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At least 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Dry Cask Shipments of ATR Spent Nuclear Fuel at INL - 19600

The High Load Charger (HLC) cask will be replacing the Advanced Test Reactor (ATR) spent fuel transfer cask that has been used at Idaho National Laboratory (INL) to wet-transfer spent ATR fuel elements from ATR to Idaho Nuclear Technology and Engineering Center (INTEC) in order to support future plans at both facilities. INTEC intends to discontinue use of the CPP-666 storage pools in compliance with the Idaho Settlement Agreement and the basins will no longer be available to store spent ATR fuel elements. In order to support the core-internals change out in 2021 for ATR, spent ATR fuel elements must be removed from the canal and shipped to INTEC. INL contractors will use the HLC cask for safe, dry transfer of spent Advanced Test Reactor (ATR) fuel elements between the ATR Complex managed by Battelle Energy Alliance, LLC and the Idaho Nuclear Technology and Engineering Center (INTEC) managed by Fluor Idaho, LLC. The HLC cask (also known as the High Load Charging Cask or High Load Charger) is a lead-shielded, stainless steel transfer container originally designed in 1954 by the Idaho Nuclear Corporation to meet the radiation and contamination controls requirements of the time for shipment of spent nuclear fuel. The cask has been modified over the years and is currently used for wet-to-dry transfers at INTEC. It is approved only for out-of-commerce transfers. The cask will arrive at the ATR where eight spent ATR fuel elements will be loaded into two, newly-designed buckets. Each bucket is a quarter section, right-cylinder shape that has four compartments. The cask will be lifted out of the ATR canal allowing most of the water to drain from the cask. Processing the 12,500 kg cask will be completed in ATR and then transported on a tractor trailer to INTEC where the loaded buckets will be removed from the cask. The loaded buckets will undergo a drying process before placing the four buckets into a canister for long-term storage in the Irradiated Fuel Storage Facility (IFSF) at INTEC. A new transportation plan that was submitted to, and approved by, the U.S. Department of Energy to provide the cask safety basis, which describes the hazard analysis and accident analysis for the cask during transport. Supporting accident analysis include criticality, thermal, dose, and drop analyses. Of particular interest for this cask was the drop analysis from which was derived the speed limit for the transport. The drop analysis addressed drops in air onto an unyielding surface and in water onto an energy-absorbing pad for the cask in the ATR facility. These analyses are typically applied as the bounding drop scenarios during transport; however, this resulted in a safety basis speed limit of 12.9 kph. In order to more accurately represent the in-transit accident scenario, LS-DYNA software was procured and the model was updated to more closely represent the potential drop heights and impact surfaces of asphalt and soil for the cask while in transit. This analysis resulted in a safety basis speed limit of 48 kph. This safety basis speed limit is more practical to implement and control. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Optimization of Dry Storage Canister Cutting Operations - 20534

A recent project investigated the optimal cutting method for the opening of a welded dry storage/shielded canister (DSC) or dual purpose canister (DPC) containing Used Nuclear Fuel (UNF). Due to the lack of a disposal path for UNF in the U.S., the majority of UNF is currently moved into welded DSCs and DPCs designed by Orano TN, NAC, and Holtec. As the DSCs/DPCs were neither designed nor licensed for disposal and may not be able to be emplaced in a geologic repository due to physical emplacement constraints, near-term thermal limitations, or long-term criticality issues, the UNF in these existing DSCs/DPCs may need to be repackaged into transportation, aging, and disposal canisters (TADs), generic standard TADs (STADs), transportation casks, new cask/canister systems, and/or cask/canister/package systems suitable for disposal. These DSCs/DPCs may also be opened to simply remove the UNF in preparation for recycling, re-storage, or placement into a waste package suitable for disposal. Regardless of need, the DSCs/DPCs would require to be cut open by a potentially large scale operation. This project examined the multiple means for cutting welded metal systems and considered several factors to determine an optimal approach. Issues arose around some of the cutting approaches when applied to systems with 2 or 3 lids and with pre-cutting gas testing required. The resulting preferences in cutting methods with an optimized application for the opening of DSCs/DPCs are presented. (authors)

