TY - CONF A1 - Zencker, Uwe T1 - Spent Fuel Characterization - Current Activities in Germany N2 - The characterization of spent fuel and radioactive waste is essential in the evaluation of the ability of these materials to meet acceptance criteria. Prior to determining a method for spent fuel characterization, it is necessary to identify the type of data required to support safety and other analyses for the storage, transportation and final disposal of spent fuel. BAM’s understanding and capabilities in the field of spent fuel characterization are presented for the scope of a coordinated research project on the topic. T2 - IAEA Technical Meeting on Spent Fuel Characterization for Management of Spent Fuel in the Back End of the Fuel Cycle CY - Vienna, Austria DA - 12.11.2019 KW - Nuclear fuel cycle KW - Back end KW - Spent fuel characterization KW - Storage KW - Transportation PY - 2019 AN - OPUS4-55866 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Simbruner, K. T1 - Embrittlement of spent fuel claddings during long-term dry interim storage - Current approach at BAM N2 - The integrity of fuel rods and cladding tubes must be demonstrated throughout the dry interim storage of spent nuclear fuel in Germany. For at least a relevant number of casks, an extension of interim storage between 40 and 100 years is foreseeable. The drying procedure for the fuel rods in conjunction with the thermomechanical storage conditions may change the microstructure of the cladding tubes and lead to their possible embrittlement. The failure probability might increase under mechanical load by handling procedures (transport after storage) on potentially embrittled cladding tubes of fuel rods especially at low temperatures after long-term storage. Results of ring compression tests on unirradiated pre-hydrided as well as irradiated samples with radial hydrides are investigated by finite element analyses to be able to describe the failure process and to identify failure criteria. T2 - EPRI Extended Storage Collaboration Program (ESCP) Winter 2019 Meeting CY - Charlotte, NC, USA DA - 04.11.2019 KW - Fuel rod KW - Cladding KW - Radial hydrides KW - Ring compression test KW - Failure analysis PY - 2019 AN - OPUS4-56005 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Simbruner, Kai T1 - Brittle failure of spent fuel claddings during long-term dry interim storage N2 - In order to demonstrate the safety of both spent fuel and the storage system, a good understanding of the processes that might cause deterioration is required. The International Atomic Energy Agency (IAEA) continued the coordinated research on Spent Fuel Performance Assessment and Research (SPAR) into a fourth phase starting in 2015 to evaluate fuel and materials performance under wet and dry storage and to assess the impact of interim storage on associated spent fuel management activities (such as handling and transport). BAM’s understanding and experiences in the field of long-term storage of spent fuel and especially spent fuel characterization are presented to assist the preparation of the final project report. T2 - IAEA CRP SPAR-IV Third Research Coordination Meeting (RCM3) CY - Buenos Aires, Argentina DA - 07.10.2019 KW - Fuel rod KW - Cladding KW - Radial hydrides KW - Ring compression test KW - Failure analysis PY - 2019 AN - OPUS4-56006 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Simbruner, Kai T1 - Embrittlement of spent fuel claddings during long-term dry interim storage - Current approach at BAM N2 - In this presentation, the actual situation of long-term dry interim storage of spent nuclear fuel in Germany is reported. An extension of the interim storage period between 40 and 100 years for at least a relevant number of casks is foreseeable. To demonstrate the integrity of fuel rods and cladding tubes throughout the extended dry interim storage, a potential cladding embrittlement should be investigated. The current approach at BAM is a research project on experimental and numerical investigations of sudden failure of unirradiated pre-hydrided cladding tubes after hydride reorientation at small deformation under Ring Compression Test conditions. A fracture mechanics based failure analysis and suggested failure criteria for a potential brittle failure of spent fuel claddings after long-term dry interim storage under mechanical loads are discussed. T2 - IAEA - Second Coordinate Research Meeting and Consultancy Meeting on Ageing Management Programmes for Spent Fuel Dry Storage Systems CY - Lemont, IL, USA DA - 29.04.2019 KW - Ageing Management KW - Extended Storage KW - Spent Nuclear Fuel KW - Cladding Embrittlement KW - Ring Compression Test PY - 2019 AN - OPUS4-48745 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Stress wave propagation with finite element mesh transitions N2 - A reliable finite element analysis presupposes a qualified model of the object to be analyzed. The finite element mesh density may significantly vary throughout a model. A transition between dissimilar meshes can be accomplished by a gradual change of element size or by an interface with constraint conditions. Some issues may arise from such mesh transitions. Generally, a finer mesh can transmit signals of higher frequency than a coarser mesh. Stress wave propagation