TY - CONF A1 - Linnemann, Konrad A1 - Ballheimer, Viktor A1 - Sterthaus, Jens A1 - Rolle, Annette A1 - Wille, Frank A1 - Nasyrow, R. A1 - Papaioannou, D. A1 - Rondinella, V. A1 - Vlassopoulos, E. A1 - Pautz, A. T1 - Finite element modeling of spent fuel rod segments under bending loads N2 - Transport packages for spent nuclear fuel have to be assessed with respect to specific transport conditions which are defined in the regulations of the International Atomic Energy Agency. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important Inputs for the evaluation of the package capabilities under These conditions. Cracks or failures in the fuel rod cladding can cause the release of gas, volatiles or fuel particles into the cavity. The amount of substances in the cavity has to be considered in the assessment of the activity release and criticality safety. The mechanical analysis of the compound system formed by the fuel rod cladding and the spent fuel pellets is very difficult due to the limited knowledge of the material properties and the insufficient understanding of the interaction between pellets and cladding and between adjacent pellets. The variation of fuel assembly properties regarding cladding material, burn-up and the history of usage makes reliable predictions of the fuel rod behavior even harder. For a better understanding about the behavior of spent fuel rods, JRC and BAM have started a joint research project. In this context, JRC has developed a test device which allows quasi-static 3-point-bending test on fuel rod segments in the hot cell. The loads are applied with respect to the boundary conditions of the activity release assessment. This paper deals with the numerical calculation of a single fuel rod segment under bending load. The aim is to identify the governing mechanical parameters by the variation of constitutive assumptions, contact conditions, inner constraints, etc. This knowledge helps for the interpretation of the experimental results. Furthermore, the improved understanding about the behavior of the cladding-pellets system will be beneficial for the assessment of spent fuel transport conditions. T2 - SMIRT24 - 24th Conference on Structural Mechanics in Reactor Technology CY - Busan, Korea DA - 20.08.2017 KW - Finite element methods KW - Spent fuel assessment KW - Transport packages PY - 2017 SP - 1 EP - 8 AN - OPUS4-45316 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Nasyrow, R. A1 - Papaioannou, D. A1 - Rondinella, V. A1 - Vlassopoulos, E. A1 - Linnemann, Konrad A1 - Ballheimer, Viktor A1 - Sterthaus, Jens A1 - Rolle, Annette A1 - Wille, Frank A1 - Caruso, St. T1 - Bending test device for mechanical integrity studies of spent nuclear fuel rods N2 - This paper presents data obtained from experiments performed using a bending test set-up developed at the Joint Research Centre (JRC) – Karlsruhe, for spent fuel segment testing. Adjustable sample holders, loading modes and other experimental conditions can be im- plemented in the experiments to study the effects of different deformation ranges up to cladding failure. The experimental set-up has been adapted to hot cell remote controlling and has a modular configuration, which allows manual and motor-driven loading option. The device has been calibrated on hydrogenated, unirradiated cladding tube segments filled with alumina pellets. The final application of present set-up is to test non-defueled spent fuel rod segments, pressurized to the original spent fuel rod pressure level. The range of applicability of this device, the scope of the experimental program and the first results from actual bending tests will be discussed. T2 - PATRAM 2016 - 18th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Kobe, Japan DA - 18.09.2016 KW - Hot cell testing KW - Spent fuel assessment KW - Transport packages PY - 2016 SP - Paper 6023, 1 AN - OPUS4-40002 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Linnemann, Konrad A1 - Ballheimer, Viktor A1 - Sterthaus, Jens A1 - Rolle, Annette A1 - Wille, Frank A1 - Nasyrow, R. A1 - Papaioannou, D. A1 - Vlassopoulos, E. A1 - Rondinella, V. T1 - Analysis of parameters affecting the bending behavior of spent fuel rods N2 - Transport packages for spent nuclear fuel have to be assessed with respect to specific transport conditions which are defined in the regulations of the International Atomic Energy Agency. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important inputs for the evaluation of the package capabilities under these conditions. Cracks or failures in the fuel rod cladding can cause the release of gas, volatiles or fuel particles into the cavity. The amount of substances in the cavity has to be considered in the assessment of the activity release and criticality safety. The mechanical analysis of the compound system formed by the fuel rod cladding and the spent fuel pellets is very difficult due to the limited knowledge of the material properties and the insufficient understanding of the interaction between pellets and cladding and between adjacent pellets. The variation of fuel assembly properties regarding cladding material, burn-up and the history of usage makes reliable predictions of the fuel rod behavior even harder. For a better understanding about the behavior of spent fuel rods, JRC-ITU and BAM have started a joint research project. In this context, JRC-ITU has developed a test device which allows quasi-static 3-point-bending test on fuel rod segments in the hot cell. The loads are applied with respect to the boundary conditions of the activity release assessment. This paper deals with the numerical calculation of a single fuel rod segment under bending load. The aim is to identify the governing mechanical parameters by the variation of constitutive assumptions, contact conditions, inner constraints, etc. This knowledge helps for the interpretation of the experimental results. Furthermore, the improved understanding about the behavior of the cladding-pellets system will be beneficial for the assessment of spent fuel transport conditions. T2 - PATRAM 2016 - 18th International symposium on the packaging and transportation of radioactive materials CY - Kobe, Japan DA - 18.09.2016 KW - Transport packages KW - Finite element methods KW - Spent fuel assessment PY - 2016 SP - Paper 2012, 1 AN - OPUS4-40000 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Linnemann, Konrad A1 - Ballheimer, Viktor A1 - Müller, Lars A1 - Rolle, Annette A1 - Wille, Frank A1 - Droste, Bernhard T1 - Aspects of spent fuel behavior assessment for transport packages N2 - Transport packages for spent nuclear fuel have to be assessed with respect to specific transport conditions which are defined in the safety regulations of the International Atomic Energy Agency. In general, gastight fuel rods constitute the first barrier of the containment system. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important inputs for the evaluation of the package safety under transport conditions. The objective of this paper is to discuss the methodologies accepted by BAM for the authority assessment of spent fuel behavior within the package design approval procedure of German package designs. In particular, cracks or failures in the fuel rod cladding can occur under regulatory transport conditions. These defects can cause the release of gas, volatiles, fuel particles or fragments into the package cavity and have to be considered properly in the safety analysis. Another issue is the transport of defective fuel rods. One concept is to use special canisters which can be handled like fuel assemblies. This concept requires additional assessment concerning drying, sealing and the mechanical and thermal design of such canisters. The package as a mechanical system is characterized by a complex set of interactions, e.g. between the fuel rods within the assembly as well as between the fuel assemblies, the basket, and the cask containment. This complexity together with the limited knowledge about the material properties and the variation of the fuel assemblies regarding cladding material, burn-up and the operation history makes an exact mechanical analysis of the fuel rods nearly impossible. The simplified approaches to consider conservatively spent fuel behavior currently accepted by BAM are presented here. T2 - International Conference on Management of Spent Fuel from Nuclear Power Reactors - An Integrated Approach to the Back-End of the Fuel Cycle CY - Wien, Austria DA - 15.06.2015 KW - Transport packages KW - Spent fuel assessment PY - 2015 SP - 1 AN - OPUS4-38102 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -