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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Developments in the Regulatory Framework of Interim Storage of Spent Nuclear Fuel (SNF) in Germany - 18137

According to article 6 of the German Atomic Energy Act, storage of nuclear fuel requires a license granted by the competent authority in this field, which is the Federal Office for Safety of Nuclear Waste Management (BfE). In 1995, a license for storage of spent nuclear fuel (SNF) and vitrified high level waste (HAW) was granted for the interim storage facility in Gorleben. Dry Interim storage at the NPP sites in its current form started in 2002 in the interim storage facility next to the nuclear power plant NPP Lingen. Since this time twelve NPPs erected their own storage facilities additionally to the three central storage facilities (Ahaus, Gorleben and Rubenow). In all of these facilities the concept of dry storage in metallic dual purpose casks (DPCs) is implemented. The actual storage licenses for SNF and the HAW (which is stored in Gorleben and Rubenow) are limited to 40 years after closing the transport and storage cask. The first license to expire will be the storage license for the interim storage facility in Gorleben in the year 2034. In Germany the decision about a repository site will be taken in the year 2031. To enable the utilities to hand over the SNF and HAW to the repository regime while the current licenses are still valid would require a very fast process licensing and construction of the required facilities at this location. As a delay cannot be excluded, necessary preparations have to be taken on the side of the regulator. In the German Atomic Energy Act precautions have already been taken and consulting the German parliament (Bundestag) is mandatory if inevitable events would require an extension of the currently licensed interim storage period. The new structure with the Federal Office BfE, and two companies for final disposal 'Bundesgesellschaft fuer Endlagerung' (BGE mbH) and interim storage 'Bundesgesellschaft fuer Zwischenlagerung' (BGZmbH) was established in 2017. The Federal Office for Safety of Nuclear Waste Management (BfE) is responsible for regulation in the field of nuclear disposal since 30.07.2016. So the package design approval for DPCs and the storage licenses for SNF will be granted by this office. Furthermore, it is the regulator for the repository and will finally grant a license for the German deep geological repository. Site selection and construction and operation of the repository will be the responsibility of BGE. On January 1, 2018, the BGZ will take over all remaining interim storage facilities operating under a license due to article six of the Atomic Energy Act. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Neutron radiation shielding material polyethylene: consequences of gamma irradiation - 15062

High and ultra-high molecular weight polyethylenes are high performance materials, which are predestined for a wide range of applications due to characteristics like high density, low weight, good dimensional stability, high chemical resistance, and high hydrogen content. These polymers are used in two demanding areas: as implant material in medical technology (only ultra-high molecular weight polyethylenes) and as a component for neutron shielding purposes in casks for storage and transport of radioactive waste. In the medical field (joint replacements) as well as during neutron radiation shielding application, high and ultra-high molecular weight polyethylenes are exposed to gamma irradiation: in the first case requested as sterilization process and for surface-crosslinking and in the second case existing as a side effect of inserting the radioactive material in the cask. Given that polyethylene as shielding material has to withstand any type of degradation affecting safety relevant aspects to be applicable for long-term radiation shielding purposes for instance over a period of 40 years, the durability of the material is of special interest. High molecular weight polyethylene (HMW-PE; LUPOLEN 5261Z; Lyondell Basell) and ultra-high molecular weight polyethylene (UHMW-PE; GUR 4120; Ticona) were subjected to gamma radiation and afterwards thermally treated. The gamma doses used are in the range of 50 to 600 kGy and irradiation takes place at RT using a Co-60 source. The planned thermal treatment will take place at a temperature of 125 deg.C for periods of minimum 30 days and maximum of five years. With the applied conventional analytical techniques it is possible to detect structural changes of both types of polyethylene induced by gamma irradiation and certainly of thermal treatment. Through gamma irradiation melting temperature, crystallinity, and density, respectively increased. Furthermore both polyethylenes get oxidized and cross-linked. With regard to the special application as neutron radiation shielding material in casks for storage and transport of radioactive materials, the impact of irradiation lead to changes of material properties. A consolidated view indicates that the detected changes of the irradiated (U)HMW-PE are not safety relevant for long-term neutron radiation shielding purposes over a period of 40 years in Germany. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Strategy for Fuel Rod Receipt, Characterization, Sample Allocation for the Demonstration Sister Rods

