Douglas James Ammerman

Numerical Analyses of Locomotive Impacts on a Spent Fuel Truck Cask and Trailer

During the transportation of spent nuclear fuel by truck, the possibility exists that a train could run into the spent fuel cask at a grade crossing. Sandia National Laboratories has conducted a numerical study to assess the possibility of cask breach or material release in the event of a high-speed, broadside locomotive collision. A numerical approach has the advantage over conducting a physical test as was done in the 1970s [1] in that varying parameters can be examined. For example, one of the criticisms of the 1970s test was the height of the cask. In the test, the centerline of the cask was above the main frame-rails of the locomotive. In this study the position of the cask with respect to the locomotive was varied. The response of the cask and trailer in different collision scenarios was modeled numerically with LS-DYNA [2]. The simulations were performed as a collaborative endeavor between Sandia National Laboratories (SNL), Applied Research Associates, Inc. (ARA) and Foster-Miller, Inc (FMI). ARA developed the GA-4 Spent Fuel Cask and Cask Transporter models described in this report. These models were then combined with two existing FMI heavy freight locomotive finite element models to create the overall simulation scenarios. The modeling effort, results, and conclusions are presented in this paper.Copyright

COMPARISON OF UF6 PACKAGES UNDER CERTIFICATION TEST AND ACCIDENT CONDITIONS.

Abstract Currently there are three packages approved by the NRC for US domestic shipments of fissile quantities of UF6: NCI-21PF-1, UX-30, and ESP30X. For approval by the NRC, packages must be subjected to a sequence of physical tests to simulate transportation accident conditions as described in 10 CFR part 71. The primary objective of this project was to compare conditions experienced during these tests to conditions potentially encountered in actual accidents and to estimate the probabilities of such accidents. Comparison of the effects of actual accident conditions to 10 CFR part 71 tests was achieved by means of computer modelling of structural effects on the packages due to impacts with actual surfaces, and thermal effects resulting from tests and other fire scenarios. In addition, the likelihood of encountering bodies of water during transport over representative truck routes was assessed. Modelled effects and their associated probabilities, accident rates, and other characteristics gathered from representative routes were combined with existing event tree data to derive generalized probabilities of encountering accident conditions comparable to or exceeding the 10 CFR part 71 test conditions. This analysis suggests that the regulatory conditions are unlikely to be exceeded in real accidents.

Pipe Overpack Container Fire Testing: Phase II-A

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.

Spent fuel transportation risk assessment: cask fire analyses

Abstract The US Nuclear Regulatory Commission has recently completed an updated Spent Fuel Transportation Risk Assessment, NUREG-2125. This assessment considered the response of three certified casks to a range of fire accidents in order to determine whether or not they would lose their ability to contain the spent fuel or maintain effective shielding. The casks consisted of a lead shielded rail cask that can be transported either with or without an inner welded canister, an all steel rail cask that is transported with an inner welded canister, and a DU shielded truck cask that is transported with directly loaded fuel. For the two rail casks, large pool fires that were concentric (fully engulfing), offset from the casks by 3 m, and offset from the cask by 18 m were analysed using the computational fluid dynamics CAFE-3D fire modelling code coupled with the finite element analysis PATRAN-Thermal heat transfer code. All of the fires were assumed to last for 3 h. In addition to these extraregulatory fires, the regulatory 30 min fire was analysed using both the regulatory uniform 800°C boundary condition and the more realistic CAFE-3D fire modelling code. For the truck cask, only the engulfing fire case was analysed using a 1 h fire duration. In all of the fire analyses, the seal region of the cask stayed below the failure temperature; therefore, there would be no release of radioactive material. In addition, the temperature of the fuel rods stayed below their burst rupture temperature, providing another barrier to release. For the lead shielded cask, very severe fires cause some of the lead to melt. There is no leak path for this molten lead to exit the shield region, but its expansion during the melting and subsequent contraction due to solidification during cool down results in a reduction in gamma shielding effectiveness.

Spent fuel transportation risk assessment: overview

Abstract The US Nuclear Regulatory Commission (NRC) is responsible for issuing regulations for the packaging of spent fuel (and other large quantities of radioactive material) for transport that provide for public health and safety during transport [Title 10 of the Code of Federal Regulations (10 CFR) Part 71, ‘Packaging and transportation of radioactive waste’, dated 26 January 2004]. In September 1977, the NRC published NUREG 0170, ‘Final environmental statement on the transportation of radioactive material by air and other modes’, which assessed the adequacy of those regulations to provide safety assurance. In that assessment, the measure of safety was the risk of radiation doses to the public under routine and accident transport conditions, and the risk was found to be acceptable. Since that time, there have been two affirmations of this conclusion for spent nuclear fuel (SNF) transportation, each using improved tools and information that supported the earlier studies. This report presents the results of a fourth investigation into the safety of SNF transportation. The risks associated with SNF transportation come from the radiation that the spent fuel gives off, which is attenuated, but not eliminate, by the transportation casks shielding, and the possibility of the release of some quantity of radioactive material during a severe accident. This investigation shows that the risk from the radiation emitted from the casks is a small fraction of naturally occurring background radiation and the risk from accidental release of radioactive material is several orders of magnitude less. Because there have been only minor changes to the radioactive material transportation regulations between NUREG 0170 and this risk assessment, the calculated dose due to the external radiation from the cask under routine transport conditions is similar to what was found in earlier studies. The improved analysis tools and techniques, improved data availability, and a reduction in the number of conservative assumptions has made the estimate of accident risk from the release of radioactive material in this study approximately five orders of magnitude less than what was estimated in NUREG 0170. The results demonstrate that NRC regulations continue to provide adequate protection of public health and safety during the transportation of SNF.