through the model could be affected by the transition region or the artificially induced interface. Stress waves generated within the fine mesh could be partly encapsulated by the surrounding coarse mesh. Therefore poor mesh transitions could result in reflections or shape changes of stress waves. A thin rod is investigated to demonstrate the effects. This study supports the development of the ASME Guidance Document "Use of Explicit Finite Element Analysis for the Evaluation of Nuclear Transport and Storage Packages in Energy-Limited Impact Events". T2 - 19th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 2019) CY - New Orleans, LA, USA DA - 04.08.2019 KW - Stress Wave Propagation KW - Thin Rod KW - Finite Element Mesh Transition PY - 2019 AN - OPUS4-48916 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Simbruner, Kai T1 - Brittle failure of spent fuel claddings during long-term dry interim storage N2 - The current research project BRUZL (Fracture mechanical analysis of spent fuel claddings during long-term dry interim storage) is presented. General aim of the project is the development of methods for risk assessment and identification of boundary conditions to prevent potential sudden brittle failure of spent fuel claddings at small deformation during long-term dry interim storage and subsequent transportation. The project is funded by the Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561. T2 - 3rd GRS Workshop on Safety of Extended Dry Storage of Spent Nuclear Fuel CY - Garching, Germany DA - 05.06.2019 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2019 AN - OPUS4-48867 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Zencker, Uwe T1 - Numerical approach to determine a package dependent bar length for the iaea pin drop test T2 - Proceedings of PATRAM 2019 N2 - The Federal Institute for Materials Research and Testing (BAM) is assessing the mechanical and thermal safety performance of packages for the transport of radioactive materials. Drop testing and numerical calculations are usually part of the safety case concepts, where BAM is performing the regulatory tests at their own test facility site. Among other mechanical tests the 1 meter drop onto a steel puncture bar shall be considered for accident safe packages. According to the IAEA regulations “the bar shall be of solid mild steel of circular section, 15.0 ± 0.5 cm in diameter and 20 cm long, unless a longer bar would cause greater damage…”. Particularly with regard to the German transport- and storage cask designs, often made from ductile cast iron, an accurate determination of the puncture bar length to guarantee a load impact covering the worst case scenario can be imperative. If the fracture mechanical proof for the cask material shall be provided by a test, small deviations in the concentrated load applied can be decisive for the question if the cask fails or not. The most damaging puncture bar length can be estimated by iterative procedure in numerical simulations. On the one hand, a sufficient puncture bar length shall guarantee that shock absorbers or other attachments do not prevent or reduce the local load application to the package, on the other hand, a longer and thus less stiff bar causes a smaller maximum contact force. The contrary influence of increasing puncture bar length and increasing effective drop height shall be taken into account if a shock absorber is directly placed in the target area. The paper presents a numerical approach to identify the bar length that causes maximum damage to the package. Using the example of two typical package masses the sensitivity of contact forces and puncture bar deformations to the initial length are calculated and assessed with regard to the international IAEA package safety requirements. T2 - PATRAM 2019 CY - New Orleans, LA, USA DA - 04.08.2019 KW - IAEA KW - 1-m-punch-bar-drop-test KW - Numerical approach KW - Bar length KW - Finite element analysis PY - 2019 UR - https://www.inmm.org/INMM-Resources/Proceedings-Presentations/PATRAM-Proceedings.aspx SP - 1 EP - 10 AN - OPUS4-49016 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kömmling, Anja A1 - Simbruner, Kai A1 - Grelle, Tobias A1 - Jaunich, Matthias A1 - Zencker, Uwe A1 - Wolff, Dietmar A1 - Orellana Pérez, Teresa A1 - Völzke, Holger T1 - Research at BAM for evaluating long term safety of container systems and components T2 - Annual Meeting on Nuclear Technology N2 - Safety of long term containment and storage of radioactive waste is an essential issue, which is increasingly gaining international attention. Extending interim storage beyond initial license periods is one of the major challenges worldwide. BAM has been involved in all kinds of radioactive waste container safety evaluation for storage and transportation including their long term performance from the very beginning. Today, all work related to the safe management of radioactive waste is linked by the activity field “Nuclear Waste Management” within the Focus Area “Energy” at BAM and in conjunction with a research strategy and agenda. Most recently, the working group “Long Term Safety of Container Systems and Components” was established to launch and perform research projects related to safety issues concerning the extended interim storage of