Sandia National Laboratories (SNL) has completed a comparative evaluation of three design assessment approaches for a 2-liter (2L) capacity containment vessel (CV) of a novel plutonium air transport (PAT) package designed to survive the hypothetical accident condition (HAC) test sequence defined in Title 10 of the United States (US) Code of Federal Regulations (CFR) Part 71.74(a), which includes a 129 meter per second (m/s) impact of the package into an essentially unyielding target. CVs for hazardous materials transportation packages certified in the US are typically designed per the requirements defined in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PVC) Section III Division 3 Subsection WB “Class TC Transportation Containments.” For accident conditions, the level D service limits and analysis approaches specified in paragraph WB-3224 are applicable. Data derived from finite element analyses of the 129 m/s impact of the 2L-PAT package were utilized to assess the adequacy of the CV design. Three different CV assessment approaches were investigated and compared, one based on stress intensity limits defined in subparagraph WB-3224.2 for plastic analyses (the stress-based approach), a second based on strain limits defined in subparagraph WB-3224.3, subarticle WB-3700, and Section III Nonmandatory Appendix FF for the alternate strain-based acceptance criteria approach (the strain-based approach), and a third based on failure strain limits derived from a ductile fracture model with dependencies on the stress and strain state of the material, and their histories (the Xue-Wierzbicki (X-W) failure-integral-based approach). This paper gives a brief overview of the 2L-PAT package design, describes the finite element model used to determine stresses and strains in the CV generated by the 129 m/s impact HAC, summarizes the three assessment approaches investigated, discusses the analyses that were performed and the results of those analyses, and provides a comparison between the outcomes of the three assessment approaches.
A series of extensively instrumented tests was performed on the Structural Evaluation Test Unit in the early 1990s. The purpose of these tests was to determine the response of a minimally designed cask to impacts that were more severe than the design basis impact. This test series provides an excellent opportunity for benchmarking explicit dynamic finite element analysis programs for behaviors that may be experienced by casks during regulatory and extra-regulatory impact events. This report provides the results of the four tests that were conducted. It is meant to go along with a companion report that defines the benchmark problem and gives the locations for the instrumentation and inspection points.
A series of extensively instrumented tests was performed on the Structural Evaluation Test Unit in the early 1990s. The purpose of these tests was to determine the response of a minimally designed cask to impacts that were more severe than the design basis impact. This test series provides an excellent opportunity for benchmarking explicit dynamic finite element analysis programs for behaviors that may be experienced by casks during regulatory and extra-regulatory impact events. This report provides the parameters of the test unit, the locations of instrumentation, the locations of inspection points, and the parameters of the four tests that were conducted. A companion report provides the results of the tests.
Sandia National Laboratories (SNL) was contracted by the United States Department of Energy Environmental Management (DOE-EM), Los Alamos Field Office to perform mechanical and thermal scoping calculations as part of a study seeking to understand the ignitability risk of the Remediated Nitrate Salts (RNS) waste drums during transportation from the Waste Control Specialists (WCS) facility to Waste Isolation Pilot Plant (WIPP) and permanent disposal of the waste at WIPP. The scoping thermal simulations described in this report pertain to thermal calculations performed with a packaging system consisting of one Standard Waste Box (SWB) loaded with drums placed inside a Standard Large Box 2 (SLB2). During transportation, the SLB2 is inside Transuranic Package Transporter Model III (TRUPACT-III), which provides the third layer of the packaging. Once at the WIPP, it is assumed the SLB2 is extracted from the TRUPACT-III and maintained above ground, and then subsequently placed underground for permanent disposal. In these proposed configurations, the space between the SLB2 and the SWB is always filled by a layer of insulation consisting of air-filled glass microbubbles except for the bottom which rests directly on the SLB2. The thermal scoping calculations described in this report specifically address whether the introduction of external heat inputs, combined with the contributions from the internally generated radiolytic decay heat and chemical reactions, lead to an unstable thermal state during the time of its movement and placement in the permanent disposal location. The external heat inputs are of two forms: 1) ambient thermal irradiation (e.g., solar and ambient storage/disposal temperatures) and 2) accident-induced fire. Three scoping calculation scenarios were derived as representative, conservative scenarios: 1A) TRUPACT-III transient transportation, 1B) SLB2 48-hour outdoor storage with solar radiation, and 2) fully-engulfing fire during SLB2 handling or emplacement following a steady-state analysis in a 38 °C environment. All the simulated scenarios are conservative relative to the operational conditions expected for handling the waste package during transportation and placement in the WIPP underground disposal unit. The predictions obtained from simulating the three exposure scenarios revealed that adding the SLB2 and the air-filled glass microbubbles to the transport and storage/disposal configurations provides additional thermal protection of the drums beyond what the SWB provides alone, both during long-term above ground insolation and underground during a fire accident. Under the current transportation/storage/disposal concepts, the degree of protection provided by the packaging concept is sufficient to prevent the waste from being ignitable. The simulation results demonstrate that there is adequate margin to safely transport and place the RNS waste from WCS to the WIPP under the current operational concept.
The Spent Fuel Waste Disposition (SFWD) program is planning to conduct a full-scale seismic shake table test on the dry storage systems of spent nuclear fuel (SNF) to close the gap related to seismic loads on fuel assemblies in dry storage systems. This test will allow for quantifying the strains and accelerations on surrogate fuel assembly hardware and cladding during earthquakes of different magnitudes and frequency content. Full-scale testing is needed because a dry storage system is a complex and highly nonlinear system making it hard to predict (model) the responses to seismic excitations. The non-linearity arises from the multiple spatial gaps in the system – between fuel rods and the basket, between the basket and dry storage canister, between the dry storage canister and the storage cask (overpack), and ventilation gaps. The non-linearities pose significant limitations on the value of tests with scaled systems.
The PRO-X program is actively supporting the design of nuclear systems by developing a framework to both optimize the fuel cycle infrastructure for advanced reactors (ARs) and minimize the potential for production of weapons-usable nuclear material. Three study topics are currently being investigated by Sandia National Laboratories (SNL) with support from Argonne National Laboratories (ANL). This multi-lab collaboration is focused on three study topics which may offer proliferation resistance opportunities or advantages in the nuclear fuel cycle. These topics are: 1) Transportation Global Landscape, 2) Transportation Avoidability, and 3) Parallel Modular Systems vs Single Large System (Crosscutting Activity).
This report documents the updated seismic shake table test plan. The report describes the shake table inputs (ground motions), test hardware, shake table facility, friction experiment, and proposed instrumentation.
This report is a preliminary test plan of the seismic shake table test. The final report will be developed when all decisions regarding the test hardware, instrumentation, and shake table inputs are made. A new revision of this report will be issued in spring of 2022. The preliminary test plan documents the free-field ground motions that will be used as inputs to the shake table, the test hardware, and instrumentation. It also describes the facility at which the test will take place in late summer of 2022.
The 30 cm drop is the remaining NRC normal conditions of transport (NCT) regulatory requirement (10 CFR 71.71) for which there are no data on the response of spent fuel. While obtaining data on the spent fuel is not a direct requirement, it allows for quantifying the risk of fuel breakage resulting from a cask drop from a height of 30 cm or less. Because a full-scale cask and impact limiters are very expensive, 3 consecutive drop tests were conducted to obtain strains on a full-scale surrogate 17x17 PWR assembly. The first step was a 30 cm drop of a 1/3 scale cask loaded with dummy assemblies. The second step was a 30 cm drop test of a full-scale dummy assembly. The third step was a 30 cm drop of a full-scale surrogate assembly. The results of this final test are presented in this paper. The test was conducted in May 2020. The acceleration pulses on the surrogate assembly were in good agreement with the expected pulses derived from steps 1 and 2. This confirmed that during the 30 cm drop the surrogate assembly experienced the same conditions as it would have if it had been dropped in a full-scale cask with impact limiters. The surrogate assembly was instrumented with 27 strain gauges. Pressure paper was inserted between the rods within the two long and two short spacer grid spans in order to register the pressure in case of rod-to-rod contact. The maximum observed peak strain on the surrogate assembly was 1,724 microstrain at the bottom end of the assembly. The pressure paper sheets from the two short spans were blank. The pressure paper sheets from the two long spans, except a few middle ones, showed marks indicating rod-to-rod contact. The maximum estimated contact pressure was 4,100 psi. The longitudinal bending stress corresponding to the maximum observed strain value (calculated from the stress-strain curve for low burnup cladding) was 22,230 psi. Both values are significantly below the yield strength of the cladding. The major conclusion is that the fuel rods will maintain their integrity following a 30 cm drop inside of a transportation cask.
