BAM Production Control Programme for Containers for Transport and Storage of Nuclear Materials

Abstract

In the process of testing spent fuel casks, BAM is gaining a lot of relevant data regarding the quality level of ductile cast iron (DCI). The basic parameters governing the material behaviour of ferritic and ferritic pearlitic DCI are dicussed and the development of container quality over recent years is summarised. The high quality level of German DCI containers is outlined. The effect of microstructure, sample size and loading rate on the fracture toughness of DCI is discussed in the second part of the paper.     

Design Assessment for Transport and Storage Casks

Abstract

The design assessment concerning the mechanical behaviour of transport and storage casks for radioactive material to fulfil nuclear safety criteria has to be based on two essential considerations: (1) Effective analysis of the stress–strain state of the cask components under both normal operational and test conditions including hypothetical accident scenarios with suitable accepted methods. (2) Economic estimation of the required properties and the structural state of the cask components with sufficient exactness. In an overview of the codes which are available at GNS/GNB for cask impact strength analyses (ANSYS, ADINA, VDI Codes), procedures and aspects of benchmarking and validation of calculation codes are described. The results of experimental full size cask drop test programs (CASTOR, POLLUX) and corresponding pre-test calculational analyses show the suitability of the codes used. The influence of dynamic effects on the mechanical properties of material (ductile cast iron, wood) has been investigated experimentally. By consideration of these dynamic values in strength analyses of casks at impact a good agreement between experimental and calculational results has been achieved.     

Development of Sealed Type B Transport Containers for the Transport of Intermediate Level Radioactive Waste

Abstract

UK Nirex Ltd have developed three possible concepts of sealed transport containers for the safe transport of immobilised intermediate level radioactive waste. Computer based finite element impact and thermal analysis has been carried out on each concept and compliance with both the IAEA regulatory requirements and specified Nirex design aims has been demonstrated. One single concept will be selected at a later date following further development and confirmatory testing to provide a fleet of transport containers.     

Application of IAEA TECDOC 717 to the Assessment of Brittle Fracture in Transport Containers with Plastic Flow Shock Absorbers

Abstract

UK Nirex is developing re-usable shielded transport containers (RSTCs) in a range of shielding thicknesses (from 70 nun to 285 nun) to transport immobilised intermediate level radioactive waste (ILW) to a deep repository. The RSTCs are being designed to meet the requirements of the IAEA Transport Regulations for Type B packages, including the requirement to maintain shielding and containment following a drop of 9 m onto an unyielding surface. The RSTCs are essentially monolithic in construction and the heaviest version weighs up to 65 tonnes when loaded with contents. They rely principally on plastic flow of the material of construction to absorb the high energies involved in impact events. Specific features of the designs, such as the solid metal comer shock absorbers and side ribs have been optimised for this purpose. Nirex has investigated the feasibility of manufacturing the RSTCs from ductile cast iron (DCI) or cast steel instead of from forgings, since this would bring advantages of reduced manufacturing time and costs. In this paper the methodology set out in IAEA-TECDOC-717 is applied to the Nirex RSTC, including the application of elastic plastic fracture mechanics methods.     

The Second Technical Seminar on Brittle Fracture Safety Assessment of Ductile Cast Iron Containers for Shipping and Storage of Radioactive Materials

Abstract

A facility was built in 1997 to enable drop tests to be carried out on samples of ?KODA casks in accordance with IAEA standards. Such drop tests have been instrumented and measured using hardware and software which is described further in the paper. The measurements have been validated by specific crash software. The whole measurement process was certified by the Czech Institute for Accreditation (Cesky institut pro akreditaci). Both the drop test facility and the measurement technology have the status of an accredited test shop (test room).     

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.     

Computerised Modelling of Radiolysis and Corrosion Effects in Transport Containers for Fuel Spent Nuclear

Abstract

The radiolysis of water and/or gases within transport containers for spent nuclear fuel may result in the generation of hydrogen and oxygen gases and also the enhanced corrosion of the materials in contact with the water. These effects are important, particularly when the fuel container is also used for storage post-transport prior to reprocessing or disposal. The behaviour of a range of radiolytic systems has been studied. Plant behaviour has been simulated in numerous laboratory experiments: plant and experimental results have been linked by a computerised model describing the radiolysis mechanism and predicting the quantities and production rate of gaseous and corrosive species. This allows prediction of plant performance over a long time scale. The model is based on a well-accepted radiolysis mechanism supplemented with specific measurements made at the Harwell laboratory. Model capabilities include inert atmospheres, materials corrosion, variations in water and gas volumes or aqueous chemistry. The model has been applied to design stage radiolysis assessments of transport containers; information from operating plant has been interpreted to advise on design improvement, e.g. diminution of gas production using easily corroded scavengers to remove oxygen. Radiolysis in gas filled dry storage containers for spent nuclear fuel has been studied; corrosive product production (e.g. nitric acid), which is important for fuel cladding integrity has been assessed. The development and use of this computerised model is described with a current summary.     

