Safety of ocean shipments of plutonium between Europe and Japan

Abstract

Sandia National Laboratories (SNL) has conducted an extensive study of emergency response planning applicable to sea transport of plutonium for the Japan Nuclear Cycle Development Institute (JNC). This work covered four separate areas to better define the accident environment for long range sea transport of nuclear materials. A probabilistic safety analysis evaluated technical issues for the transport of plutonium between Europe and Japan. An engine room fire aboard a purpose built ship was used to analyse the vulnerability of plutonium packaging designed to International Atomic Energy Agency (IAEA) standards. A comprehensive corrosion study estimated the time required for sea water to breach a containment boundary in submerged generic plutonium packaging. A survey of worldwide commercial recovery capabilities provided a compilation of information on the capabilities of salvaging high value cargo from sunken ships. This paper addresses salvage modes from harbour depths to the deepest ocean trenches. Previous studies (J. L. Sprung et al., SNL reports SAND98-1171/1 and SAND98-1171/2, May 1998) included a probabilistic risk assessment of the overall safety, source term evaluations and finite element structural dynamics calculations to determine the effects of ship to ship collisions on nuclear material containers and the effects of ship fires on transport packaging as determined by actual fire experiments conducted on board a test ship. The previous studies, together with this work, form a comprehensive technical basis that encompasses the overall safety of sea transport of plutonium between Europe and Japan. Based on these technical analyses, transport of nuclear materials by sea in Type B packaging, approved in accordance with US Nuclear Regulatory Commission (NRC) and IAEA regulations, and carried in purpose built ships with adequate surveillance, has a very high degree of safety for the failure modes studied. Non-purpose built ships do not have the redundancy in safety features provided by the newer purpose built ships. However, SNL studies on non-purpose built ships have shown accident environments to be within NRC and IAEA regulatory assumptions for Type B packaging. These studies were carried out for both structural ship to ship collisions and engine room fires by analysis (for the collisions) and direct experimentation and analysis (for the fires). Thus, land transport mode regulations are applicable for sea transport accident conditions.     

Mechanical safety analysis for high burn-up spent fuel assemblies under accident transport conditions

Abstract

Transport packages for spent fuel have to meet the requirements concerning containment, shielding and criticality as specified in the International Atomic Energy Agency regulations for different transport conditions. Physical state of spent fuel and fuel rod cladding as well as geometric configuration of fuel assemblies are, among others, important inputs for the evaluation of correspondent package capabilities under these conditions. The kind, accuracy and completeness of such information depend upon purpose of the specific problem. In this paper, the mechanical behaviour of spent fuel assemblies under accident conditions of transport will be analysed with regard to assumptions to be used in the criticality safety analysis. In particular the potential rearrangement of the fissile content within the package cavity, including the amount of the fuel released from broken rods has to be properly considered in these assumptions. In view of the complexity of interactions between the fuel rods of each fuel assembly among themselves as well as between fuel assemblies, basket, and cask body or cask lid, the exact mechanical analysis of such phenomena under drop test conditions is nearly impossible. The application of sophisticated numerical models requires extensive experimental data for model verification, which are in general not available. The gaps in information concerning the material properties of cladding and pellets, especially for the high burn-up fuel, make the analysis more complicated additionally. In this context a simplified analytical methodology for conservative estimation of fuel rod failures and spent fuel release is described. This methodology is based on experiences of BAM acting as the responsible German authority within safety assessment of packages for transport of spent fuel.     

Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

Abstract

The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal, off-normal and accident conditions. The environmental temperature is assumed to be 27°C under the normal condition. The off-normal condition has an environmental temperature of 40°C. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The accident condition is defined as a 100% blockage of air inlet ducts. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of the ventilation system have been carried out for the determination of the optimum duct size and shape. The finite-volume computational fluid dynamics code FLUENT was used for the thermal analysis. From the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal, off-normal and accident conditions.     

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.     