radioactive waste. For example, within our project LaMEP, we investigate ageing effects of metal and elastomer seals as well as polymeric neutron shielding materials, which are container components that fulfil major safety functions. The results of the experiments help to understand the long term behaviour of these components. Likewise, our project BRUZL started at BAM in 2018 with the aim of developing a fracture mechanics approach to understand and describe a potential brittle failure mechanism of fuel cladding during long-term interim storage. BAM also contributes to the GRS project “Development of methods and models and international exchange for safety evaluation in the frame of interim storage of radioactive waste”. Additionally, BAM is involved in several international research programmes dealing with extended interim storage, such as the European Joint Programme on Radioactive Waste Management (EURAD), the Extended Storage Collaboration Program (ESCP) and IAEA Coordinated Research Projects. The present contribution provides a comprehensive overview of BAM research activities contributing to national and international gap analyses and to filling knowledge gaps regarding the safe long term management of radioactive waste. T2 - Annual Meeting on Nuclear Technology 2019 CY - Berlin, Germany DA - 07.05.2019 KW - Radioactive Waste KW - Long-term safety PY - 2019 SN - 978-3-926956-95-8 SP - 1 EP - 5 AN - OPUS4-51486 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Qiao, Linan A1 - Keller, Christian A1 - Zencker, Uwe A1 - Völzke, Holger T1 - Three-dimensional finite element analysis of O-ring metal seals considering varying material properties and different seal diameters JF - International Journal of Pressure Vessels and Piping N2 - Metal seals of O-ring form are often used in lid-systems of transport and storage casks for radioactive waste in Germany. To investigate their mechanical behaviour, three dimensional (3D) finite element (FE) models were created using solid elements for all of the seal components. The material behaviour of each component is described with a unified static elastic-plastic material model. The total strain is defined as the sum of linear elastic strain and plastic strain with power-law hardening. The model was carefully validated by comparison of Simulation results with experimental results. The influence of material fluctuation of each seal component due to varying properties and the sensitivity of different seal diameters on the seal force are analysed and discussed. The results show that the material properties of helical spring have major influence on seal force and that the influence of seal diameters is negligible small in the studied range. This is very important to use the test results from seals with small diameter for the assessment of seal behaviour with larger diameter as used in transport and storage cask. KW - Sensitivity analysis KW - Transport and storage cask KW - O-ring metal seal KW - Lid-system KW - Finite element analysis KW - Stochastic variation of material properties PY - 2019 DO - https://doi.org/10.1016/j.ijpvp.2019.103953 VL - 176 SP - 102953 PB - Elsevier Ltd. CY - Amsterdam AN - OPUS4-49263 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Qiao, Linan A1 - Völzke, Holger T1 - Fracture mechanical analysis of a cylindrical cast iron cask N2 - The safety evaluation of cask components made of ductile cast iron includes investigations to prevent brittle fracture. Generally, ductile cast iron is endangered by brittle fracture especially at low temperatures (down to -40°C) and in combination with existing crack-like material defects. An applicable method is the assessment of fracture resistance using fracture mechanics according to the IAEA guidelines. The approach is based on the prevention of fracture initiation. For application of these principles for drop loads, account must be taken both of dynamic stresses within the component and dynamic material behavior. Basically, the dynamic stress intensity factor of postulated pre-existing crack-like defects is compared with the dynamic fracture toughness of the material. Applicable numerical and experimental methods for the safety assessment of cask components are demonstrated for the case of an artificially pre-cracked cylindrical cast iron cask which undergoes dynamic loading conditions as result of the hard impact between the cask and a concrete target. The proposed evaluation procedure is a combination of numerical and experimental steps. Exemplarily, the calculated stress intensity factor is compared with measured fracture toughness values from single edge notched bending specimens. T2 - 19th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 2019) CY - New Orleans, LA, USA DA - 04.08.2019 KW - Ductile Cast Iron KW - Brittle Fracture KW - Cylindrical Cask PY - 2019 AN - OPUS4-48915 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Simbruner, Kai A1 - Völzke, Holger T1 - Brittle failure of spent fuel claddings under long-term dry interim storage conditions – Preliminary analysis N2 - The evaluation of cladding integrity is a major issue to be demonstrated in Germany for extended interim storage periods up to 100 years and subsequent transportation considering operational and accidental conditions with