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DOE OSTI2015

New container development for the international market - 15656

Designs of casks and containers for the disposal of radioactive waste have to meet safety requirements for transport and storage. Since nuclear law is a national issue, the legal specifications often differ significantly. To improve the flexibility of a given basic design, the development of a set of generic design elements and the application of favorable material characteristics are of major importance. As an example for the international approach the design development for an IP-2 container for radioactive non-fuel waste made of Ductile Cast Iron (DCI) for the UK market in the frame of the specific national waste disposal concept is described. The results of drop tests performed during the licensing procedure and of parametric analysis work regarding a wide spectrum of accidental scenarios and impacts show the excellent behavior of the container concept and confirm the related favorable manufacturing procedures. Furthermore, the development work provided a significant set of proposals for the future detailed work. (authors)

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DOE OSTI2015

A novel application of cosmic rays for spent nuclear fuel monitoring - 15738

Monitoring nuclear waste and controlling spent nuclear fuel at its source is one of the main strategies to minimize the risks of nuclear proliferation and reduce potential homeland threats. Storage of high level waste from a fission reactor typically involves placing spent fuel in sealed dry casks. After the spent nuclear fuel has been placed inside the dry cask, the cask is welded, not allowing for visual inspection. Conventional methods for examining the interior of materials e.g., X-rays, are limited by the fact that they cannot penetrate very dense well-shielded objects. Instead, cosmic ray muons have the potential to allow for assessment of spent nuclear fuel stored within sealed dense dry casks. Objectives: 1.Exploit the ability of cosmic ray muons to penetrate high density materials allowing the distribution of materials within the object to be inferred from the muon tracks. 2. Develop a monitoring system that tracks the path of cosmic ray muons, including their deflection and attenuation, to identify the contents of a closed dry cask. Methods: - Monte-Carlo simulations using Geant4; - Bayesian decision theory for signal classification; - Wavelet analysis for noise cancellation; - Chaotic Neural Networks for signal encryption. Results: 1. Preliminary statistical analysis proved that the method is feasible and has the ability to identify the presence of spent nuclear fuel in a dry cask. 2. Using monoenergetic muons, a dry cask with one fuel assembly missing can be correctly identified with error rate <1%. Broader merit: Once completed, the proposed project for cosmic ray muon transport will enable a new and potentially transformative approach towards efficient, inexpensive and remote cosmic ray muon applications in archaeology, medicine and safeguarding of nuclear materials.

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DOE OSTI2015

The German product quality control of the compacted metallic nuclear waste - 15311

The treatment and safe disposal of spent nuclear fuel is a key challenge of nuclear energy utilization and has to be accomplished regardless the future nuclear of electricity production or phase-out policies. Until June 2005 approx. 6662 tons of spent fuel from German nuclear power plants was sent to the La Hague (France) and Sellafield (UK) plants for reprocessing. An equivalent amount of nuclear waste compounds has contractually to be returned to Germany, which is currently in progress. The returnable waste streams comprise mainly highly-active vitrified waste containing a major part of fission products and actinides, and intermediate-active super-compacted metallic waste containing reprocessed fuel element's structural materials and some technological waste from the reprocessing process itself. Shearing and dissolution of spent fuel leaves scraps of fuel metal cladding, the end-pieces and spacers which are contaminated, activated and irradiating from their life in a reactor core. In fact, this metallic intermediate level radioactive waste makes up a bit more than half of the total volume of the reprocessed waste. At La Hague the metallic waste is super-compacted. Several, typically eight, compacted discs are stacked in a stainless steel container, named CSD-C as for 'Colis Standard de Dechets Compactes'. 4104 containers of such CSD-Cs in 152 transport/storages casks have to be returned to Germany. Each one of them has to be checked and quality certified for its conformity with German national product specifications and acceptance criteria by independent national product quality control inspectors (PKS). The residue package documentation has to be complete and the characterization methods have to be comprehensive. Due to the heterogeneity of metallic waste the characterization and control of its properties is very complex. The nuclear characterization is performed by means of non-destructive measurements, where only a few key nuclides are measured directly. In fact, the proper calculation of the whole nuclear inventory relies on numerous supplementary simulations. The measurement and calculation methodology is scrutinized by PKS in detail and key points are benchmarked with our own simulations for the sake of an independent nuclear inventory declaration check. With respect to the vast amount of documentation and parameters, which have to be checked and verified, PKS had to optimize their checking procedures. For this purpose dedicated proof tools have been developed that shall be applied for selected indicative cross-checks of the relevant parameters. This contribution is dedicated to the procedure of quality control by PKS and examples of actually performed computational verifications of declared nuclear inventory. (authors)

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DOE OSTI2019

SNL Contribution: Consequence Analysis for Moisture Remaining in Dry Storage Canisters After Drying