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 the research and development activities related to storage, transportation, and disposal of used nuclear fuel and high-level radioactive waste. 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 (UNF) and wastes generated by existing and future nuclear fuel cycles. The UFDC Storage and Transportation staffs are responsible for addressing issues regarding the extended or long-term storage of UNF and its subsequent transportation. The near-term objectives of the Storage and Transportation task are to use a science-based approach to develop the technical bases to support the continued safe and secure storage of UNF for extended periods, subsequent retrieval, and transportation. While low burnup fuel [that characterized as having a burnup of less than 45 gigawatt days per metric tonne uranium (GWD/MTU)] has been stored for nearly three decades, the storage of high burnup used fuels is more recent. The DOE has funded a demonstration project to confirm the behavior of used high burnup fuel under prototypic conditions. The Electric Power Research Institute (EPRI) is leading a project team to develop and implement the Test Plan to collect this data from a UNF dry storage system containing high burnup fuel. The Draft Test Plan for the demonstration outlines the data to be collected; the high burnup fuel to be included; the technical data gaps the data will address; and the storage system design, procedures, and licensing necessary to implement the Test Plan. To provide data that is most relevant to high burnup fuel in dry storage, the design of the test storage system must closely mimic real conditions high burnup SNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage. To document the initial condition of the used fuel prior to emplacement in a storage system, “sister ” fuel rods will be harvested and sent to a national laboratory for characterization and archival purposes. This report supports the demonstration by describing how sister rods will be shipped and received at a national laboratory, and recommending basic nondestructive and destructive analyses to assure the fuel rods are adequately characterized for UFDC work. For this report, a hub-and-spoke model is proposed, with one location serving as the hub for fuel rod receipt and characterization. In this model, fuel and/or clad would be sent to other locations when capabilities at the hub were inadequate or nonexistent. This model has been proposed to reduce DOE-NE’s obligation for waste cleanup and decontamination of equipment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Investigation of Thermal-Hydraulic Effects of Dry Storage Canister Helium Backfill Loss Using the Horizontal Dry Cask Simulator

A previous investigation produced data sets that can be used to benchmark the codes and best practices presently used to determine cladding temperatures and induced cooling air flows in modern horizontal dry storage systems. The horizontal dry cask simulator (HDCS) was designed to generate this benchmark data and add to the existing knowledge base. The objective of the previous HDCS investigation was to capture the dominant physics of a commercial dry storage system in a well-characterized test apparatus for a wide range of operational parameters. The close coupling between the thermal response of the canister system and the resulting induced cooling air flow rate was of particular importance. The previous investigation explored these parameters using helium backfill at 100 kPa and 800 kPa pressure as well as air backfill with a series of simulated decay heats. The helium tests simulated a horizontal dry cask storage system at normal storage conditions with either atmospheric or elevated backfill pressure, while the air tests simulated horizontal storage canisters following a complete loss of helium backfill, in which case the helium would be replaced by air. The present HDCS investigation adds to the previous investigation by exploring steady-state conditions at various stages of the loss of helium backfill from a horizontal dry cask storage system. This is achieved by using helium/air blends as a backfill in the HDCS and running a series of tests using various simulated decay heats to explore the effects of relative helium/air molar concentration on the thermal response of a simulated horizontal dry cask storage system. A total of twenty tests were conducted where the HDCS achieved steady state for various assembly powers, representative of decay heat. The power levels tested were 0.50, 1.00, 2.50, and 5.00 kW. All tests were run at 100 kPa vessel pressure. The backfill gases used in these tests are given in this report as a function of mole fraction of helium (He), balanced by air: 1.0, 0.9, 0.5, 0.1, and 0.0 He. Steady-state conditions (where the steady-state start condition is defined as where the change in temperature with respect to time for the majority of HDCS components is less than or equal to 0.3 K/h) were achieved for all test cases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Storage, transportation and disposal system for used nuclear fuel assemblies

An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Storage, transportation and disposal system for used nuclear fuel assemblies

An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Concepts for Small-Scale Testing of Used Nuclear Fuel

This report documents a concept for a small-scale test involving between one and three Boiling Water Rector (BWR) high burnup (HBU) fuel assemblies. This test would be similar to the DOE funded High Burn-Up (HBU) Confirmatory Data Project to confirm the behavior of used high burn-up fuel under prototypic conditions, only on a smaller scale. The test concept proposed would collect data from fuel stored under prototypic dry storage conditions to mimic, as closely as possible, the conditions HBU UNF experiences during all stages of dry storage: loading, cask drying, inert gas backfilling, and transfer to an Independent Spent Fuel Storage Installation (ISFSI) for multi-year storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Review Facility Design Drawings M3FT-16LA040105011 - Safeguards and Security by Design for Used Fuel Extended Storage: 1.02.04.01.05 FT – 16LA04010501