The 30 cm drop is the remaining NRC normal conditions of transport (NCT) regulatory requirement (10 CFR 71.71) for which there are no data on the actual surrogate fuel. While obtaining data on the actual fuel is not a direct requirement, it provides definitive information which aids in quantifying the risk of fuel breakage resulting from a cask drop from a height of 30 cm or less. The 30 cm drop test with the full-scale surrogate assembly conducted in May 2020 was the last step needed for quantifying the strains on the surrogate assembly rods under NCT. The full-scale surrogate assembly used in the 2020 30 cm drop test was built using a new 17x17 Pressurized Water Reactor (PWR) Westinghouse skeleton filled with the copper rods and 3 zircaloy rods from the full-scale surrogate assembly used in the Multi-Modal Transportation Test (MMTT). Felt pads were attached to the surrogate assembly bottom prior to the 30 cm drop to adequately represent the effects of the impact limiters and the cask. Note that felt "programming material" has been used extensively in past drop tests and is known to be a good material for programming a desired shock pulse. The felt pad configuration was determined during a previous series of tests reported in. The acceleration pulses observed on the surrogate assembly during the test were in good agreement with the expected pulses. This confirmed that during the 30 cm drop the surrogate assembly experienced the same conditions as it would if it was dropped in the cask with the impact limiters.
Currently, spent nuclear fuel (SNF) is stored in onsite independent spent fuel storage facilities (ISFSIs), which is a dry storage facility, at 55 nuclear power plant sites. The majority of SNF in dry storage is in welded metal canisters (2,917 canisters at the end of 2019). The canisters are loaded for storage in storage overpacks (vertical casks or horizontal storage modules) and placed on outdoor concrete pads. Because the SNF will be stored at ISFSIs for an extended period of time, there is growing concern with regards to the behavior of the SNF within these dry storage systems during earthquakes. To address these concerns, the SFWST program is considering conducting an earthquake shaker table test. The goal of this test is to determine the strains and accelerations on fuel assembly hardware and cladding during earthquakes of different magnitudes to better quantify the potential damage an earthquake could inflict on spent nuclear fuel rods. The seismic integrity of the storage system has been addressed in the past by the US Nuclear Regulatory Commission and is not the focus of this potential test. Instead the DOE would benefit from knowing the condition of the fuel cladding from storage, transportation, to disposal so that it can ascertain repository performance for the fuel and packaging in its final state. A seismic event is part of the possible loading events that the fuel could experience in its lifetime. This report proposes several earthquake shaker table tests with different degrees of complexity. Alternative 1 was defined in the FY20 work scope. Alternatives 2 and 3 were recently developed to take advantage of the NUHOMS 32PTH dry storage canister that may be available in FY21 for this test at a minimum cost to the project. The selection of the alternative(s) will depend on the available budget and the SFWST program priorities for the near future.
The goal of this transportation analysis (TA) is to update the 2008 TA in order to evaluate the impacts associated with the transportation of transuranic (TRU) waste from waste generator sites to the Waste Isolation Pilot Plant (WIPP) facility and from waste generator sites to the Idaho National Laboratory (INL).