CONSTOR® V/TC drop tests: pre-test analysis using the finite-element method

Abstract

CONSTOR^® is a family of steel–CONSTORIT–steel sandwich cask designs that have been developed with special consideration for an economical and effective method of manufacture by using conventional mechanical engineering technologies and common materials. The CONSTOR^® concept fulfils both the internationally valid IAEA criteria for transport and the requirements for long-term intermediate storage in the USA and various European countries. A full-scale prototype test cask, CONSTOR^® V/TC, of the latest CONSTOR^® design has been developed, with a heat removal capacity of up to 32 kW. A comprehensive drop testing programme consisting of five 9 m drops onto a flat unyielding target and seven 1 m drops onto a punch is to be carried out by BAM at the test facilities in Horstwalde during Autumn 2004, with the first 9 m side drop to be carried out during PATRAM 2004. The drop tests will form part of the application for a transport licence in both Germany and the USA. Extensive pre-test calculations have been performed using finite-element methods. The objectives of the analyses are as follows: (1) As an intermediate step in demonstrating the performance of the package in fulfilling the requirements of 10 CFR 71 and the IAEA transport regulations. (2) To justify the selection of drop tests. (3) To predict the performance of the V/TC in the drop tests. (4) To estimate the strain and acceleration–time history at measuring points to aid the setting up of the instrumentation. (5) To develop an analysis model that can be used in future safety analyses for transport and storage licence applications to confidently demonstrate the performance of the package. This paper will: present an overview of the analyses; discuss the methodology of the analysis, including the design and make-up of the models taking into account the behaviour of the package, the requirements of the licensing regimes and the present and future purposes of the model; discuss the modelling techniques used; present key results from the analyses; and discuss the behaviour of the package.     

Transport of large nuclear power plant components: experiences in mechanical design assessment

Abstract

In the course of decommissioning of power plants in Germany large nuclear components (steam generator, reactor pressure vessel) must be transported over public traffic routes to interim storage facilities, where they are dismantled or stored temporarily. Since it concerns surface contaminated objects or low specific activity materials, a safety evaluation considering the IAEA transport regulations mainly for industrial packages (type IP-2) is necessary. For these types of industrial packages the requirements from normal transport conditions are to be covered for the mechanical proof. For example, a free drop of the package from a defined height, in dependence of its mass, onto an unyielding target, and a stacking test are required. Since physical drop tests are impossible generally due to the singularity of such 'packages', a calculation has to be performed, preferably by a complex numerical analysis. The assessment of the loads takes place on the basis of local stress distributions, also with consideration of radiation induced brittleness of the material and with consideration of recent scientific investigation results. Large nuclear components have typically been transported in an unpackaged manner, so that the external shell of the component provides the packaging wall. The investigation must consider the entire component including all penetration areas such as manholes or nozzles. According to the present IAEA regulations the drop position is to be examined, which causes the maximum damage to the package. In the case of a transport under special arrangement a drop only in an attitude representing the usual handling position (administratively controlled) is necessary. If dose rate values of the package are higher than maximum allowable values for a public transport, then it is necessary that additional shielding construction units are attached to the large component.     

Letter to the Editor

Abstract

Transport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies (109 assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing) were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this packaging design features such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of the transport control centre, communication training, and accompanying of the radiation monitoring expert. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule.     

Transport of Fresh MOX Fuel Assemblies for the MONJU Initial Core

Abstract

Transport of fresh MOX fuel assemblies for the prototype FBR MONJU initial core started in July 1992 and ended in March 1994. As many as 205 fresh MOX fuel assemblies (109 assemblies for an inner core, 91 assemblies for an outer core and 5 assemblies for testing) were transported in nine transport missions. The packaging for fuel assemblies, which has shielding and shock absorbing material inside, meets IAEA regulatory requirements for Type B(U) packaging including hypothetical accident conditions such as the 9 m drop test, fire test, etc. Moreover, this packaging design features such advanced technologies as high performance neutron shielding material and an automatic hold-down mechanism for the fuel assemblies. Every effort was made to carry out safe transport in conjunction with the cooperation of every competent organisation. This effort includes establishment of, the transport control centre, communication training, and accompanying the radiation monitoring expert. No transport accident occurred during the transport and all the transport missions were successfully completed on schedule.     

Computational and experimental validation of protective features of transport packages for long-term storage of spent nuclear fuel from nuclear power plants

Abstract

The purpose of this paper is to describe the work of our Institute (RFNC-VNIITF) on the development of a unique long transport container using new manufacturing technologies. The container is designed for transport and long-term storage of spent fuel elements more than 10 m long. The disposal of such elements usually implies their cutting. This requires considerable expense in building special cutting plants. Another way is to use long protective shells to contain and store the entire fuel cell. Development of such containers has been made possible by the use of rolled-strip technology ensuring maximum strength of containers. The paper contains certain results of calculations and experimental research justifying and confirming the protective properties of the long package. To confirm reliability with minimum material consumption an available package model, manufactured by new technology, is used. This package was enhanced in the flange area to simulate strictly a flange joint of a full-scale transport package. Calculations and experimental work were performed according to IAEA regulations. During investigations the full-scale cask was considered as both a model and a prototype as it simulated completely the bottom and the flange joint 'lid-cask body' (i.e. being a model) while being shorter and thinner than the cylindrical part of the body (i.e. being a prototype). Here it is called simply a prototype. The complete simulation was not set as an objective. At the same time the cask body loading for tests simulating an air crash was done with regard to the full-scale design. The paper contains results obtained analytically on the basis of experimental data to justify the properties of a full-scale package without experimentally testing it: therefore only some photographs and experimental results for the prototype are presented here. The title of the paper indicates that the analytical study was conducted with use of numerical constants obtained experimentally. It is impossible to present all the information, including experimental research, here because of the large scope of the work performed and the limited space available.     