respect to reactor operation, cask drying and dry interim storage. The chemical reaction between the zirconium fuel cladding and the cooling water in water-cooled reactors produces hydrogen and zirconium oxide. Hydrogen diffuses into the cladding and precipitates as zirconium hydrides when the solubility limit is reached, preferably oriented in hoop direction. At high temperatures during vacuum drying procedures, the hydrides can dissolve. Over a succeeding period of slow cooling with existing hoop stress the hydrides precipitate again, but partly reoriented along the radial direction of the cladding. This change of microstructure in combination with a decreasing temperature (0.5...2 K/year) during (extended) interim storage and additional mechanical load by handling procedures or under accident conditions could lead to a potential cladding embrittlement and consequently increased failure probability. The current research project BRUZL (Fracture mechanical analysis of spent fuel claddings under long-term dry interim storage conditions) has been launched by BAM to investigate potential sudden brittle failure of spent fuel claddings at small deformation under long-term dry interim storage conditions and is presented. T2 - IAEA International Conference on the Management of Spent Fuel from Nuclear Power Reactors CY - Vienna, Austria DA - 24.06.2019 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2019 AN - OPUS4-48868 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Simbruner, Kai A1 - Völzke, Holger T1 - Brittle failure of spent fuel claddings under long-term dry interim storage conditions – Preliminary analysis T2 - Proceedings of the IAEA International Conference on the Management of Spent Fuel from Nuclear Power Reactors N2 - The evaluation of cladding integrity is a major issue to be demonstrated in Germany for extended interim storage periods up to 100 years and subsequent transportation considering operational and accidental conditions with respect to reactor operation, cask drying and dry interim storage. The chemical reaction between the zirconium fuel cladding and the cooling water in water-cooled reactors produces hydrogen and zirconium oxide. Hydrogen diffuses into the cladding and precipitates as zirconium hydrides when the solubility limit is reached, preferably oriented in hoop direction. At high temperatures during vacuum drying procedures, the hydrides can dissolve. Over a succeeding period of slow cooling with existing hoop stress the hydrides precipitate again, but partly reoriented along the radial direction of the cladding. This change of microstructure in combination with a decreasing temperature (0.5...2 K/year) during (extended) interim storage and additional mechanical load by handling procedures or under accident conditions could lead to a potential cladding embrittlement and consequently increased failure probability. The current research project BRUZL (Fracture mechanical analysis of spent fuel claddings under long-term dry interim storage conditions) has been launched by BAM to investigate potential sudden brittle failure of spent fuel claddings at small deformation under long-term dry interim storage conditions and is presented. T2 - IAEA International Conference on the Management of Spent Fuel from Nuclear Power Reactors CY - Vienna, Austria DA - 24.06.2019 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2019 SP - Paper IAEA-CN-272/49, 1 EP - 8 AN - OPUS4-48869 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Qiao, Linan A1 - Völzke, Holger T1 - Fracture mechanical analysis of a cylindrical cast iron cask T2 - Proceedings of the 19th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 2019) N2 - The safety evaluation of cask components made of ductile cast iron includes investigations to prevent brittle fracture. Generally, ductile cast iron is endangered by brittle fracture especially at low temperatures (down to -40°C) and in combination with existing crack-like material defects. An applicable method is the assessment of fracture resistance using fracture mechanics according to the IAEA guidelines. The approach is based on the prevention of fracture initiation. For application of these principles for drop loads, account must be taken both of dynamic stresses within the component and dynamic material behavior. Basically, the dynamic stress intensity factor of postulated pre-existing crack-like defects is compared with the dynamic fracture toughness of the material. Applicable numerical and experimental methods for the safety assessment of cask components are demonstrated for the case of an artificially pre-cracked cylindrical cast iron cask which undergoes dynamic loading conditions as result of the hard impact between the cask and a concrete target. The proposed evaluation procedure is a combination of numerical and experimental steps. Exemplarily, the calculated stress intensity factor is compared with measured fracture toughness values from single edge notched bending specimens. T2 - 19th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 2019) CY - New Orleans, LA, USA DA - 04.08.2019 KW - Ductile Cast Iron KW - Brittle Fracture KW - Cylindrical Cask PY - 2019 SP - Paper 1209, 1 EP - 7 AN - OPUS4-48914 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -