heimaey, Iceland for a lava flow with a much higher effusion rate and much longer eruption duration than would occur at INL. The large capacity of the Snake River Plain aquifer is likely to be able to sustain those pumping rates for several days to several weeks because the water would be pumped from several high-capacity wells. The objective of this mitigation action is to cool and solidify a portion of the lava flow front and flank as it approaches INTEC, so that lava is forced to flow around the facilities.

The following factors provide a high degree of confidence that a lava flow could be diverted around INTEC:

1.

The Big Lost River Valley is broad and flat, with a gentle gradient to the north, so lava will not have a strong tendency to flow in any particular path down the valley.

2.

The distance from the volcanic vent will be at least 12 km (55th percentile length) and more likely in the range of 16 km (70th percentile length), and the effusion rate is likely to be waning by the time it reaches the ISFSI area.

3.

Although the INTEC is within the Big Lost River valley, there are no constrictions in the valley at or near the ISFSI that would cause ponding of lava.

4.

Local alluvium available for barrier construction is dense and easily moved with available equipment.

5.

Equipment and workforce available at INL could construct a barrier in as little time as one week.

These mitigation actions provide assurance that a basaltic lava flow will not adversely affect ISFSI facility performance because:

1.

A seismic network is maintained such that characteristic volcanic earthquakes can be readily distinguished from tectonic earthquakes.

2.

Criteria has been developed to determine what level of activity (e.g., earthquake clustering or rate of decrease in epicenter depth) constitutes a potentially significant volcanic event.

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SAR-II-8.4 CH 8 Rev. 6 8-85 of 8-98

3.

An alert will be declared when the INL Seismic Monitoring Program reports to the Warning Communications Center that seismic activity recorded by seismographs indicates that a volcanic eruption is possible in the near future.

4.

A site area emergency will be declared when the INL Seismic Network reports shallow seismicity accompanied by observed ground fissures.

5.

Mitigation plans consider that lava would not reach the site until a week after vent formation and thus action could be implemented before a new vent actually forms.

6.

Lava diversion takes into account site topographic and lava flow characteristics.

7.

Property removal, as a mitigation strategy is not required.

8.2.11.3 Accident Dose Calculations. In the unlikely event of a future basaltic lava flow, the ISFSI would experience no structural, thermal, or radiological consequences due to the implementation of the above mitigation actions.

8.2.11.4 Recovery. Adequate time is available for implementation of mitigation actions to preclude any impact on the ISFSI performance, therefore, recovery from this unlikely event is not required.

Table 8.2-1. Postulated Accident Loading Identification.

ACCIDENT LOAD TYPE SECTION REFERENCE NUHOMS-12T COMPONENT AFFECTED DSC SHELL ASSEMBLY DSC INTERNAL BASKET DSC SUPPORT STRUCTURE HSM ON-SITE TRANSFER CASK Tornado Wind 8.2.2 X

X Tornado Missiles 8.2.2 X

X Earthquake 8.2.3 X

X X

X X

Accident Cask Drop 8.2.5 X

X X

Loss of Cask Neutron Shield 8.2.5 X

Lightning 8.2.6 X

DSC Leakage 8.2.7 (Radiological Consequence Only)

DSC Accident Internal Pressure 8.2.8 X

Fire and Explosion 8.2.9 X

Load Combinations 8.3 X

X X

X X

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SAR-II-8.4 CH 8 Rev. 6 8-86 of 8-98 Table 8.2-2. Maximum HSM Reinforced Concrete Bending Moments and Shear Force for Accident Loads.

STRUCTURAL SECTION FORCE COMPONENT HSM INTERNAL FORCES (kip/ft, in-k/ft)(1)

TORNADO WIND AND TORNANDO MISSILE LOAD SEISMIC Floor Slab Shear 1.2 0.86 Moment 127.3 478.1 Side Wall Shear 7.4 0.78 Moment 212.4 61.2 Front Wall Shear 25.6 10.5 Moment 567.7 84.3 Rear Wall Shear 17.5 10.8 Moment 95.3 66.7 Roof Slab Shear 9.8 1.4 Moment 329.1 11.2 (1)

Maximum loads shown are irrespective of location.