Results of full scale CONSTOR® V/TC prototype 9 m horizontal drop test

Abstract

In the context of the research on the mechanical safety of packages for radioactive material, full scale drop tests with spent fuel and high activity waste transport and storage casks have been performed by the Federal Institute for Materials Research and Testing (BAM). The research reflects national and international interest in acquiring comparative knowledge of full and reduced scale model drop tests as well as in finite element calculations. This paper presents the experimental, analytical and first numerical results of the full scale drop test with the full scale CONSTOR^® V/TC prototype, manufactured by GNS, Gesellschaft für Nuklear-Service mbH, Germany. The prototype was tested by BAM in a 9 m horizontal drop test onto the unyielding target of the BAM drop test facility in Horstwalde, Germany.     

Preliminary Work Related to the Transport of Low and Intermediate Level Radioactive Waste in the Republic of Croatia

Abstract

In the Republic of Croatia, there is a project for a repository for low level (LL) and intermediate level (IL) radioactive waste disposal. Among many preliminary proceedings related to the proposed construction of the repository (such as site selection, repository project design, public acceptance, and the like), are the problems related to the transport of LL/IL radioactive waste from the place of its generaiion or storage to the location of the final disposal site. In this phase of the preliminary works— prior to site selection and further working out of project documentation for the facility, it has only been possible to commence with some study papers related to the problem. During 1992/1993 the first version of the generic study related to transport was worked out with the aim of preparing for a more detailed study, and technical and investment support for further working out of the project. The study is firstly a literature abstrac,, with an analysis of the transport problem for Croatia and recommendations for a transport system, type of transport, transport equipment and dynamics for the requirements of the Repubiic of Croatia.     

Ten years of Romanian experience on risk assessment owing to radioactive materials transport in Romania

Abstract

The transport of radioactive materials is a very important problem considering the potential risks and radiological consequences in carrying out the present activity. Based on the International Atomic Energy Agency (IAEA)'s Safety Standard TS-R-1 (1996 edition, as amended 2003), Romanian National Nuclear Regulatory Body – Romanian National Commission for Nuclear Activities Control (CNCAN) was adopted and implemented by act no. 374/October 2001, the safety regulations for the transport of radioactive materials in Romania under the title 'Fundamental regulations for a safe transport of radioactive materials, in Romania'. The present paper will present the main sources of radioactive materials in Romania, their transport routes with a particular interest paid to the radioactive wastes. Hypothetical scenarios for specific problems related to the identification and evaluation of the risks and potential radiological consequences associated with the transport of radioactive materials in Romania, for all these situations: routine transport (incident free) and possible accidents.     

WAGR Decommissioning: The Transport of WAGR Heat Exchangers to Drigg Low Level Waste Repository

Abstract

The Windscale Advanced Gas-Cooled Reactor (WAGR) operated between 1963 and 1981 and was used as a prototype for experimental research and development related to the Advanced Gas-Cooled Reactor programme. WAGR is now the UK's demonstration project for power reactor decommissioning. A major operation in the decommissioning of the WAGR was the removal of the four heat exchangers (REs) from their concrete bioshields inside the secondary containment sphere and transporting them to Drigg low level waste repository. The issues involved with the transport of the four heat exchangers are described, covering details of the heat exchangers themselves as a transport package. Preparation of the route along public roads is described along with details of the transport vehicles and transport operations.     

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.     

Facilities and Systems for Transport of Encapsulated Spent Fuel to a Final Repository in Sweden

Abstract

In 2002 two communities in Sweden, Östhammar and Oskarhamn, accepted that SKB, the Swedish Nuclear Fuel and Waste Management Company, could continue the work of site investigations for a final repository for spent fuel. This was a big and important step towards the completion of the last link in the chain for a complete system for handling and finalstorage of spent fuel in Sweden. In 2008 the site investigations should be ready and one of the two sites will be chosen for the final repository in Sweden. The repositoryis scheduled for start of operation in 2017. This paper describes how facilities and transport systems could be designed, building on the experience of the current system in Sweden which has been in operation since 1985.     