This work package focuses on developing Best Practices for the design of security for, and domestic safeguarding (e.g. MC&A) of, a pilot-scale independent spent/used fuel storage facility consistent with conceptual design efforts in Nuclear Fuels Storage and Transportation (NFST) and Used Fuel Disposal (UFD) campaigns. This is a review of the basic design of the facility to determine a candidate list of accounting and control requirements that could be considered for safeguards and security purposes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Evaluation of the Frequency for Gas Sampling for the High Burnup Confirmatory Data Project

This report provides a technically based gas sampling frequency strategy for the High Burnup (HBU) Confirmatory Data Project. The evaluation of: 1) the types and magnitudes of gases that could be present in the project cask and, 2) the degradation mechanisms that could change gas compositions culminates in an adaptive gas sampling frequency strategy. This adaptive strategy is compared against the sampling frequency that has been developed based on operational considerations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

MP197HB; First licensed transportation package with canistered high burnup fuel after dry storage - 15580

AREVA TN has recently received the NRC 10 CFR Part 71 transportation license CoC 9302 Revision 7 for the MP197HB transportation package that allows transportation of canistered high burnup fuel assemblies. The Nuclear Horizontal Modular Storage (NUHOMS{sup R}) system canisters containing high burnup PWR and BWR fuel assemblies are now licensed for transportation as a payload in MP197HB after these canisters were in dry storage at the utility ISFSIs. Enhancements have been made to the MP197HB transportation package to improve thermal performance and also provide superior shielding capabilities. Design features of existing AREVA TN transportation packages such as radial fins, thermal shield, neutron resin and optimized gamma and neutron shielding were used to accommodate transportation of high burnup fuel assemblies with heat loads up to 32 kW. The transportation package is designed for exclusive use by rail, truck or marine transport. State-of-the-art analyses methodologies were used to demonstrate compliance with regulatory requirements of 10 CFR 71.55 for the high burnup fuel after dry storage at any ISFSI. Some of the issues addressed in the license application related to the transportation of high burnup fuel included potential degradation of fuel cladding, uncertainty associated with high burnup fuel cladding material properties, and larger radiation and thermal sources due to higher burnups. Bounding conservative assumptions were made where limited test data was available. AREVA TN's licensing approach did not rely on the cladding integrity of high burnup fuel cladding. Instead, the criticality, thermal and shielding analyses were carried out assuming high burnup fuel reconfiguration. With the approval of this transportation package for PWR and BWR canistered high burnup fuel from dry storage at ISFSIs, the utilities can now transport these canisters which are in dry storage at their sites. This is also a major step forward for the canistered fuel that is currently in dry storage at the shutdown plants because they can use this AREVA TN MP197HB transportation package to ship that fuel to any interim storage facility (ISF) or to a repository when they become available. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Demonstration of friction stir Welding (FSW) Technology for packaging of used nuclear fuel - 15025

With uncertainty surrounding the disposal of US Used Nuclear Fuel (UNF), the nuclear industry is making an effort to identify an interim storage solution. One of the alternatives is to simply manage UNF on-site until a permanent repository can be established - extended storage. In an effort to provide a technical basis for extended storage, the Department of Energy (DOE) performed a gap analysis identifying several issues that need to be addressed. One of the issues relates to the performance of the UNF dry storage container (DSC). DSC container welds (fabricated using conventional fusion welding processes) are sensitive to Stress Corrosion Cracking (SCC), and their extended-term performance is questionable. Fluor, the Pacific Northwest National Laboratory (PNNL) and the Savannah River National Laboratory (SRNL) believe that Friction Stir Welding (FSW) technology may be able to address the DSC issue. The Fluor team collaborated on a project to demonstrate feasibility of FSW for DSC packaging (closure-welding). This paper reports on a 2-year project in which FSW process conditions were established (for DSC container materials and dimensions) and used in the preparation of mechanical and corrosion specimens for testing and evaluation. In addition to FSW test specimens, fusion welded specimens were prepared and tested in like manner for comparative purposes. The results show excellent mechanical properties (meeting construction code requirements) and significant improvement in corrosion performance - over the fusion welded specimens. It is believed these findings will be of significant value to efforts underway to provide a technical basis for DSC extended storage performance. FSW technology appears to be an acceptable alternative or candidate process, capable of addressing the SCC sensitivity issues associated with fusion-welded DSC fabrication welds. In addition, as an alternative to on-site, extended storage of UNF, the DOE has proposed the development and construction of a Consolidated Interim Storage Facility (CISF). UNF would be accepted from the many, local sites for consolidation at one or more CISFs until a permanent repository could be established. A recent CISF study indicates that UNF currently packaged into DSCs will likely be re-packaged into containers suitable for permanent storage. The study also indicates that re-packaging is best performed under water in a pool. FSW technology has several specific advantages for this application, 1) because it utilizes machine-tool equipment, a welding station can easily perform a cutting operation, and 2) FSW has been successfully demonstrated capable of joining container materials under water. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