Table 8.2-3. Maximum NUHOMS-12T DSC Stresses for Drop Accident Loads.

DSC COMPONENTS STRESS TYPE CALCULATED STRESS (ksi)(1, 2)

VERTICAL HORIZONTAL DSC Shell Primary Membrane 10.1 14.5 Membrane + Bending 17.7 40.4 Top Shield Plug Primary Membrane 1.1 14.9 Membrane + Bending 38.9 26.6 Top Cover Plate Primary Membrane 2.4 8.5 Membrane + Bending 3.8 14.1 Inner Bottom Cover Plate Primary Membrane 4.5 9.6 Membrane + Bending 5.2 12.7 Outer Bottom Cover Plate Primary Membrane 4.7 7.9 Membrane + Bending 7.4 12.8 Top Cover Plate Weld Primary 2.6 11.4 Bottom Cover Plate Weld Primary 2.3 10.6 (1) Values shown are maximums irrespective of location.

(2) DSC was also included in corner drop analysis for cask, however, stresses are enveloping.

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SAR-II-8.4 CH 8 Rev. 6 8-87 of 8-98 Figure 8.2-1. MP187 Cask Postulated Drop Accident Scenarios.

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SAR-II-8.4 CH 8 Rev. 6 8-88 of 8-98 Figure 8.2-2. Tornado Missile Impact Model.

Security-Related Information Figure Withheld Under 10 CFR 2.390.

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SAR-II-8.4 CH 8 Rev. 6 8-89 of 8-98 Figure 8.2-3. Horizontal and Vertical Seismic Design Response Spectra.

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SAR-II-8.4 CH 8 Rev. 6 8-90 of 8-98 Figure 8.2-4. TMI-2 ISFSI Volcanic Mitigation Map and Cross Section

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SAR-II-8.4 CH 8 Rev. 6 8-91 of 8-98 8.3 Load Combination Evaluation The load categories associated with normal operating conditions, off-normal conditions, and postulated accident conditions are described and analyzed in previous sections. The load combination results for the NUHOMS-12T components important to safety are presented in this section. Fatigue effects on the DSC confinement boundary are also addressed in this section.

8.3.1 DSC Confinement Boundary Load Combination Evaluation As described in Section 3.2, the stress intensities in the DSC confinement boundary at various critical locations for the appropriate normal operating condition loads are combined with the stress intensities experienced by the DSC confinement boundary during postulated accident conditions. It is assumed that only one postulated accident event occurs at any one time. Since the postulated cask drop accidents are by far the most critical, the load combinations for these events envelope all other accident event combinations.

Table 8.3-1 through Table 8.3-3 tabulate the maximum stress intensity for each component of the DSC confinement boundary calculated for the enveloping normal operating, off-normal, and accident load combinations. For comparison, the appropriate ASME Code allowables are also presented in these tables.

8.3.2 DSC Confinement Boundary Fatigue Evaluation Table 8.3-1 through Table 8.3-3 present the calculated enveloping normal, off-normal and accident stress intensities for the DSC confinement boundary components. Fatigue effects on the DSC confinement boundary are addressed using the criteria contained in NB-3222.4 of the ASME Code [8.19]. Fatigue effects need not be specifically evaluated provided the six criteria contained in NB 3222.4(a) are met. As demonstrated in Appendix C.4.1 of the NUHOMS generic SAR [8.9], an evaluation using these six criteria has been performed to show that the ASME Code fatigue requirements are satisfied for the DSC confinement boundary.

8.3.3 MP187 Cask Load Combination Evaluation The MP187 cask load combination evaluations are addressed in the MP187 cask SAR [8.18] and the Rancho Seco ISFSI SAR [8.12].

8.3.4 MP187 Cask Fatigue Evaluation Fatigue effects on the MP187 cask are addressed in the MP187 cask 10 CFR Part 72 evaluation presented in the Rancho Seco ISFSI SAR [8.12].

8.3.5 HSM Load Combination Evaluation The maximum bending moments and shear forces induced in the HSM for the individual normal and off-normal loads are listed in Table 8.1-15. Similarly, the maximum moments and shears induced in the HSM for the individual accident loads are listed in Table 8.1-15. As described in Section 3.2.5.l, the load combination procedure of Section 6.17.3.1 of ANSI 57.9 [8.2] is used to combine the factored normal operation, off-normal, and postulated accident loadings imposed on the reinforced concrete HSM. Many of the general event combinations, shown in Table 3.2 4, are enveloped by others that contain the same load factors with additional applied load cases.