This report discusses several possible sources of water that could persist in SNF dry storage canisters through the drying cycle. In some cases, the water is trapped in occluded geometries in the cask such as dashpots or damaged fuel. Persistence of water or ice in such locations seems unlikely, given the high heat load of the canistered fuel; this is especially true in the case of vacuum drying, where a strong driver exists to remove water vapor from the headspace of such occluded geometries. Water retention in Boral® core material is a known problem, that has in the past resulted in the need for much extended drying times. Since the shift to slightly higher porosity "blister resistant" Boral®, water drainage appears to be less of a problem. However, high surface areas for the Boral® core material will provide a trap for significant amounts of adsorbed water, at least some of which is certain to survive the drying process. Moreover, if corrosion within the cores produces hydrous aluminum corrosion products, these may also survive.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Fission Gases in Spent Fuel Pool Criticality Analysis

Historically all fission products, including fission gases have been included in the criticality analysis of many spent fuel pools (SFP). For transport casks ISG-8 Rev. 3 states, 'Credit for neutron absorbing isotopes other than those identified in Tables 1 and 2 should be accompanied by assurance that such isotopes are non-volatile, and non-gaseous,..' Recent questions from the NRC have challenged using fission gases in spent fuel pool criticality analyses, noting that the time the fuel is contained in the spent fuel pool can be long. This paper establishes a conservative approach for using some of the fission gases in the criticality analysis. Calculations have been performed for Indian Point Unit 2 to determine the impact of the fission gas release. The calculations have been run with the noble gases reduced by 32%, the halogens reduced by 10% and the Rb reduced by 44%. At 43 GWd/MTU, the impact of reducing the isotopes described here is only 0.0014 in k. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

A burnup credit approach for margin estimation of loaded boiling water reactor canisters in UNF-ST-DARDS

This paper discusses the modeling approach applied in the Used Nuclear Fuel Storage, Transportation, and Disposal Analysis Resource and Data System (UNF-ST-DARDS) to take credit for the reduced reactivity associated with the burnup of boiling water reactor (BWR) fuel for canister-specific as-loaded criticality analysis. Burnup credit is routinely used for the analysis of pressurized water reactor (PWR) spent nuclear fuel (SNF) storage (wet), transportation, and disposal. The fresh fuel assumption has historically been applied to BWR SNF criticality analyses for dry storage and transportation, and the peak reactivity method of burnup credit has been applied to pool storage. The fresh fuel assumption ignores the credit for gadolinium burnable absorbers in BWR fuel, as well as the burnup of the fuel, and is therefore conservative. This work employs the framework traditionally used for PWR burnup credit analyses and applies available data and recently published regulatory guidance to justify a new approach to model the as-loaded burnup and enrichment of SNF assemblies in currently loaded dry casks. The following features of this criticality analysis approach are presented in this paper: (1- selection of axial burnup profiles for BWR fuel from publicly available sources using methods derived from recently published research, (2- justification of the selected axial burnup profiles as relevant to all types of BWR fuel, (3- justification for modeling the fuel assemblies with a uniform axial and radial enrichment, and (4- justification for modeling the axial void profile and control blade insertion during depletion. This analysis approach was used to evaluate 153 already-loaded BWR casks at 8 sites as a function of time in UNF-ST-DARDS. Finally, the results of the as-loaded criticality safety margin assessments are presented. The results show the k{sub eff} calculated for storage and transportation cases range from 0.71 to 0.83, and the eigenvalues calculated for the disposal cases range from 0.85 to 1.03, though the highest cases are dominated by very limiting hypothetical damaged fuel assumptions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Development of the B-25 Low Level Radioactive Waste Container in Taiwan - 1900x

Chinshan Nuclear Power Plant will be permanently stopped its operation in the middle of 2019 and will be the first nuclear power plant stepping into the decommissioning phase in Taiwan. According to Nuclear Reactor Facilities Regulation Act, a nuclear facility must be dismantled in 25 years after being permanently ceased operation. In its decommissioning plan, the B -25 low level radioactive waste container was one of the proposed containers used for this decommissioning task. However, the B-25 waste container was not used or developed in Taiwan before. To meet this need, Institute of Nuclear Energy Research developed a new type of B-25 waste container and will apply to the authority for its usage license. This article provided the information of the low level radioactive waste containers have been approved in Taiwan and described the design specifications of the developed B-25 waste container. Besides, the numerical evaluation and test results were also included in this article. Based on the evaluation and test results, the developed B-25 waste container satisfied the requirements of the storage container and industrial package type 2. It is believed that the usage license would be granted soon after applying it to the authority. (authors)