Mechanical Properties of Ductile Cast Iron and Cast Steel for Intermediate Level Waste Transport Containers

Abstract

UK Nirex Ltd is developing Type B re-usable shielded transport containers (RSTCs) in a range of shielding thicknesses to transport intermediate level radioactive waste (ILW) to a deep repository. The designs are of an essentially monolithic construction and rely principally on the plastic flow of their material to absorb the energies involved in impact events. Nirex has investigated the feasibility of manufacturing the RSTCs from ductile cast iron (DCI) or cast steel instead of from forgings, since this would bring advantages of reduced manufacturing time and costs. However, cast materials are perceived to lack toughness and ductility and it is necessary to show that sufficient fracture toughness can be obtained to preclude brittle failure modes, particularly at low temperatures. The mechanical testing carried out as part of that programme is described. It shows how the measured properties have been used to demonstrate avoidance of brittle fracture and provide input to computer modelling of the drop tests.     

秦山三期重水堆核电站废过滤器滤芯转运与收贮

简述秦山三期重水堆核电站废过滤器滤芯的产生系统及产生量,废滤芯的转运的准备、实施,收贮入库等实施操作流程及其资料的整理。     

The Design of New Concept Containers for the Transport of U Powders and Enriched UF6

Abstract

With the imminent introduction of the IAEA STI and the problems associated with existing 30″ Hex cylinder protective secondary packages (PSPs), BNFL made the decision to design and have ownership of a new PSP which overcomes shortcomings in other packages in order to ensure continued deliveries to their customers. The design ideas and the approach to the UK competent authority in order to obtain transport approval are outlined.     

The Development of a Transportable Storage Cask

Abstract

There are a number of different technologies for implementing interim storage at reactor sites, as shown in Table 1. It is generally accepted that, if possible, expanding the capacity of existing fuel pools through the installation of compact racks and the use of fuel rod consolidation are the most economical ftrst steps. Once these have been carried out, other alternatives must be employed if further capacity expansion is required. It is not the purpose of this paper to discuss the relative economics of these alternatives, since under speciftc constraints and conditions each one can be shown to have an economic beneftt. However, it is the reduction in plant operations, the minimising of radiation exposure, the inherent flexibility and corresponding overall favourable economics that have led to the development of the dual purpose storage and transport cask in the past few years.     

International Working Group for Sabotage Concerns of Transport and Storage Casks

Abstract

The International Working Group for Sabotage Concerns of Transport and Storage Casks (IWGSTSC), gathers multiple organisations from different countries (for US party Department of Energy, Nuclear Regulatory Commission, and Sandia National Laboratories; for German party Gesellschaft für Anlagen- und Reaktorsicherheit and Fraunhofer Institut; for the French party Institut de Radioprotection et de Sûreté Nucléaire). The goal of the IWGSTSC is to continue cooperation to improve the analytic capabilities, through information sharing and collaborative research and development plus modelling, to understand the potential adverse public health effects and environmental impacts of radiological sabotage directed at or associated with the transport and storage of civilian nuclear material or other civilian radioactive materials. The Parties may also undertake collaborative research and development in other areas of the physical protection of civilian nuclear materials or other radioactive materials. Since 2000, the IWGSTSC has conducted an extensive test programme for the assessment of the aerosol source term produced in the case of spent fuel transport sabotage by a high energy density device, after having examined several scenarios. The major goal of this programme is to produce an accurate estimate of the so called spent fuel ratio in the domain of respirable, aerosol particles produced. All the reports prepared by Sandia National Laboratories have precisely emphasised the important efforts they have made from the beginning and the amount of work already accomplished. In parallel, the International Atomic Energy Agency (IAEA), assisted by technical experts from different countries, has provided a draft document promised to become guidance for the security of radioactive or nuclear materials during transport. The IAEA document contains general guidance addressed to anyone who intends to implement or improve the security of material transports, but the text is, as of today, limited to rather general recommendations. Based on all the knowledge accumulated from past experiments and also based on the work carried out in Vienna at the IAEA, the IWGSTSC members have decided to work on the development of a method for the evaluation of the vulnerability and the source term. So for doing that, joint projects for the research, development, testing and evaluation of the consequences of the malevolent actions during transport are being pursued and are described in this paper.     