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SAR-II-8.4 CH 8 Rev. 6 8-92 of 8-98 The governing calculated bending moments and shears for each load combination are tabulated in Table 8.3-4. The tabulated results represent the bounding shears and moments for either a single free-standing HSM or the array of HSMs. For comparison, the ultimate moment and shear capacity of the HSM for the controlling load combinations are also shown in Table 8.3-4. Comparison of the reported bending moment and shear for each load combination with the corresponding ultimate capacity shows that the design strength of the HSM is greater than the strength required for the most critical load combination.

8.3.6 Thermal Cycling of the HSM The largest mean daily change of temperature at the TMI-2 ISFSI site is 50°F. Because of the massive concrete sections used in the HSM, a period of time greater than one week is needed to obtain steady state temperatures and a steady state thermal gradient. For conservatism, it is assumed that the 50°F maximum daily change could produce a steady state gradient every day for 50 years, for a total of 18,250 thermal cycles. For the TMI-2 HSM, the maximum moment caused by the worst-case steady state normal operating thermal loads is bounded by the steady state off-normal operating thermal loads. The maximum moment caused by the off-normal operating thermal load for the concrete components of the HSM is 928.8 kip-in/ft and occurs in the rear wall. This loading is 46% of the minimum ultimate strength of the roof slab. Assuming this load amplitude is cycled daily, and referring to the S-N curve of Figure 6-46 of Reference 8.7, the number of cycles that occur before failure is greater than 10,000,000. Since this value is far greater than the postulated worst case 18,250 cycles, thermal cycling has a negligible effect on the HSM reinforced concrete.

8.3.7 DSC Support Structure Load Combination Evaluation The applicable loads for the DSC support structure inside the HSM include the DSC and support assembly dead weight, seismic loads, and DSC handling loads. Three load combinations are evaluated. Load combination one consists of the DSC plus the support structure dead weight, plus the DSC handling loads for a typical normal operating load case. Load combination two includes the dead weight of the support structure plus DSC handling loads in the jammed condition representing an off-normal loading. The third load combination includes the total dead weight plus design basis seismic loads for an accident event. The resulting maximum stresses as compared to AISC code allowables [8.39] are shown in Table 8.3-5.

The same load combinations are used for the DSC support structure connecting elements. All end connection components are designed to meet the AISC Code requirements for these design loads. The structural steel design is based on the requirements of the AISC code, and the embedments are designed in accordance with the requirements of ACI 349.

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SAR-II-8.4 CH 8 Rev. 6 8-93 of 8-98 Table 8.3-1. NUHOMS-12T DSC Enveloping Load Combination Results for Normal and Off-Normal Loads (ASME Service Levels A and B).

DSC COMPONENTS STRESS TYPE STRESS (ksi)

CONTROLLING LOAD COMBINATION CALCULATED ALLOWABLE DSC Shell Primary Membrane 16.3 23.1 A

Membrane + Bending 32.7 34.6 A

Primary + Secondary 54.7 69.3 B

Inner Bottom Cover Plate Primary Membrane 1.8 23.1 A

Membrane + Bending 2.6 34.6 A

Primary + Secondary 3.9 69.3 B

Outer Bottom Cover Plate Primary Membrane 7.9 23.1 B

Membrane + Bending 28.2 34.6 B

Primary + Secondary 34.7 69.3 B

Top Cover Plate Primary Membrane 9.5 23.1 A

Membrane + Bending 14.0 34.6 A

Primary + Secondary 14.8 69.3 B

Top Shield Plug Primary Membrane 5.3 19.3 A

Membrane + Bending 7.8 29.0 A

Primary + Secondary 8.0 57.9 A

Table 8.3-2. NUHOMS-12T DSC Enveloping Load Combination Results for Accident Loads (ASME Service Level C).

DSC COMPONENTS STRESS TYPE STRESS (ksi)

CALCULATED ALLOWABLE DSC Shell Primary Membrane 16.4 34.6 Membrane + Bending 28.0 51.9 Inner Bottom Cover Plate Primary Membrane 2.2 34.6 Membrane + Bending 3.6 51.9 Outer Bottom Cover Plate Primary Membrane 5.1 34.6 Membrane + Bending 9.1 51.9 Top Cover Plate Primary Membrane 10.7 34.6 Membrane + Bending 15.5 51.9 Top Shield Plug Primary Membrane 5.4 32.8 Membrane + Bending 8.5 49.2

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SAR-II-8.4 CH 8 Rev. 6 8-94 of 8-98 Table 8.3-3. NUHOMS-12T DSC Enveloping Load Combination Results for Accident Loads (ASME Service Level D).

DSC COMPONENTS STRESS TYPE CALCULATED STRESS (ksi)

CALCULATED ALLOWABLE DSC Shell Primary Membrane 26.4 49.0 Membrane + Bending 51.6 70.0 Inner Bottom Cover Plate Primary Membrane 11.1 49.0 Membrane + Bending 15.0 70.0 Outer Bottom Cover Plate Primary Membrane 11.9 49.0 Membrane + Bending 20.5 70.0 Top Cover Plate Primary Membrane 17.7 49.0 Membrane + Bending 27.7 70.0 Top Shield Plug Primary Membrane 19.8 40.6 Membrane + Bending 46.5 58.0 Top Cover Plate Weld Primary 17.9 56.0 Bottom Cover Plate Weld Primary 13.7 56.0 Table 8.3-4. HSM Enveloping Load Combination Results.

LOAD (1)

COMBINATION LOADING COMBINATION DESCRIPTION GOVERNING LOAD (2, 3)

CAPACITIES Vmax (k/ft)

Mmax (k-in/ft)

Vu (k/ft)(4)

Mu (k-in/ft) 1 1.4D+1.7L+1.7Ro 60.4 834.2 100.9 1683.0 2

0.75(1.4D+1.7L+1.7To+1.7W) 20.0 2193.8 31.0 2576.0 3

0.75(1.4D+1.7L+1.7To+1.7Ro) 20.1 1471.2 31.0 1683.0 4

D+L+To+E 15.8 1791.9 31.0 2576.0 5

D+L+To+Wt 59.8 1197.2 100.9 1683.0 6

D+L+Ta+Ra 54.7 1265.3 100.9 1683.0 D =

Dead Weight E =

Earthquake Load L =

Live Load To =

Normal Condition Thermal Load Ta =

Off-Normal or Accident Condition Thermal Load Wt =

Tornado Wind and Missile Loads Ro =

Normal Handling Loads Ra =

Off-Normal Handling Loads W =

Normal Wind Load (1)

Load combinations are based on ANSI-57.9 as shown in Table 3.2.4.

(2)

Governing loads shown are irrespective of locations. Loads reported have minimum margin to design capacity.

(3)

Results of load combinations 2 through 6 are based on cracked section. Others based on uncracked sections.

(4)

See Table 3.2-4 for definition of load terms.

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SAR-II-8.4 CH 8 Rev. 6 8-95 of 8-98 Table 8.3-5. DSC Support Structure Enveloping Load Combination Results COMPONENT LOAD COMBINATION CALCULATED STRESS (1)

INTERACTION (2)

(Calc/Allowable)

ALLOWABLE(2)

Shear Stress (ksi)

AXIAL (ksi)

STRONG AXIS BENDING (ksi)

WEAK AXIS BENDING (ksi)

SHEAR (ksi)

SUPPORT RAIL Normal Operation D + L + Ro(3) 1.42 6.73 7.81 5.12 0.73 13.1 Off-Normal Operation D + L + Ra(3) 2.84 10.47 9.94 5.70 1.09(4) 17.4 Accident D + L + E(3) 0.02 6.23 5.36 4.91 0.051(5) 18.3 (1)

Maximum stresses reported irrespective of location.

(2)

Allowable stresses taken at 200°F for all combinations.

(3)

See Table 3.2-4 for definition of load terms.

(4)

The permissible interaction ratio for this load combination is 1.33.

(5)

The permissible interaction for this load combination is 1.6.