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DOE OSTI2015

Evaluation of DUSTRAN Software System for Modeling Chloride Deposition on Steel Canisters

The degradation of steel by stress corrosion cracking (SCC) when exposed to atmospheric conditions for decades is a significant challenge in the fossil fuel and nuclear industries. SCC can occur when corrosive contaminants such as chlorides are deposited on a susceptible material in a tensile stress state. The Nuclear Regulatory Commission has identified chloride-induced SCC as a potential cause for concern in stainless steel used nuclear fuel (UNF) canisters in dry storage. The modeling of contaminant deposition is the first step in predictive multiscale modeling of SCC that is essential to develop mitigation strategies, prioritize inspection, and ensure the integrity and performance of canisters, pipelines, and structural materials. A multiscale simulation approach can be developed to determine the likelihood that a canister would undergo SCC in a certain period of time. This study investigates the potential of DUSTRAN, a dust dispersion modeling system developed by Pacific Northwest National Laboratory, to model the deposition of chloride contaminants from sea salt aerosols on a steel canister. Results from DUSTRAN simulations run with historical meteorological data were compared against measured chloride data at a coastal site in Maine. DUSTRAN’s CALPUFF model tended to simulate concentrations higher than those measured; however, the closest estimations were within the same order of magnitude as the measured values. The decrease in discrepancies between measured and simulated values as the level of abstraction in wind speed decreased suggest that the model is very sensitive to wind speed. However, the influence of other parameters such as the distinction between open-ocean and surf-zone sources needs to be explored further. Deposition values predicted by the DUSTRAN system were not in agreement with concentration values and suggest that the deposition calculations may not fully represent physical processes. Overall, results indicate that with parameter refinement, DUSTRAN has the potential to simulate atmospheric chloride dispersion on steel canisters.

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DOE OSTI2018

An Industrial Perspective of Research Within the DISTINCTIVE Program - 18045

UK Engineering and Physical Sciences Research Council (EPSRC) and three industrial partners, the Nuclear Decommissioning Authority (NDA), the National Nuclear Laboratory (NNL) and Sellafield Ltd (SL) are supporting the DISTINCTIVE (Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories) consortium of eleven universities to carry out research in four theme areas of importance to the UK nuclear industry: Advanced Gas-cooled Reactor (AGR), Magnox and exotic spent fuels; PuO{sub 2} and fuel residues; legacy ponds and silo wastes; and structural integrity. This paper describes the main industry challenges in each of these theme areas and some of industry led activities to address these challenges, and provides examples of where the DISTINCTIVE research output is assisting the nuclear industry in meeting these challenges. (authors)

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DOE OSTI2020

Cask RT-100 Type B(U) Self-Certification for DOT 7A, Type A Nuclear Waste Transportation - 20098

The purpose of this Type A certification is to diversify the use of the RT-100 cask by allowing the end user to transport either Type A or Type B radioactive waste in accordance with the handling procedures for each waste category. The project's scope included the development of a matrix that contains all applicable requirements for a Type A package from the Code of Federal Regulations (CFR), Title 49 for the Department of Transportation (DOT) with a justification for each requirement. Moreover, the scope included the writing of conditions and restrictions list for the transportation of Type A nuclear waste, the writing of a compliance statement, and the design and fabrication of a Type A placard to be permanently placed on the RT-100 cask. Type A packages are self-certified by the end user using one of three methods as permitted by 49 CFR 173.461, either by performance of tests with prototypes or samples of the specimens representing the packaging, or reference to a previous satisfactory demonstration of compliance of a sufficiently similar nature, or calculations or reasoned evaluation using reliable and conservative procedures and parameters. Since the RT-100 cask was licensed by the Nuclear Regulatory Commission (NRC) as a Type B cask, Robatel relied on the method of reference to a previous satisfactory demonstration of compliance of a sufficiently similar nature for Type A certification. This method was determined to be acceptable, since Type B packages are evaluated for more demanding transport conditions than those required for Type A packages. This paper presents Robatel's process for certifying the RT-100 cask for Type A shipments, an overview of the DOT regulations related to Type A packages, and the documents that were generated for the end user. This paper is concluded with a discussion on the advantages of having the option to transport both Type A and Type B nuclear waste, and possible future projects to further improve the operation of the cask when making Type A shipments. (authors)