SPENT FUEL TRANSPORTATION RISK ASSESSMENT: TRANSPORTATION ACCIDENT ANALYSIS.

Abstract The US Nuclear Regulatory Commission (NRC) has recently completed an updated Spent Fuel Transportation Risk Assessment, NUREG-2125. This assessment considered four types of accidents that could interfere with routine transportation of spent nuclear fuel: those in which the spent fuel cask is not affected, those in which there is loss of lead gamma shielding, those in which radioactive material is released and those that could result in a criticality event. The probability of a particular type of accident is the product of the probability that the vehicle carrying the spent fuel cask will be in an accident and the conditional probability that the accident will be of a certain type. An accident in which the spent fuel cask is not damaged or affected at all is the most probable: 99·95% of vehicle accidents are less severe than the regulatory hypothetical accident, and most accidents that are more severe than this still do not lead to loss of shielding or release, which occur in fewer than one accident in a billion. If a lead shielded cask is involved in one of these impacts, the lead shield can slump, and a small section of the spent fuel in the cask will be shielded only by the steel shells. The resulting external doses are significant but would result in neither acute illness nor death. The collective dose risks are vanishingly small. Consequences and risks of an accidental release of radioactive material are similar, since only very small amounts of material would be released, and only through damaged cask seals. The study also examined the probabilities and risks associated with several possible fire scenarios previously analysed by the NRC, and showed that even such events do not result in significant risks. Inclusion of such events increases the estimated risk by only a small fraction. Another accident type that is of potential concern is one that leads to a criticality event. This study has shown that the combination of factors necessary to produce such an event is so unlikely that the event is not credible.

SPENT FUEL TRANSPORTATION RISK ASSESSMENT: CONCLUSIONS.

Abstract The US Nuclear Regulatory Commission (NRC) has recently completed an updated Spent Fuel Transportation Risk Assessment, NUREG-2125. The study reached the following findings. First, the collective dose risks from routine transportation are vanishingly small. These doses are about four to five orders of magnitude less than collective background radiation doses. Second, the routes selected for this study adequately represent the routes for spent nuclear fuel transport, and there was relatively little variation in the risks per kilometre over these routes. Third, radioactive material would not be released in an accident if the fuel is contained in an inner welded canister inside the cask. Fourth, only rail casks without inner welded canisters would release radioactive material, and only then in exceptionally severe accidents. Fifth, if there were an accident during a spent fuel shipment, there is less than one in a billion chance the accident would result in a release of radioactive material. Sixth, if there were a release of radioactive material in a spent fuel shipment accident, the dose to the maximally exposed individual would be <2 Sv (200 rem) and would not cause an acute fatality. Seventh, the collective dose risks for the two types of extraregulatory accidents (accidents involving a release of radioactive material and loss of lead shielding) are negligible compared to the risk from a no release, no loss of shielding accident. Eight, the risk of loss of shielding from a fire is negligible. Ninth, none of the fire accidents investigated in this study resulted in a release of radioactive material. Based on these findings, this study reconfirms that radiological impacts from spent fuel transportation conducted in compliance with NRC regulations are low. In fact, this study’s radiological impact estimates are generally less than the already low estimates reported in earlier studies. Accordingly, with respect to spent fuel transportation, this study reconfirms the previous NRC conclusion that the regulations for transportation of radioactive material are adequate to protect the public against unreasonable risk.

Strain-Based Acceptance Criteria for Inelastic Analysis

Modern finite element codes used in the design of nuclear material transportation and storage casks can readily calculate the response of the packages beyond the elastic regime. These packages are designed to protect workers, the public, and the environment from the harmful effects of the transported radioactive material following a sequence of hypothetical accident conditions. Hypothetical accidents considered for transport packages include a 9-meter free drop onto an essentially unyielding target and a 1-meter free fall onto a 30-cm diameter puncture spike. For storage casks, accident conditions can include drops, tip-over, and aircraft impact. All of these accident events are energy-limited rather than load-limited, as is typically the case for boilers and pressure vessels. Therefore, it makes sense to have analysis acceptance criteria that are more closely related to absorbed energy than to applied load. Strain-based acceptance criteria are the best way to meet this objective. As cask vendors’ ability to perform non-linear impact analysis has improved, the need for a code-based method to interpret the results of this type of analysis has increased. The ASME Section III Working Group on Design of Division 3 Containments is working with Section III Working Group Design Methodology to develop strain-based acceptance criteria to use within the ASME Code for energy-limited events. This paper will briefly discuss the efforts within the ASME, detail the advantages of using strain-based criteria, discuss the problem areas associated with establishing strain-based criteria, and provide insights into inelastic analyses as applied to radioactive material transportation and storage casks in general. The views expressed represent those of the authors and not necessarily those of their respective organizations or the ASME.Copyright

Response of a Dry Storage Spent Fuel Cask to Hypothetical Loading

This paper addresses the analytical methods used to determine the response of a dry storage spent fuel cask to hypothetical loading. Because of the sensitive nature of the topic under discussion, the response of the cask is described in qualitative terms, and the paper is intentionally vague on the parameters and results. This research was sponsored by the U.S. Nuclear Regulatory Commission (NRC) Spent Fuel Project Office. The work was performed under contract from the Sandia National Laboratory (SNL), Transportation Risk and Packing organization. The analytical effort was performed at the Oak Ridge National Laboratory (ORNL) facilities with loading specified by SNL.Copyright