8.4 Site Characteristics Affecting Safety Analysis All site characteristics affecting the safety analysis of the NUHOMS-12T system are noted throughout this SAR where they apply. Specific site characteristics that were taken into consideration in the design were:

1.

Wind Loads

2.

Missile Loads

3.

Seismic Events

4.

Temperatures

5.

Precipitation

6.

Flooding

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SAR-II-8.4 CH 8 Rev. 6 8-96 of 8-98 8.5 References 8.1 U.S. Nuclear Regulatory Commission (U.S. NRC), "Standard Format and Content for the Safety Analysis Report for an Independent Spent Fuel Storage Installation (Dry Storage)," Regulatory Guide 3.48, August 1989.

8.2 American National Standard, "Design Criteria for an Independent Spent Fuel Storage Installation (Dry Storage Type)," ANSI/ANS 57.9-1984, American Nuclear Society, La Grange Park, Illinois, 1984.

8.3 Henry H. Bednar, P.E., Pressure Vessel Design Handbook, Van Nostrand Reinhold Co., 1981.

8.4 U. S. Nuclear Regulatory Commission, "Barrier Design Procedures," Standard Review Plan NUREG-0800, (Formerly NUREG-75/087) Revision 1, July 1981.

8.5 American Concrete Institute, Code Requirements for Nuclear Safety Related Concrete Structures and Commentary, ACI 349-85 and ACI 349R-85, American Concrete Institute, Detroit, MI, 1980.

8.6 C. Wang and C. G. Salmon, Reinforced Concrete Design, Third Edition, Harper and Row, New York, N. Y., 1979.

8.7 M. Fintel, Handbook of Concrete Engineering, Van Nostrand Reinhold Co., New York, N.Y., 1985.

8.8 Prestressed Concrete Institute, PCI Design Handbook, 2nd Edition, Prestressed Concrete Institute, 1978.

8.9 Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel, NUH-003, Revision 4A, VECTRA Technologies, Inc., June 1996, File No. NUH003.0103.

8.10 H. K. Hilsdorf, J. Kropp, and H. J. Koch, "The Effects of Nuclear Radiation on the Mechanical Properties of Concrete," Paper 55-10, Douglas McHenry International Symposium on Concrete and Concrete Structures, American Concrete Institute, Detroit, MI, 1978.

8.11 American Nuclear Society, "American National Standard Guidelines on the Nuclear Analysis and Design of Concrete Radiation Shielding for Nuclear Power Plants," ANSI/ANS-6.4-1977, American National Standards Institute, Inc., 1977.

8.12 Safety Analysis Report for the Rancho Seco Independent Spent Fuel Storage Installation, Sacramento Municipal Utility District, October 1993, USNRC Docket Number 72-11.

8.13 R. J. Roark and W. C. Young, Formulas for Stress and Strain, Sixth Edition, McGraw-Hill, New York, N.Y., 1989.

8.14 Climatography of the Idaho National Engineering Laboratory, DOE/ID-12118, 2nd Edition, December 1989.

8.15 SCALE 4.2, Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluation, NUREG/CR-0200, Revision 4, ORNL/NUREG/CSD-2/V2/R4, Prepared by Oak Ridge National Laboratory for USNRC, CCC-S45, Released November 1993.

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SAR-II-8.4 CH 8 Rev. 6 8-97 of 8-98 8.16 F. Kreith, Principle of Heat Transfer, Third Edition, Harper and Row Publishers.

8.17 TMI-2 Canister Dry Storage: Hydrogen Gas Generation and Transport Evaluation, GPU Nuclear, Inc., September 1996.

8.18 Safety Analysis Report for the NUHOMS-MP187 Multi-Purpose Cask, NUH-005, Revision 2, VECTRA Technologies, Inc., February 1996, USNRC Docket Number 71-9255.

8.19 American Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code,Section III, Division l, Subsections NB, NC, and Appendices, 1992 Edition with 1993 Addenda.

8.20 J. A. Buckholz, Scoping Design Analyses for Optimized Shipping Casks Containing 1-, 2-, 3-, 5-,

7-, or 10-Year Old PWR Spent Fuel, ORNL/CSD/TM-149, Oak Ridge National Laboratory/Union Carbide Nuclear Division, p. 95, 1983.