A Novel Mobile Examination and Remediation Facility for On-Site Remediation of Dry Storage Systems - 18389

Most of the spent nuclear fuel (SNF) in dry storage in the United States is safely and securely stored in dual-purpose (i.e., storage and transportation) canisters [1]. However, some of the SNF is stored in storage-only (i.e., non-transportable) packages, or packages that may become nonconforming for continued storage or transportation for a variety of reasons. Currently, the capability to remediate dry storage systems (DSSs) for SNF that are not transportable or that need to be repaired to maintain confinement boundary integrity during storage service does not exist without on-site access to a spent fuel pool. Hence, there is a potential future need for a mobile facility with remediation capabilities that can be used at multiple storage installations. These capabilities would permit needed repairs to an existing DSS to support continued storage or for the dry transfer of fuel to a new conveyance suitable for transport, if necessary. This is a technical paper that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in canisters to be an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. A novel design concept for a Mobile Examination and Remediation Facility (MERF) is proposed with functions and capabilities to remediate a DSS. The proposed MERF would be a multi-purpose facility, remotely operated and computer-controlled with semi-automated equipment to allow trained operators to safely perform remediation activities while keeping worker radiation dose as low as reasonably achievable. It would include a shielded rotating platform with support towers and would use commercial off- the-shelf examination and repair equipment. A slip ring arrangement underneath the rotating platform would allow for the power supply without impeding the continuous rotation and would simplify cable management. The support towers would be mounted on the rotating platform, and each tower would have vertical and horizontal positioning capability for axial and radial tool location. The rotating platform would be attached to the main MERF structure. Capabilities would include visual, surface, and volumetric examinations to fully characterize physical conditions on canister surface before and after a repair; repair techniques, e.g., laser beam welding, friction stir welding, or cold spray, to perform needed remediation activities; and associated equipment for surface preparation such as cutting and grinding equipment. The proposed design would be simple, modular, robust, economically viable, and easily transportable and would have a small footprint. The modular design would present several advantages including compatibility with a variety of different DSSs, on-site assembly, and transportability; it also would be built in interchangeable sections to facilitate decontamination. Further, shielding would be supplemented through interface with the existing transfer cask and storage overpack. Robustness and canister stability would be achieved using pneumatic holders and steel support structures designed with vibration absorbers to reduce impacts and/or a 'pendulum' effect during a seismic event. Thus, the proposed MERF would be designed to: (i) minimize handling of an affected DSS to the extent practicable, (ii) provide radiation shielding and minimize dose to operators and the public during remediation activities, and (iii) enable examination, repair, and repackaging. This paper describes the proposed MERF design concept and capabilities, main components, equipment, different modes of operation, and operating procedures. Structural and safety assessments would need to be performed before final design, licensing, and deployment. The MERF is evaluated with respect to regulatory impact (first of a kind, prior experience), operational impact (safety, security, staffing, ongoing operations, constraints), worker dose (low, medium, high), cost (low, medium, high), schedule (duration, time sequence), and technology readiness level. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Self-powered Wireless Through-wall Data Communication for Nuclear Environments

In the nuclear industry, many vital components, such as spent fuel storage canisters and nuclear reactor pressure vessels (RPV), are entirely enclosed by metal and surrounded by thick concrete walls that manage the potentially harmful radiation and prevent release to the environment. Due to the casks’ long storage, monitoring temperature, pressure, radiation, humidity, structural health, etc., within these enclosed vessels is crucial to ensure the fuel containment safety and security. In this paper, a self-powered wireless through-wall data communication system for nuclear environments was designed. The whole package includes: a radiation energy harvester with power management; ultrasound wireless communication using high-temperature piezoelectric transducers; electronics modules for harvesting, sensing, and data transmission; and radiation shielding for electronics and sensors. The package is able to harvest tens to hundreds of mW power from a nuclear canister environment directly and provides a path forward for continuous monitoring of the dry cask for 50 years.