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DOE OSTI2018

Design and Licensing of Irradiated Uranyl Nitrate Liquid Containers and Applicable Loading and Packaging System - 18481

Irradiated highly enriched uranyl nitrate liquid (HEUNL) solutions are typically residual materials resulting from the processing of uranium target materials in the production of radioisotopes. Traditionally, these residual materials have been processed and stored at the site of origin, due to the lack of an option for packaging, transportation and disposition. Modern alternatives to process, stabilize, package and dispose of these materials have been developed, but until recently there has not been a viable Type B package solution that could be used to ship these materials in sufficient quantities to justify the cost of an off-site disposition strategy. For this reason, most facilities that generate irradiated HEUNL materials have implemented on-site storage tanks or containers, resulting in the significant accumulation of HEUNL inventories worldwide. This paper provides an overview of the design and licensing activities for the development of sealed irradiated HEUNL transport containers to permit safe transportation of this material in the NAC Legal Weight Truck (NAC-LWT) cask. A description of the robust HEUNL container, licensing and fabrication is also presented. In addition, NAC has developed a specialized loading system to optimize the process to load, package and un-package HEUNL materials to minimize worker exposure during loading, and execute the packaging process efficiently for safe transport and retrieval. (authors)

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DOE OSTI2021

Aerosol Deposition Inside a Spent Fuel Transportation and Storage Canister (Interim Report)

This report documents the progress in preparing the ANSYS/Fluent ® and MELCOR models to perform characterization studies of aerosols dispersal and deposition within a select spent fuel cask system. Steady state thermal-hydraulic cask response is modeled with both codes, at present, while the MELCOR source code is being modified to allow imposed thermal-hydraulic conditions with aerosol physics calculations. This will allow the MELCOR model to assume the thermal-hydraulic calculation from the ANSYS/Fluent ® , while only computing the aerosol physics. Detailed results are presented on the thermal-hydraulic analysis of the MAGNASTOR ® cask for the current ANSYS/Fluent ® model, with convergent conditions observed over two fidelities. While the MELCOR computation computes the steady state conditions, they differ sufficiently from that computed with ANSYS/Fluent ® . The MELCOR analysis include a set of UO 2 sources to investigate system response for interim reporting. The airborne concentrations and evolving distributions are presented. Model development is anticipated to continue as additional components impacting flow conditions or aerosol deposition within the MAGNASTOR ® cask and Westinghouse fuel assemblies are identified.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Feasibility Study For Use Of Commercial Cask Vendor Dry Transfer Systems To Unload Used Fuel Assemblies In L-Area

The purpose of this study is to determine whether a commercial dry transfer system (DTS) could be used for loading or unloading used nuclear fuel (UNF) in L-Basin and to determine if a DTS pool adapter could be made for L-Basin Transfer Pit #2 that could accommodate a variety of DTS casks and fuel baskets or canisters up to 24” diameter.[1, 2] This study outlines the technical feasibility of accommodating different vendor dry transfer systems in the L-Basin Transfer Bay with a general work scope. It identifies equipment needing development, facility modifications, and describes the needed analyses and calculations. After reviewing the L-Basin Transfer Bay area layout and information on the only DTS system currently in use for the Nuclear Assurance Corporation Legal Weight Truck cask (NAC LWT), the authors conclude that use of a dry transfer cask is feasible. AREVA was contacted and acknowledged that they currently do not have a design for a dry transfer cask for their new Transnuclear Long Cask (TN-LC) cask. Nonetheless, this study accounted for a potential future DTS from AREVA to handle fuel baskets up to 18” in diameter. Due to the layout of the Transfer Bay, it was determined that a DTS cask pool adapter designed specifically for spanning Pit #2 and placed just north of the 70 Ton Cask lid lifting superstructure would be needed. The proposed pool adapter could be used to transition a fuel basket up to 24” in diameter and ~11 feet long from a dry transfer cask to the basin. The 18” and 24” applications of the pool adapter are pending vendor development of dry transfer casks that accommodate these diameters. Once a fuel basket has been lowered into Pit #2 through a pool adapter, a basket cart could be used to move the basket out from under the pool adapter for access by the 5 Ton Crane. The cost to install a dry transfer cask handling system in L-Area capable of handling multiple vendor provided transport and dry transfer casks and baskets with different diameters and lengths would likely be on the same order of magnitude as the Basin Modifications project. The cost of a DTS capability is affected by the number of design variations of different vendor transport and dry transfer casks to be considered for design input. Some costs would be incurred for each vendor DTS to be handled. For example, separate analyses would be needed for each dry transfer cask type such as criticality, shielding, dropping a dry transfer cask and basket, handling and auxiliary equipment, procedures, operator training, readiness assessments, and operational readiness reviews. A DTS handling capability in L-Area could serve as a backup to the Shielded Transfer System (STS) for unloading long casks and could support potential future missions such as the Idaho National Laboratory (INL) Exchange or transferring UNF from wet to dry storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Richard Hodes Memorial Award Lecture - Development and Certification of New Type-B Containers for the Recovery of High-Activity Sealed Sources - 18710