This report will describe the one test conducted during phase III of the Pipe Overpack Container (POC) test campaign, present preliminary results from these tests, and discuss implications for the Criticality Control Overpack (CCO). The goal of this test was to see if aerosol surrogate material inside the Criticality Control Container (CCC) gets released when the drum lid of the CCO comes off during a thirty-minute long, fully-engulfing, fire test. As expected from POC tests conducted in Phase I and II of this test campaign, the CCO drum lid is ejected about one minute after the drum is exposed to fully-engulfing flames. The remaining pressure inside the drum is high enough to eject the top plywood dunnage a considerable distance from the drum. Subsequently, most of the bottom plywood dunnage supporting the CCC burns off during and after the fire. High pressure buildup inside the CCC and inside two primary containers holding the surrogate powder also results in damage to the filter media of the CCC and the filter-house, thread attachment of the primary canisters. No discernable release of surrogate powder material was detected from the two primary containers when pre- and post-test average mass were compared. However, when the average masses are corrected to account for possible uncertainties in mass measurements, error overlap does not preclude the possibility that some surrogate powder mass may have been lost from these primary canisters. Still, post-test conditions of the secondary canisters enclosing these two primary canisters suggest it is very unlikely this mass loss would have escaped into the CCC.
The data from the multi-modal transportation test conducted in 2017 demonstrated that the inputs from the shock events during all transport modes (truck, rail, and ship) were amplified from the cask to the spent commercial nuclear fuel surrogate assemblies. These data do not support common assumption that the cask content experiences the same accelerations as the cask itself. This was one of the motivations for conducting 30 cm drop tests. The goal of the 30 cm drop test is to measure accelerations and strains on the surrogate spent nuclear fuel assembly and to determine whether the fuel rods can maintain their integrity inside a transportation cask when dropped from a height of 30 cm. The 30 cm drop is the remaining NRC normal conditions of transportation regulatory requirement (10 CFR 71.71) for which there are no data on the actual surrogate fuel. Because the full-scale cask and impact limiters were not available (and their cost was prohibitive), it was proposed to achieve this goal by conducting three separate tests. This report describes the first two tests — the 30 cm drop test of the 1/3 scale cask (conducted in December 2018) and the 30 cm drop of the full-scale dummy assembly (conducted in June 2019). The dummy assembly represents the mass of a real spent nuclear fuel assembly. The third test (to be conducted in the spring of 2020) will be the 30 cm drop of the full-scale surrogate assembly. The surrogate assembly represents a real full-scale assembly in physical, material, and mechanical characteristics, as well as in mass.
The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) for disposal. In 1996 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provided during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire. This test resulted in one of the POCs generating sufficient internal pressure to pop off its drum lid and expose the top of the pipe container (PC) to the fire environment. The initial contents of the POCs were inert materials that would not generate large internal pressure within the PC if heated. However, POCs are now being used to store combustible Transuranic (TRU) waste at Department of Energy (DOE) sites. At the request of DOE's Office of Environmental Management (EM) and National Nuclear Security Administration (NNSA), SNL started conducting a new series of fire tests in 2015 to examine whether PCs with combustibles would reach a temperature that could result in: (1) decomposition of inner contents and (2) subsequent generation of sufficient gas to cause the PC to over-pressurize and release its inner contents. In 2016, Phase II of the tests showed that POCs tested in a pool fire failed within 3 minutes of ignition with the POC lid ejecting. These POC lids were fitted with a NUCFIL-019DS filter and revealed that this specific filter did not relieve sufficient pressure to prevent lid ejection. In the Fall of 2017, Phase II-A was conducted to expose POCs to a 30-minute pool fire with similar configurations to those tested in Phase II, except that the POC lids were fitted with an UltraTech (UT) 9424S filter instead. That specific filter was chosen because of its design to help relieve internal pressure during the fire and thus prevent lid ejection. In Phase II-A, however, setups of two POCs stacked upon one another were never tested, which led to this phase of tests, Phase II-B. This report will describe the various tests conducted in Phase II-B, present results from these tests, and implications for the POCs based on the test results will be discussed.
This report describes the Shaker Table Test conducted on September 12, 2018, at the Dynamic Certification Laboratories (DCL) in Sparks, Nevada. This report satisfies Milestone M3SF-19SN010202021 Shaker Table Test, Sandia National Laboratories (SNL) Work Package (Parent WBS # 1.08.01.02.02; Work Package #SF-19SN01020202). The Shaker Table Test is related to the Multi-Modal Transportation Test (MMTT) conducted in 2017. During the MMTT, accelerations and strains were measured on the transportation platform, ENsa UNiversal (ENUN) 32P dual-purpose rail cask, cradle, basket, and three surrogate 17x17 pressurized water reactor (PWR) assemblies (one from SNL, one from Spain and one from Korea).
