Publications

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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.

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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.

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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.

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