Wu, Yongjia↗
DOE OSTI2017

Sensitivity and uncertainty analysis for a combined depletion and criticality calculation chain of a BWR FA

This paper examines the prerequisites and boundary conditions for extending burnup credit from pressurized water reactors to boiling water reactors with validated depletion codes by performing uncertainty and sensitivity analyses on a modern boiling water reactor fuel assembly design. We developed a computational chain consisting of an axial boiling water reactor axial irradiation model, followed by a series of depletion calculations and a three-dimensional criticality calculation. The irradiation model puts into relation the profiles of power, burnup and moderator density of the fuel assembly at 24 axial locations. In a simplified fashion it also includes the insertion of control blades. The depletion calculations of the fuel assembly were performed in axial slices with the 2D-depletion code HELIOS 1.12. The criticality calculations were done in 3D-geometry with KENO-Va from the SCALE 6.2.1 package, in which several fuel assemblies are placed into a generic transport cask. For the uncertainty and sensitivity analyses a Monte Carlo sampling based approach is used applying the GRS tool SUnCISTT. We performed several generic investigations testing the capabilities and the limitations of the whole computational chain. The impact of variations of the irradiation model on the resulting reactivity of the fuel assemblies inside a transport cask or storage pool was investigated using an uncertainty and sensitivity analysis on the whole computational chain. Geometrical uncertainties were also included in the evaluation and a parameter study on the use of control blades during irradiation was performed. The goal was to develop the procedures and tools to be able to obtain quantitative uncertainty data which can be used in the future to derive conservative irradiation profiles without unnecessary excess reactivity margins. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

IN-SERVICE INSPECTION OF EXTENDED DRY STORAGE OF SPENT NUCLEAR FUEL, PART I: CRAWLER TECHNOLOGY DEVELOPMENT

This paper describes development and demonstration of remote crawling systems to support periodic examinations of interim dry storage system (DSS) canisters for spent nuclear fuel in the USA. Specifically, this work relates to robotic crawler developments for “canister” based DSS systems which form the majority population of DSSs, in the USA, for interim storage of spent nuclear fuel. Consideration of potential degradation of the welded stainless-steel canister in these systems is required for continued usage in the period of extended operation (PEO) beyond their initial licensed or certified terms. Challenges associated with performing the periodic examinations are associated with physical access to the canister surface which is constrained due to narrow annulus spaces between the canister and the overpack, tortuous entry pathways, and high temperatures and radiation doses that can be damaging to materials and electronics. Motivations for performing periodic examinations and developing robotic crawlers for performing those examinations remotely will be presented and several activities to demonstrate robotic crawlers for different DSS systems are summarized.

nondestructive examination (NDE), Dry Cask Storage↗
DOE OSTI2017

Spent Nuclear Fuel Transportation: Rail Conveyance Transmissibility

Equipos Nucleares, S.A. (ENSA) and the US Department of Energy (DOE) are preparing a full scale Spent Nuclear Fuel (SNF) transportation test. This transportation testing will include road, rail, coastal, and trans-Atlantic shipments. The test campaign will use a full scale commercial transportation package and cradle. The package will be loaded with two instrumented fuel assemblies to measure strains at cladding locations and accelerations on the fuel assemblies, and thirty “dummy” assemblies. This testing is designed to closely match an actual SNF shipment. Accelerometers will also be used at various locations throughout the full conveyance system (i.e. rail car/truck, cask, transport cradle, and basket) to study the transmission of loads through the system and to provide validation for numerical models. Previous testing and 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/truck, transport cradle, cask, basket, and fuel) (Ref 1, 2). This paper describes preliminary models that were constructed to estimate load transmission during rail transport, from the bottom of the cradle to an individual fuel rod within the package. The modeling studies in this paper evaluate the system response to postulated shock pulses and random vibration loads. These models describe the transmissibility of the conveyance, and demonstrate how loads can be amplified or attenuated as they are transmitted through the structure. This is done by coupling the rail vehicle dynamics code NUCARS, to the general finite element modeling code ANSYS, and the explicit dynamics code LS-DYNA. Models such as the ones presented herein will be used during the test campaign to help analyze and evaluate the test data as it is collected.

Cladding, Spent Fuel and Radiactive Waste, Used Fu↗