Use of impact analysis in licensing industrial container for transporting new fuel

Abstract

The finite element (FE) method is a powerful tool for the simulation of mechanical and thermal behaviour of structures. In recent years, the explicit FE method has increasingly been used in the development of transport packages and as part of approval applications to demonstrate the performance of packages. Testing and analysis are the two methods specified in the IAEA Regulations for the Safe Transport of Radioactive Material for demonstrating the structural and thermal performance of a transport package against the requirements of the Transport Regulations. The roles of testing and analysis, and the relative prominence of the two, may vary between Competent Authorities in different countries. This can range from analysis being regarded as the primary mode of demonstration with testing as confirmatory, to testing being the primary mode of demonstration supplemented by analysis. This paper describes the use of the non-linear FE code LS-DYNA in the licensing of a new container for the transport of new nuclear fuel. The package was classified as an Industrial Package (Fissile) in accordance with the IAEA Regulations, and hence it was necessary, among other things, to demonstrate that criticality criteria were satisfied under postulated impact conditions. Physical drop tests were carried out and the results are compared with LS-DYNA computer calculations using the same FE models developed to support the design of the new container. The analyses and tests clearly demonstrate the novel use of polyurethane foam as the container main energy absorber. The FE predictions are compared for accelerations, bolt loadings and global deformations of the container. In general good correlation was obtained between predictions and tests and the differences, which did occur, particularly for accelerations, are discussed and reconciled. The paper concludes that explicit analysis codes are now so reliable for container impact calculations that minimal test work should be pursued basically for key confirmatory impact scenarios.     

Study of the Accident Environment During Sea Transport of Nuclear Material: Analysis of An Engine-Room Fire on a Purpose-Built Ship

Abstract

The programme goal was to show that the IAEA safe transport regulations adequately cover the thermal effects of an engine-room fire on plutonium transport packages stowed aboard a purpose-built ship. The packages are stored in transport containers located in a cargo hold of the ship. For this study, it was assumed that the packages in No 5 Hold, adjacent to an engine-room, could be subject to heating due to a fire in the engine-room. The No 5 hold and the engine-room are separated by a water-filled bulkhead. This study addressed the heat transfer from an engine-room fire that could heat and evaporate water out of the water-filled bulkhead and the resulting temperature conditions around the packages and inside the packages near their elastomeric seals.     

COG-OP-30B: A New COGEMA Overpack for Type 30B Cylinders

Abstract

Facing the difficulties encountered in the United States of America in obtaining a durable certificate of approval for the existing overpacks, COGEMA desired to free itself from the constraints imposed by these packagings for the transport of enriched uranium hexafluoride. Transnucl6aire therefore initiated the design of a new Type B(U)F overpack, on behalf of COGEMA, for use with the 30B cylinders containing uranium hexafluoride enriched up to 6% in ²³⁵U, obtained from natural or reprocessed uranium. The external aspect of the new overpack is quite standard (two cylindrical halves closed with ten toggles) and it keeps the same stowage system so that it can be used with the methods of transport which exist already. The main features of COGEMA's overpack are a reinforced internal structure to protect the 30B cylinder's skirt from bending and better absorption of the drop energy to limit the acceleration under impact. The drop tests and a fire test were performed successfully in February and March 1997, the French Certificate of Approval was obtained at the end of 1998, the Swedish validation by mid-1999 and the Canadian validation at the beginning of 2000.     

TRANSPORT OF MOX FUEL

Abstract

The regulatory framework which governs the transport of MOX fuel is set out, including packages, transport modes and security requirements. Technical requirements for the packages are reviewed and BNFL's experience in plutonium and MOX fuel transport is described. The safety of such operations and the public perception of safety are described and the question of gaining public acceptance for MOX fuel transport is addressed. The paper concludes by emphasising the need for proactive programmes to improve the public acceptance of these operations.     

Safe transport of radioactive material in Brazil

Abstract

The objectives of this article are to briefly present an updated review of the regulatory framework and activities related to the transport of radioactive material in Brazil, to provide an analysis of the appraisal service performed by the International Atomic Energy Agency (IAEA) in 2002 and to identify questions that require action plans from the Brazilian Nuclear Energy Commission (CNEN), including those actions which will involve neighbouring countries regulatory authorities.