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The regulatory compliance of the containment system is of essential importance for the design assessment of transport packages for radioactive materials. The requirements of the IAEA transport regulations SSR-6 for accident conditions implies high load on the containment system of Type B(U) packages. The integrity of the containment system has to be ensured under the mechanical and thermal tests. The containment system of German transport packages for spent nuclear fuel (SNF) and high level waste (HLW) usually includes bolted lids with metal gaskets. BAM Federal Institute for Materials Research and Testing as the German competent authority for the mechanical and thermal design assessment of approved transport packages has developed the guideline BAM-GGR 012 for the analysis of bolted lid and trunnion systems. According to this guideline the finite element (FE) method is recommended for the calculations. FE analyses provide more accurate and detailed information about loading and deformation of such kind of structures. The results allow the strength assessment of the lid and bolts as well as the evaluation of relative displacements between the lid and the cask body in the area of the gasket groove. This paper discusses aspects concerning FE simulation of lid systems for SNF and HLW transport packages. The work is based on the experiences of BAM within safety assessment procedures. The issues considered are the assessment methods used in the BAM-GGR 012 for bolted lid systems along with the nominal stress concept which is applied for bolts according to that guideline. Additionally, modeling strategies, analysis techniques and the interpretation of the results are illustrated by the example of a generalized bolted lid systems under selected accident conditions of transport.

In recent years BAM was involved in several licensing procedures of new package designs for the transport of radioactive material, where the cask body was made of Ductile Cast Iron (DCI). According to IAEA regulations type B(U) packages must withstand the defined accident conditions of transport. For the cask material DCI, it is necessary to determine the brittle fracture behaviour. Due to the complex structure of the cask body and the dynamic loading a fracture mechanical assessment in an analytical way is not always possible. Numerical calculations are necessary to determine the fracture mechanical load in the component. At the first step a numerical analysis has to be done to identify the loading state at the cask body. Secondly an analysis of a detail of the cask body is made considering the boundary conditions of the global model. An artificial flaw is considered in this detailed model to calculate the fracture mechanical loading state. The size of the artificial flaw is characterized by the ultrasonic inspection used for the quality assurance of the package. The applicant developed additional analysis tools for calculation of stress intensity factor and/or J-Integral. The paper describes the authority assessment approach for the DCI fracture mechanics analysis.

For disposal of the research reactor of the Technical University Munich FRM II a new transport and storage cask design was under approval assessment by the German authorities on the basis of International Atomic Energy Agency (IAEA) requirements. The cask body is made of ductile cast iron and closed by two bolted lid systems with metal seals. The material of the lids is stainless steel. On each end of the cask the wood-filled impact limiters are installed to reduce impact loads to the cask under drop test conditions. In the cavity of the cask a basket for five spent fuel elements is arranged. This design has been assessed by the Bundesanstalt für Materialforschung und -prüfung (BAM) in view to the mechanical and thermal safety analyses, the activity release approaches, and subjects of quality assurance and surveillance for manufacturing and operation of the package. For the mechanical safety analyses of the package a combination of experimental testing and analytical/numerical calculations were applied. In total, four drop tests were carried out at the BAM large drop test facility. Two tests were carried out as a full IAEA drop test sequence consisting of a 9m drop test onto an unyielding target and a 1m puncture bar drop test. The other two drop tests were performed as single 9m drop tests and completed by additional analyses for considering the effects of an IAEA drop test sequence. The main objectives of the drop tests were the investigation of the integrity of the package and its safety against release of radioactive material as well as the test of the fastening system of the impact limiters. Furthermore, the acceleration and strain signals measured during the tests were used for the verification of finite-element (FE) models applied in the safety analysis of the package design. The FE models include the cask body, the lid system, the inventory and the impact limiters with the fastening system. In this context special attention was paid to the modeling of the encapsulated wood-filled impact limiters. Additional calculations by using the verified numerical model were done to investigate e.g. the brittle fracture of the cask body made of ductile cask iron within the package design approval procedure. The thermal safety assessment was based on analytical energy balance calculations and FE analyses. As an additional point of evaluation in frame of approval procedure, the effect of possible impact limiter burning under accident conditions of transport was considered by the applicant and assessed by BAM. This paper describes the package design assessment from the point of view of the competent authority BAM including the applied assessment strategy, the conducted drop tests and the additional calculations by using numerical and analytical methods.