The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) for disposal. In 1996 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provided during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire, resulting in one of the 7A drum overpacks generating sufficient internal pressure to pop off its lid and expose the top of the pipe container (PC) to the fire environment. The initial contents of the POCs were inert materials, which would not generate large internal pressure within the PC if heated. However, POCs are now being used to store combustible Transuranic (TRU) waste at Department of Energy (DOE) sites. At the request of DOE's Office of Environmental Management (EM) and National Nuclear Security Administration (NNSA), SNL started conducting a new series of fire tests in 2015 to examine whether PCs with combustibles would reach a temperature that would result in (1) decomposition of inner contents and (2) subsequent generation of sufficient gas to cause the PC to overpressurize and release its inner content. In 2016, Phase II tests showed that POCs tested in a pool fire failed within 3 minutes of ignition with the POC lid ejecting. These POC lids were fitted with an all-metal (NUCFIL019DS) filter and revealed that this specific filter did not relieve sufficient pressure to prevent lid ejection. For the test phase discussed in this report, Phase II-A, the POCs are exposed to a 30-minute pool fire, with similar configurations to those tested in Phase II, except that the POC lids are fitted with a hybrid metal-polyethylene (UT9424S) filter instead. This report will: describe the various tests conducted in Phase II-A, present results from these tests, and discuss implications for the POCs based on the test results.
The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) for disposal. In 1996 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provided during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire, resulting in one of the 7A drum overpacks generating sufficient internal pressure to pop off its lid and expose the top of the pipe container (PC) to the fire environment. The initial contents of the POCs were inert materials, which would not generate large internal pressure within the PC if heated. POCs are now being used to store combustible TRU waste at Department of Energy (DOE) sites. At the request of DOE’s Office of Environmental Management (EM) and National Nuclear Security Administration (NNSA), starting in 2015 SNL conducted a series of fire tests to examine whether PCs with combustibles would reach a temperature that would result in (1) decomposition of inner contents and (2) subsequent generation of sufficient gas to cause the PC to over-pressurize and release its inner content. Tests conducted during 2015 and 2016 were done in three phases. The goal of the first phase was to see if the PC would reach high enough temperatures to decompose typical combustible materials inside the PC. The goal of the second test phase was to determine under what heating loads (i.e., incident heat fluxes) the 7A drum lid pops off from the POC drum. The goal of the third phase was to see if surrogate aerosol gets released from the PC when the drum lid is off. This report will describe the various tests conducted in phase I, II, and III, present preliminary results from these tests, and discuss implications for the POCs.
An ENUN 32P cask supplied by Equipos Nucleares S.A. (ENSA) was transported 9,600 miles by road, sea, and rail in 2017 in order to collect shock and vibration data on the cask system and surrogate spent fuel assemblies within the cask. The task of examining 101,857 ASCII data files – 6.002 terabytes of data (this includes binary and ASCII files) – has begun. Some results of preliminary analyses are presented in this paper. A total of seventy-seven accelerometers and strain gauges were attached by Sandia National Laboratories (SNL) to three surrogate spent fuel assemblies, the cask basket, the cask body, the transport cradle, and the transport platforms. The assemblies were provided by SNL, Empresa Nacional de Residuos Radiactivos, S.A. (ENRESA), and a collaboration of Korean institutions. The cask system was first subjected to cask handling operations at the ENSA facility. The cask was then transported by heavy-haul truck in northern Spain and shipped from Spain to Belgium and subsequently to Baltimore on two roll-on/roll-off ships. From Baltimore, the cask was transported by rail using a 12- axle railcar to the American Association of Railroads’ Transportation Technology Center, Inc. (TTCI) near Pueblo, Colorado where a series of special rail tests were performed. Data were continuously collected during this entire sequence of multi-modal transportation events. (We did not collect data on the transfer between modes of transportation.) Of particular interest – indeed the original motivation for these tests – are the strains measured on the zirconium-alloy tubes in the assemblies. The strains for each of the transport modes are compared to the yield strength of irradiated Zircaloy to illustrate the margin against rod failure during normal conditions of transport. The accelerometer data provides essential comparisons of the accelerations on the different components of the cask system exhibiting both amplification and attenuation of the accelerations at the transport platforms through the cradle and cask and up to the interior of the cask. These data are essential for modeling cask systems. This paper concentrates on analyses of the testing of the cask on a 12-axle railcar at TTCI.
