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Printod inThis document presents a complete description of Cycle 2 of COBRA-SFS, and consists of three main parts. Part I describes the conservation equations, constitutive models, and solution methods used in the code. Part 11 presents the User Manual, with guidance on code applications, and complete input instructions. This part also includes a detailed description of the auxiliary code RADGEN, used to generate grey body view factors required as input for radiative heat transfer modeling in the code. Part I11 describes the code structure, platform dependent coding, and program hierarchy. Installation instructions are also given for the various platform versions of the code that are available.iii

SummaryThis report fulfills the M3 milestone M3FT-12PN0810041, "Report on Realistic Temperature Profiles", under Work Package FT-12PN081004.As part of the Used Fuel Disposition Campaign of the Department of Energy (DOE), visual inspections and temperature measurements were performed on two storage modules in the Calvert Cliffs Nuclear Power Station's Independent Spent Fuel Storage Installation (ISFSI). The inspection procedure included surface temperature measurements on one end of the DSC within the storage module. The data obtained in the inspections at Calvert Cliffs provide an opportunity to develop structural and thermal models that can yield realistic temperature predictions for actual storage systems, in contrast to conservative and bounding design basis calculations.Detailed models of the concrete storage modules to be examined were developed using STAR-CCM+ (version 7.02; CD-Adapco, 2012). The immediate purpose of this modeling effort is to obtain temperature predictions in actual storage conditions for the module, DSC, and DSC contents, including preliminary estimates of fuel cladding temperatures for the SNF. The long-term goal of this work is to obtain realistic evaluations of thermal performance of actual SNF storage systems over extended periods, which will require developing a detailed COBRA-SFS (Michener, et al., 1987) model of the DSC internals, in addition to the large system models. The approach used in this study omits many of the conservatisms and bounding assumptions normally used in design-basis and safety-basis calculations for spent fuel storage systems. The results of this study cannot be used in licensing basis evaluations of the Calvert Cliffs ISFSI, or any other spent fuel storage facility.The storage modules used for this study are HSM-1 and HSM-15 in the Calvert Cliffs Nuclear Power Station's ISFSI, each containing a 24P DSC loaded with 24 CE 14x14 spent fuel assemblies. The total decay heat load for the DSC in HSM-15 was 10.8 kW at the time of loading, and was calculated to be 7.6 kW as of June 2012. The total decay heat load for the DSC in HSM-1 was calculated to be 4.1 kW as of June 2012. The base case for thermal evaluation of the 24P DSC in HSM-15 assumed an ambient temperature of 58°F (14°C). This value was determined using historical climatology data from a National Oceanic and Atmospheric Administration (NOAA) database, and verified with annual ambient temperature data from monitoring stations at the Calvert Cliffs ISFSI. Bounding sensitivity studies on the effect of ambient air temperature were performed for two cases; a 'summer case' at 77°F based on average temperatures in July, and a 'winter case' at 35°F, based on average temperatures in January. Figure On June 27 th and 28 th , 2012, visual inspections, surface sampling, and temperature measurements were performed on HSM-1 and HSM-15 at the Calvert Cliffs Nuclear Power Station ISFSI. Due to physical constraints on the accessible regions of the DSC and considerations of worker safety, reliable temperature measuremen...

In 2007, a severe transportation accident occurred near Oakland, California, at the interchange known as the “MacArthur Maze.” The accident involved a double tanker truck of gasoline overturning and bursting into flames. The subsequent fire reduced the strength of the supporting steel structure of an overhead interstate roadway causing the collapse of portions of that overpass onto the lower roadway in less than 20 minutes. The US Nuclear Regulatory Commission has analyzed what might have happened had a spent nuclear fuel transportation package been involved in this accident, to determine if there are any potential regulatory implications of this accident to the safe transport of spent nuclear fuel in the United States. This paper provides a summary of this effort, presents preliminary results and conclusions, and discusses future work related to the NRC’s analysis of the consequences of this type of severe accident.

In this effort, an assessment of bulk ultrasonic (UT) and eddy current (ECT) methods and techniques is performed for inspecting the surfaces of dry cask storage systems (DCSSs) canisters. Some DCSS canisters (especially those located in coastal environments) will be exposed to environmental conditions, which can cause atmospheric stress corrosion cracking (SCC). Information collected from the field and from laboratory studies has not been able to rule out the possibility of atmospheric SCC in DCSS canisters, although no occurrences of atmospheric SCC in DCSS canisters have been detected. UT and ECT methods and techniques are already used to inspect nuclear power plant components and this experience, along with their relative maturity, makes these methods and techniques likely frontrunners for near-term application to examination of dry storage canister surfaces. In this report, the results of several performance reliability studies for UT and ECT are reviewed. The detection, depth-sizing, and lengthsizing results are documented and summarized to quantitatively estimate the adequacy of UT and ECT for inspecting dry storage canister surfaces. In addition, this effort focuses on the implementation of NDE methods and techniques in the Holtec HI-STORM 100 system and the Transnuclear NUHOMS horizontal storage modules and considers environmental compatibility, accessibility constraints, and NDE sensor deployment options for these systems.v

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On July 18, 2001, a freight train carrying hazardous (non-nuclear) materials derailed and caught fire while passing through the Howard Street railroad tunnel in downtown Baltimore, Maryland. The United States Nuclear Regulatory Commission (USNRC), one of the agencies responsible for ensuring the safe transportation of radioactive materials in the United States, undertook an investigation of the train derailment and fire to determine the possible regulatory implications of this particular event for the transportation of spent nuclear fuel by railroad.The USNRC met with the National Transportation Safety Board (NTSB) to discuss the details of the accident and the ensuing fire. Following these discussions, the USNRC assembled a team of experts from the National Institute of Standards and Technology (NIST), the Center for Nuclear Waste Regulatory Analyses (CNWRA), and Pacific Northwest National Laboratory (PNNL) to determine the thermal conditions that existed in the Howard Street tunnel fire and analyze the potential effects of those conditions on various spent nuclear fuel transportation package designs.The Fire Dynamics Simulator (FDS) code developed by NIST was used to determine the thermal environment in the Howard Street tunnel during the fire. The FDS results were used as boundary conditions for the COBRA-SFS and ANSYS ® computer models developed to evaluate the thermal performance of different package designs. The staff concluded that larger transportation packages resembling the TransNuclear Model No. TN-68 and HOLTEC Model No. HI-STAR 100 would withstand a fire with thermal conditions similar to those that existed in the Baltimore tunnel fire event with only minor damage to peripheral components. This is due to their sizable thermal inertia and design specifications in compliance with currently imposed regulatory requirements.For the TN-68 and the NAC International Model No. LWT (legal weight truck) transportation package, the maximum temperatures predicted in the regions of the lid and the vent and drain ports exceed the seals' rated service temperatures, making it possible for a small release to occur, due to CRUD that might spall off the surfaces of the fuel rods. While a release is not expected to occur for these conditions, any release that could occur would be very small due to a number of factors. These include (1) the tight clearances maintained between the lid and cask body by the closure bolts, (2) the low pressure differential between the package interior and exterior, (3) the tendency of such small clearances to plug, and (4) the tendency of CRUD particles to settle or plate out.USNRC staff evaluated the radiological con-sequences of the package responses to the Baltimore tunnel fire. The analysis indicates that the regulatory dose rate limits specified in 10 CFR 71.51 for accident conditions would not be exceeded by releases or direct radiation from any of these packages in this fire scenario. All three packages are designed to maintain regulatory dose rate limits even with a complete ...

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