The US Department of Energy (DOE) National Nuclear Security Administration (NNSA) Off-Site Source Recovery Program (OSRP) has successfully completed a project to design, test, and certify two new Type-B transportation packages. Since about 2011 a shortage of these containers has significantly constrained the recovery and disposal of commercially licensed high activity radioactive sealed sources. These new packages will enable NNSA/OSRP to transport a majority of the high-activity Cs-137 and Co-60 devices encountered in the coming years and decades. The 435-B is an unshielded, leak-tight container, which can be used to ship common shielded devices, such as the teletherapy source drawers and irradiators that contain high-activity Co-60 and Cs-137 sources. The 380-B is a much larger, leak-tight, lead-shielded Type B cask capable of transport by truck, rail, ship, or air. Additionally, to encourage and facilitate the development of transportation capacity, DOE/NNSA intends to make the design documentation available at no cost to commercial vendors. (authors)

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DOE OSTI2018

Drop Testing of a Container for the Storage, Transport and Disposal of Intermediate Level Waste - 18078

Impact tests were performed, as part of a corresponding container's substantiation, during design development of a shielded container proposed for use at Sellafield for waste retrieval from the First Generation Magnox Storage Pond (FGMSP) in the UK. The mechanical test program comprised a 9 m free drop test onto an unyielding target in a container long lid edge down orientation at ambient conditions. Further, a 0,5 m free drop test onto a punch target was performed. Here, the container was orientated with the lid downwards in a way that the punch, a mild steel bar, impacts a filter lid. The test specimen was instrumented with strain gauges and accelerometers for the drop tests. Transient strains at selected points of the inner and outer container walls, at the shielding lid, as well as at the lid bolts were measured during the container's impact. Furthermore, decelerations of the container body, container lid, and the skip were measured. The complex geometrical changes of the container due to impact were determined by optical 3d- deformation measures using the projected fringes method in combination with multi-image photogrammetry. This paper summarizes the performance of the drop tests and various drop test results in context with the design development as well as aspects regarding the associated Finite Element (FE) analyses and post-test evaluation. (authors)

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DOE OSTI2016

Coupling the Modeled Structural Transmissibility of a Used Nuclear Fuel Conveyance to Over the Road Data - 16215

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct research and development activities related to storage, transportation, and disposal of used nuclear fuel (UNF) and high-level radioactive waste (HLW). The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel and HLW generated by existing and future nuclear fuel cycles. The Storage and Transportation staff within the UFDC is responsible for addressing issues regarding the long-term or extended storage (ES) of UNF and its subsequent transportation. Current information is insufficient to determine the ability of UNF, including high-burnup fuel, to withstand shock and vibration loads that could occur when UNF is shipped by rail from nuclear power plant sites to a storage or disposal facility after extended storage. In order to make this determination, the magnitude of the transportation loads transmitted to the UNF must be quantified. Previous preliminary modeling work has shown how the structural transmissibility of the transport system can affect the magnitude of these loads and the importance of modeling all aspects of the transport system (i.e. rail car, transport cradle, cask, canister, and fuel). The work presented herein proposes a methodology for determining the structural transmissibility of a hypothetical transport system, and conditions existing over the road test (OTR) data to predict the performance of the hypothetical system. This method utilizes the modeled frequency response of the hypothetical system to scale the existing OTR acceleration data. This work will be relevant in creating models of a UNF conveyance during transport. As such, preliminary work in coupling the rail vehicle dynamics code NUCARS{sup C}, a general purpose multi-body rail vehicle dynamics simulation package developed by Transportation Technology Center Inc. (TTCI), to existing models of UNF during transport are also discussed herein. (authors)

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