An ENUN 32P cask supplied by Equipos Nucleares S.A. (ENSA) was transported 9,600 miles by road, sea, and rail in 2017 in order to collect shock and vibration data on the cask system and surrogate spent fuel assemblies within the cask. The task of examining 101,857 ASCII data files – 6.002 terabytes of data (this includes binary and ASCII files) – has begun. Some results of preliminary analyses are presented in this paper. A total of seventy-seven accelerometers and strain gauges were attached by Sandia National Laboratories (SNL) to three surrogate spent fuel assemblies, the cask basket, the cask body, the transport cradle, and the transport platforms. The assemblies were provided by SNL, Empresa Nacional de Residuos Radiactivos, S.A. (ENRESA), and a collaboration of Korean institutions. The cask system was first subjected to cask handling operations at the ENSA facility. The cask was then transported by heavy-haul truck in northern Spain and shipped from Spain to Belgium and subsequently to Baltimore on two roll-on/roll-off ships. From Baltimore, the cask was transported by rail using a 12- axle railcar to the American Association of Railroads’ Transportation Technology Center, Inc. (TTCI) near Pueblo, Colorado where a series of special rail tests were performed. Data were continuously collected during this entire sequence of multi-modal transportation events. (We did not collect data on the transfer between modes of transportation.) Of particular interest – indeed the original motivation for these tests – are the strains measured on the zirconium-alloy tubes in the assemblies. The strains for each of the transport modes are compared to the yield strength of irradiated Zircaloy to illustrate the margin against rod failure during normal conditions of transport. The accelerometer data provides essential comparisons of the accelerations on the different components of the cask system exhibiting both amplification and attenuation of the accelerations at the transport platforms through the cradle and cask and up to the interior of the cask. These data are essential for modeling cask systems. This paper concentrates on analyses of the testing of the cask on a 12-axle railcar at TTCI.
This report describes tests conducted using a full-size rail cask, the ENSA ENUN 32P, involving handling of the cask and transport of the cask via truck, ships, and rail. The purpose of the tests was to measure strains and accelerations on surrogate pressurized water reactor fuel rods when the fuel assemblies were subjected to Normal Conditions of Transport within the rail cask. In addition, accelerations were measured on the transport platform, the cask cradle, the cask, and the basket within the cask holding the assemblies. These tests were an international collaboration that included Equipos Nucleares S.A., Sandia National Laboratories, Pacific Northwest National Laboratory, Coordinadora Internacional de Cargas S.A., the Transportation Technology Center, Inc., the Korea Radioactive Waste Agency, and the Korea Atomic Energy Research Institute. All test results in this report are PRELIMINARY – complete analyses of test data will be completed and reported in FY18. However, preliminarily: The strains were exceedingly low on the surrogate fuel rods during the rail-cask tests for all the transport and handling modes. The test results provide a compelling technical basis for the safe transport of spent fuel.
The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) for disposal. In 1996 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provided during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire, resulting in one of the 7A drum overpacks generating sufficient internal pressure to pop off its lid and expose the top of the pipe container (PC) to the fire environment. The initial contents of the POCs were inert materials, which would not generate large internal pressure within the PC if heated. However, POCs are now being used to store combustible TRU waste at Department of Energy (DOE) sites. At the request of DOE’s Office of Environmental Management (EM) and National Nuclear Security Administration (NNSA), starting in 2015 SNL conducted a new series of fire tests to examine whether PCs with combustibles would reach a temperature that would result in (1) decomposition of inner contents and (2) subsequent generation of sufficient gas to cause the PC to over-pressurize and release its inner content. Tests conducted during 2015 and 2016, and described herein, were done in two phases. The goal of the first phase was to see if the PC would reach high enough temperatures to decompose typical combustible materials inside the PC. The goal of the second test phase was to determine under what heating loads (i.e., incident heat fluxes) the 7A drum lid pops off from the POC drum. This report will describe the various tests conducted in phase I and II, present preliminary results from these tests, and discuss implications for the POCs.
