Spent fuel transportation risk assessment: cask impact analyses

Abstract

The Nuclear Regulatory Commission (NRC) 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 impact 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. Finite element analyses were performed for impacts at speeds of 48, 97, 145 and 193 kilometres per hour into a rigid target. Impacts in end-on, side-on, and CG-over-corner orientations were analysed for each cask and impact speed. Calculations were performed to equate these impacts onto rigid targets with higher speed impacts onto the yielding targets that exist in the real world. These analyses indicated that a cask with an inner welded canister or a truck cask would not release radioactive material in any impact accident and that only very high-speed impacts onto hard rock targets could result in either release of material or significant degradation of shielding for rail casks without an inner canister. Impacts other than those onto flat unyielding targets were also considered. Analyses show that an impact that bypasses the impact limiters on the ends of the casks does not result in seal failure and neither does an impact by a locomotive also between the impact limiters.     

Analysis and design of spent fuel transport cask impact limiters

Abstract

Three Latin American countries which operate research reactors, Argentina, Brazil and Chile, have joined efforts to improve the capability in the management of spent fuel elements from the reactors operated in the region. As a step in this direction, a packaging for the transport of irradiated fuel from research reactors was designed by a tri-national team and a half scale model for materials test reactor fuel was constructed in Argentina and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions.

In this paper both the numerical modelling and mechanical tests to select adequate shock absorbers materials are presented. Results of these tasks are used to improve the cask design.     

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: routine transportation

Abstract

The RADTRAN model for calculating radiation doses is based on the well understood behaviour of ionising radiation. Absorption of ionising radiation depends on the energy and type of radiation and on the absorbing material. The casks that are used to transport spent nuclear fuel have walls that absorb most of the emitted ionising radiation and thereby shield the public and the workers. For routine transportation, RADTRAN models the cask as a sphere and assumes that the longest dimension of the trailer or railcar carrying the cask is the same as that of the cask. The dose rate in Sv/h at one metre from the cask is modelled as a virtual source at the centre of a sphere whose diameter is the longest dimension of the actual spent fuel cask. People who live along the cask’s route and the people in vehicles that share the route are exposed to external radiation from the cask. The dose to workers and the public from a cask during routine transportation depends on the time that the workers or public are exposed to the cask, the distance from the cask, and the cask’s external radiation. When the vehicle carrying the cask is travelling along the route, the faster the vehicle goes, the less exposure to anyone along the vehicle’s route. Therefore, an individual member of the public receives the largest dose from a moving vehicle when he or she is as close as possible to the vehicle, and the vehicle is travelling as slowly as possible. In the present analysis, these doses are in the range of four to seven nanosieverts. Collective doses along the route depend on the size of the exposed population. In this study, such doses were of the order of 0·1 person-millisieverts. The appropriate comparison between the collective dose from a shipment of spent fuel is not a comparison between the radiation dose from the shipment and zero dose, but between the background radiation dose in the presence and absence of a shipment, e.g. 8·810096 person-Sv if there is a shipment and 8·81000 person-Sv if there is no shipment.     

An integrated risk communication system for the transport of hazardous materials

Abstract

This paper describes the development and implementation of the prototype of an internet-based risk communication system for the transport of hazardous materials. The system was designed with the objectives of (1) incorporating functionality and features that are useful for meeting a variety of risk communication needs and (2) demonstrating a high degree of interaction among system components, enabling customisation to meet the specific transport risk communication requirements of the host organisation. To demonstrate 'proof of concept', the system is applied to two scenarios: building knowledge and awareness, focusing on how information can be entered, organised and disseminated to the public and other transport stakeholders, and emergency management, utilising the system for securely managing information in responding to a transport incident involving hazardous materials. The effectiveness of the system in these applications is subsequently discussed.     

Radioactive material transport risk management initiative: AREVA’s extensive approach to operations safety

Abstract

Since 2006, when AREVA mandated its Logistics Business Unit to handle ‘Transport Risk Management’, the Unit has executed or has, at least, ensured the proper management of shipments of radioactive material, which involve particular risks. This mission is fully complementary to the strict implementation of national and international regulations regarding the safety of radioactive material transport. Taking these regulations as a starting point, the AREVA Transport Risk Management Initiative develops general principles of risk management appropriate to operations that are sensitive by nature. By applying the Transport Risk Management Initiative to the shipment of radioactive materials, AREVA has widened the precautionary principle beyond the field of safety and radioprotection: accomplishing safe transport necessarily implies the identification and management of all risks inherent in these operations (safety, physical protection, media pressure, geopolitics, etc.). The guiding principles currently in use by AREVA and the organisation and resources that have enabled the concretisation of this ambition at the operational level are outlined: shipments executed each year by the AREVA Logistics Business Unit, as well as shipments subcontracted by the AREVA group to external shipping companies, fall within the scope of this initiative, and one of the stakes is to manage suppliers, often in an international environment.     