8.21 Safety Analysis Report for the NUPAC 125-B Fuel Shipping Cask, Nuclear Packaging, Inc.,

USNRC Docket Number 71-9200.

8.22 U.S. Atomic Energy Commission, "Design Basis Tornado for Nuclear Power Plants," Regulatory Guide 1.76, April 1974.

8.23 NUREG/CR-4461 Tornado Climatology of the Contiguous United States.

8.24 SECY-93-087 USNRC Policy Issue Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs, April 2, 1993.

8.25 R. P. Kennedy, Holmes and Narver, Inc., "A Review of Procedures for the Analysis and Design of Concrete Structures to Resist Missile Impact Effects," Nuclear & Systems Sciences Group, Anaheim, California, September 1975.

8.26 J. R. McDonald, K. C. Mehta, and J. E. Minor, "Design Guidelines for Wind-Resistant Structures,"

Institute for Disaster Research and Department of Civil Engineering, Texas Tech University, Lubbock, Texas, June 1975.

8.27 U.S. Atomic Energy Commission, "Design Response Spectra for Seismic Design of Nuclear Power Plants," Regulatory Guide 1.60, Revision 1, December 1973.

8.28 U.S. Atomic Energy Commission, "Damping Values for Seismic Design of Nuclear Power Plants,"

Regulatory Guide 1.61, October 1973.

8.29 R. D. Blevins, Formulas for Natural Frequency and Mode Shape, Van Nostrand Reinhold Co.,

New York, N.Y., 1979.

8.30 American Concrete Institute, Building Code Requirements for Reinforced Concrete (ACI 318-95)

ACI, Detroit, MI, 1995.

8.31 Swanson Analysis Systems Inc., ANSYS Engineering Analysis System User's Manual, Version 5.3A, Swanson Analysis Systems Inc., Pittsburgh, PA.

8.32 U. S. Nuclear Regulatory Commission, "Combining Modal Responses and Spatial Components in Seismic Response Analysis," Regulatory Guide 1.92, Revision 1, February 1976.

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SAR-II-8.4 CH 8 Rev. 6 8-98 of 8-98 8.33 Electric Power Research Institute, "The Effects of Target Hardness on the Structural Design of Concrete Storage Pads for Spent-Fuel Casks," EPRI Report NP-4830, October 1986.

8.34 TMI-2 Defueling Canisters, Final Design Technical Report, Document 77-1153937-04, Babcock and Wilcox Company, May 24, 1985.

8.35 ICPP Interim Storage System (ISS) Lightning Risk Assessment, INEL Engineering Design File 0566 dated Sept. 4, 1996.

8.36 Flammability Characteristics of Combustible Gases and Vapors Zabetakis, M.J., Bureau of Mines Bulletin 627, United States Department of the Interior, 1965.

8.37 TMI-2 Canister Stress Analysis, Babcock and Wilcox Company, Document No. 33-1153704.

8.38 ISFSI Fire Exposure Analysis, VECTRA File No. NUH004.0422.

8.39 American Institute of Steel Construction (AISC), Allowable Stress Design, 9th Edition.

8.40 Fisher, R.V., Heiken, G., and Hulen, J.B. (1997) Volcanoes, Crucibles of Change; Princeton University Press, Princeton, N.J., p. 137-146.

8.41 Barberi, F., Carapezza, M.L., Valenza, M., and Villari, L. (1993) The control of lava flow during the 1991-1992 eruption of Mt. Etna; Journal of Volcanology and Geothermal Research, v.56, p.1-34.

8.42 McPhee, J. (1991) The Control of Nature; Noonday Press, New York, chapter 2, Cooling the Lava.

8.43 N.B. Zolders, Thermal Properties of Concrete Under Sustained elevated Temperatures, ACI Publication, Paper SP25-1, American Concrete Institute, Detroit, MI, 1970.

8.44 ASHRAE Handbook, 1981 Fundamentals, Fourth Printing, American Society of Heating, Refrigerating, and Air-Conditioning Engineering, Inc., 1983.

8.45 Radiological Evaluation of TMI-2 ISFSI Technical Specification 3.1.1, Engineering Design File No.

4728, Revision 5, November, 2022.