Most of the regulatory agencies world-wide require that containers used for the transportation of natural UF6 and depleted UF6 must survive a fully-engulfing fire environment for 30 minutes as described in 10CFR71 and in TS-R-1. The primary objective of this project is to examine the thermo-mechanical performance of 48Y transportation cylinders when exposed to the regulatory hypothetical fire environment without the thermal protection that is currently used for shipments in those countries where required. Several studies have been performed in which UF6 cylinders have been analyzed to determine if the thermal protection currently used on UF6 cylinders of type 48Y is necessary for transport. However, none of them could clearly confirm neither the survival nor the failure of the 48Y cylinder when exposed to the regulatory fire environment without the additional thermal protection. A consortium of five companies that move UF6 is interested in determining if 48Y cylinders can be shipped without the thermal protection that is currently used. Sandia National Laboratories has outlined a comprehensive testing and analysis project to determine if these shipping cylinders are capable of withstanding the regulatory thermal environment without additional thermal protection. Sandia-developed coupled physics codes will be used for the analyses that are planned. A series of destructive and non-destructive tests will be performed to acquire the necessary material and behavior information to benchmark the models and to answer the question about the ability of these containers to survive the fire environment. Both the testing and the analysis phases of this project will consider the state of UF6 under thermal and pressure loads as well as the weakening of the steel container due to the thermal load. Experiments with UF6 are also planned to collect temperature- and pressure-dependent thermophysical properties of this material.
Sandia National Laboratories has constructed an unyielding target at the end of its 2000-foot rocket sled track. This target is made up of approximately 5 million pounds of concrete, an embedded steel load spreading structure, and a steel armor plate face that varies from 10 inches thick at the center to 4 inches thick at the left and right edges. The target/track combination will allow horizontal impacts at regulatory speeds of very large objects, such as a full-scale rail cask, or high-speed impacts of smaller packages. The load-spreading mechanism in the target is based upon the proven design that has been in use for over 20 years at Sandia's aerial cable facility. That target, with a weight of 2 million pounds, has successfully withstood impact forces of up to 25 million pounds. It is expected that the new target will be capable of withstanding impact forces of more than 70 million pounds. During construction various instrumentation was placed in the target so that the response of the target during severe impacts can be monitored. This paper will discuss the construction of the target and provide insights on the testing capabilities at the sled track with this new target.
The ASME Task Group on Computational Mechanics for Explicit Dynamics is investigating the types of finite element models needed to accurately solve various problems that occur frequently in cask design. One type of problem is the 1-meter impact onto a puncture spike. The work described in this paper considers this impact for a relatively thin-walled shell, represented as a flat plate. The effects of mesh refinement, friction coefficient, material models, and finite element code will be discussed. The actual punch, as defined in the transport regulations, is 15 cm in diameter with a corner radius of no more than 6 mm. The punch used in the initial part of this study has the same diameter, but has a corner radius of 25 mm. This more rounded punch was used to allow convergence of the solution with a coarser mesh. A future task will be to investigate the effect of having a punch with a smaller corner radius. The 25-cm thick type 304 stainless steel plate that represents the cask wall is 1 meter in diameter and has added mass on the edge to represent the remainder of the cask. The amount of added mass to use was calculated using Nelm's equation, an empirically derived relationship between weight, wall thickness, and ultimate strength that prevents punch through. The outer edge of the plate is restrained so that it can only move in the direction parallel to the axis of the punch. Results that are compared include the deflection at the edge of the plate, the deflection at the center of the plate, the plastic strains at radius r=50 cm and r=100 cm , and qualitatively, the distribution of plastic strains. The strains of interest are those on the surface of the plate, not the integration point strains. Because cask designers are using analyses of this type to determine if shell will puncture, a failure theory, including the effect of the tri-axial nature of the stress state, is also discussed. The results of this study will help to determine what constitutes an adequate finite element model for analyzing the puncture hypothetical accident.