Advancing US public acceptance of spent fuel storage and transport: proposed outreach services for ionising radiation education support

Abstract

Expansion of commercial nuclear energy could be one of the future US sources for clean, safe, reliable and economic electricity. However, no federal policy has effectively achieved wide acceptance of nuclear energy, with such policies having fallen victim to the politics of public radiation fears from nuclear energy usage and from spent fuel storage and transport. Many experts have described the foundation of public fear as not so much nuclear technology, but the ionising radiation to which people fear they might be exposed, and this issue has been talked and written about, yet gone substantially unaddressed with respect to public education for more than three decades. In the USA, the Blue Ribbon Commission Final Report is just the latest of clear statements where such an educational need is firmly asserted. The lamentable fact is that no one has made that substantive and concerted effort to do anything about it. Indeed, the only effort seems to have been talk about ‘better communication’, with a focus on risk based communication. Any rejuvenation of public acceptance of commercial nuclear energy in the USA, including spent fuel storage and transport, can only be sustained using a different strategy from that of earlier decades. This paper highlights professional opinion on the radiation fear issue and why current industry efforts in risk based information for and communication with the public have not achieved the desired success. Education to expand the public’s understanding of comparative radiation sources and exposures while ameliorating concern about radiation from nuclear energy is the proposed alternative. In addition, here, the clear linkage between education supporting nuclear energy and facilitating necessary spent fuel storage and transport is unmistakable. The paper summarises a concept for outreach services for ionising radiation education support for application in the US, as well as key elements of such a process: its basis for success, its education content and potential implementation approaches. Comparative radiation education of the public can prove effective using current research, which has been effective in other industries. Additionally, while this discussion addresses the US situation, much of the content is likely applicable to many of the world’s nuclear energy producing countries.     

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.     

Modelling of polyurethane foam thermal degradation within annular region subject to fire conditions

Abstract

Nuclear materials are placed in shielded, stainless steel packaging for storage or transport. These drum type packages often employ a layer of foam, honeycomb, wood or cement that is sandwiched between thin metal shells to provide impact and thermal protection during hypothetical accidents, as those prescribed in the Code of Federal Regulations (10 CFR 71·73). The present work discusses the modelling of the thermal degradation of polyurethane (PU) foam within an annular region during an 800°C fire. Measurements and analysis by Hobbs and Lemmon [M. L. Hobbs and G. H. Lemmon: ‘Polyurethane foam response to fire in practical geometries’, Polym. Degrad. Stab., 2004, 84, 183–197.] indicate that at elevated temperatures, PU foam exhibits a two-stage, endothermic degradation. The first stage produces a degraded solid and a combustible gas; the second stage reaction consumes the degraded solid and produces another combustible gas. As a result, during a prolonged fire, a gas filled void develops beside the outer metal shell and grows inward toward the inner shell and the containment vessel. As a result of the radial symmetry in the drum geometry, a one-dimensional finite difference model is constructed for the annular foam region. Heat flux is applied to the inner surface to model the decay heat of the containment vessel contents. Thermal radiation and convection boundary conditions with a specified environmental temperature are applied to the outer surface. The material and reaction rate properties determined by Hobbs and Lemmon are applied to the foam. The annular region temperature and composition are determined as functions of radius and time after the environmental conditions are changed from room temperature to those of an 800°C fire. The effects of surface to surface radiation between the package’s outer skin and the undegraded foam and the reaction rate reduction due to material damage during the reaction are evaluated for fires lasting 20 h. The peak package liner temperature caused by a 30 min fire is predicted to be 129°C, well below the short term limit for containment vessel seal (377°C).     

Numerical Simulation of Dynamic Deformation Of Spent Fuel Transport Packages for Nuclear Power Plants in Accidental Mechanical Impacts

Abstract

Numerical simulation results and their analysis are presented for dynamic deformations of the TUK-117 package, intended for the transport of spent nuclear fuel from nuclear power plants, subjected to accidental 9 m drops onto an unyielding surface at different angles. This paper focuses on the analysis of the deformation behaviour of container shock limiters. It is demonstrated that maximum loading affects the package during a side drop. For a side drop, the maximum strain levels are determined for the different construction elements, including the cask's body and the bolts securing the sealing lid. Dynamic simulation of the behaviour of the construction elements was carried out using the LS-DYNA code, version 970.     

The Application of Fracture Mechanics to The Safety Assessment of Transport Casks For Radioactive Materials

Abstract

BAM is the responsible authority in Germany for the assessment of the mechanical and thermal design safety of packages for the transport of radioactive materials. The assessment has to cover the proof of brittle fracture safety for package components made of potentially brittle materials. This paper gives a survey of the regulatory and technical requirements for such an assessment according to BAM's new 'Guidelines for the application of ductile cast iron for transport and storage casks for radioactive materials'. Based on these guidelines, higher stresses than before will be permissible, but it is necessary to put more effort into the safety assessment procedure. The fundamentals of such a proof using the methods of fracture mechanics are presented. The recommended procedure takes into account the guidelines of the IAEA's advisory material which are based on the prevention of crack initiation. Examples of BAM's research and safety assessment practices are given. Recommendations for further developments towards package designs with higher acceptable stress levels will conclude the paper.