Packaging for the Transport and Storage of Spent Sealed Sources in Brazil

Abstract

Since mid-1994, the Centro de Desenvolvimento da Tecnologia Nuclear has been assigned the task of receiving and safely conditioning spent sealed sources until a federal disposal site is available. At the moment (October 1995) there are approximately 1300 sources in the CDTN interim storage hall. As part of the measures taken to accomplish this task, the transport group has developed a simple and low cost packaging, which consists of an outer 200 litre drum surrounding a cylindrical lead shielding, the intermediate space being filled with concrete. In the primary concept a concrete internal lid allowed the sources to be retrieved for future re-encapsulation. In view of a failure in the drop test, a modification was introduced to gather additional information about the ultimate packaging strength, although the resulting concept does not allow future recovery of the contents. The next improvement to be introduced will be the use of a shell-type shock absorber to protect the packaging closing system.     

Adaptation of Approved Transport and Storage Cask Lines to Individual Customer Requirements

Abstract

Within the decommissioning programmes of the Italian nuclear power plants, the Italian multi-utility company ENEL decided to rely on on-site dry storage while waiting for the availability of the national interim storage site. SOGIN (Società Gestione Impianti Nucleari SpA, Rome, Italy), now in charge of all nuclear power plant (NPP) decommissioning activities was created in the ENEL group but is now owned by the Italian government. In 2000 it ordered 30 CASTOR® casks for the storage of its spent fuel not covered by existing or future reprocessing contracts. Ten CASTOR X/A17 casks will contain the Trino pressurised water reactor (PWR) fuel and the Garigliano boiling water reactor (BWR) fuel currently stored in pools at the nuclear power plant Trino and the Avogadro nuclear facility at Saluggia. Additionally 20 CASTOR X/B52 casks will contain the BWR fuel assemblies, which are stored in the pool at the Caorso nuclear power plant. GNB (Gesellschaft fuer Nuklear-Behaelter mbH, Essen, Germany) has completed detailed studies for the design of both types of cask. The tailored cask design is based on the well-established and proven design features of CASTOR reference casks and is responsive to the needs and requirements of the Italian fuel and handling conditions. The design of the CASTOR X/A17 for up to 17 Trino PWR fuel assemblies or 17 Garigliano BWR fuel assemblies and the CASTOR X/B52 cask holding up to 52 Caorso BWR fuel assemblies is suitable for the following conditions of use: loading of the casks in the fuel pools of the nuclear installations at Trino, Caorso and Avogadro; no upgrading of the Current on-site crane capacities; transport of the fuel assemblies, which are currently stored at the Saluggia facility to the nuclear power plant Trino; on-site storage in a vertical or horizontal position with the possibility of transfer to another temporary storage or a final repository, even after a number of years; the partial loading of mixed oxide (MOX) and failed fuel; loading and drying of bottled Garigliano fuel assemblies. On the basis of the CASTOR V/19 and CASTOR V/52 cask lines, the design of the CASTOR X/A17 and X/B52 casks aims at optimising safety and economics under the given boundary conditions. The long time for which fuel is kept in intermediate wet storage results in a reduced shielding and thermal-conduction requirement. This is used to meet the tight mass and geometry restrictions while allowing for the largest cask capacity possible.     

A Family of Transport and Storage Flasks Containing Vitrified High Level Waste

Abstract

The flasks designed for the transport of high level vitrified waste in Germany are described.     

Abstracts on Packaging,Transport, Storage and Security of Radioactive Material

Abstract

The design testing of packages for radioactive materials considers normal operating conditions and accident conditions. A mechanical test, especially under accident conditions, must include the safety assessment of possibly undetected material defects. BAM has developed improved assessment methods, using fracture mechanics, for cracks in the most highly stressed regions of cubic containers made of ductile cast iron. Postulated surface cracks in the centre of the container walls and grooves are investigated numerically. In the static case relations between the crack tip parameters (stress intensity factor or the J integral, respectively), stress load, crack depth, container geometry and material behaviour are derived. In the dynamic case it can be shown by numerical simulations of the drop test of containers onto different targets, even without shock absorbers, that the dynamic crack tip parameter may be estimated by static formulae with the dynamic stress inserted in the intact component. This somewhat surprising result can be explained by the fact that the drop event happens over milliseconds. That is slow enough for the crack to behave quasistatically although the crack is loaded with a dynamic, i.e. time-dependent, stress. Based on these calculations, the critical crack depth is given as a function of the stress, the material quality (defined by the fracture toughness) and the wall thickness for surface cracks in the centre of walls as well as in grooves of a cubic container.     

新型乏燃料贮存格架自主研制

自主研发一种新型乏燃料贮存格架。基于该乏燃料贮存格架的结构,从临界安全、热工水力和力学抗震等方面论证新型乏燃料贮存格架的安全性、可靠性。为了验证新型乏燃料贮存格架的设计合理性和制造工艺可行性,制造1台工程样机,并进行支腿调节试验和模拟组件插拔试验。通过样机的制造和试验表明,新型乏燃料贮存格架设计合理、制造可行,可以批量化生产。