TY - CONF A1 - Zencker, Uwe A1 - Gaddampally, Mohan Reddy A1 - Völzke, Holger T1 - Fracture Mechanics Analysis of Spent Fuel Claddings during Long-Term Dry Interim Storage N2 - The prevention of brittle fracture of spent fuel claddings during long-term dry interim storage is based on experimental investigations, numerical analyses and assessment methods for predicting the mechanical behavior and determining limiting conditions. The ring compression test (RCT) is an established experimental method for characterizing cladding material. Test results for various high-burnup pressure water reactor zirconium-based fuel cladding alloys (e.g., ZIRLO®, M5®) are publicly available. To reduce the effort associated with irradiated samples in hot cells, it is helpful to perform studies on unirradiated surrogate cladding material. Based on such experimental data, load-displacement curves were numerically analyzed for selected cladding materials. In the presence of radial hydrides, a sample may suddenly fail by fracture even at small deformations. Noticeable load drops in the RCT occur associated to unstable crack propagation through the radial hydride network. The failure mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth, and coalescence in the zirconium matrix, with ductile tearing patches connecting neighboring hydrides. The failure process was simulated by cohesive zones controlled by the fracture energy and the cohesive strength. A modeling approach is presented in which the radial hydride morphology is taken into account. Based on the developed fracture mechanics approach with cohesive zone modeling, not only the deformation behavior but also the failure behavior of irradiated as well as unirradiated claddings with radial hydrides under RCT loading conditions can be adequately described. T2 - 20th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 2022) CY - Juan-les-Pins, France DA - 11.06.2023 KW - Extended Interim Storage KW - Nuclear Fuel Cladding KW - Numerical Failure Analysis KW - Ring Compression Test KW - Spent Fuel Characterization PY - 2023 AN - OPUS4-59147 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 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 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 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 - TY - CONF A1 - Zencker, Uwe A1 - Gaddampally, Mohan Reddy A1 - Völzke, Holger T1 - Fracture Mechanics Analysis of Spent Fuel Claddings during Long-Term Dry Interim Storage N2 - The prevention of brittle fracture of spent fuel claddings during long-term dry interim storage is based on experimental investigations, numerical analyses and assessment methods for predicting the mechanical behavior and determining limiting conditions. The ring compression test (RCT) is an established experimental method for characterizing cladding material. Test results for various high-burnup pressure water reactor zirconium-based fuel cladding alloys (e.g., ZIRLO®, M5®) are publicly available. To reduce the effort associated with irradiated samples in hot cells, it is helpful to perform studies on unirradiated surrogate cladding material. Based on such experimental data, load-displacement curves were numerically analyzed for selected cladding materials. In the presence of radial hydrides, a sample may suddenly fail by fracture even at small deformations. Noticeable load drops in the RCT occur associated to unstable crack propagation through the radial hydride network. The failure mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth, and coalescence in the zirconium matrix, with ductile tearing patches connecting neighboring hydrides. The failure process was simulated by cohesive zones controlled by the fracture energy and the cohesive strength. A modeling approach is presented in which the radial hydride morphology is taken into account. Based on the developed fracture mechanics approach with cohesive zone modeling, not only the deformation behavior but also the failure behavior of irradiated as well as unirradiated claddings with radial hydrides under RCT loading conditions can be adequately described. T2 - 20th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 2022) CY - Juan-les-Pins, France DA - 11.06.2023 KW - Extended Interim Storage KW - Nuclear Fuel Cladding KW - Numerical Failure Analysis KW - Ring Compression Test KW - Spent Fuel Characterization PY - 2023 SP - 1 EP - 8 AN - OPUS4-59146 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - RPRT A1 - Caruso, Stefano A1 - Vlassopoulos, Efstathios A1 - Dagan, Ron A1 - Fiorito, Luca A1 - Herm, Michel A1 - Jansson, Peter A1 - Kromar, Marjan A1 - Király, Márton A1 - Leppanen, Jaakko A1 - Feria Marquez, Francisco A1 - Metz, Volker A1 - Papaioannou, Dimitrios A1 - Herranz, Luis Enrique A1 - Rochman, Dimitri A1 - Schillebeeckx, Peter A1 - Seidl, Marcus A1 - Hernandez Solis, Augusto A1 - Stankovskiy, Alexey A1 - Alvarez Velarde, Francisco A1 - Verwerft, Marc A1 - Rodriguez Villagra, Nieves A1 - Zencker, Uwe A1 - Žerovnik, Gasper T1 - EURAD - Work Package 8 - Deliverable 8.1 - State-of-the-art report N2 - A state-of-the-art (SOTA) review on characterisation of spent nuclear fuel (SNF) properties in terms of source term and inventory assessment (neutron, gamma-ray emission, decay heat, radionuclide inventory, elemental content) and in terms of out-of-core fuel performance (cladding performance and fuel integrity in view of the safety criteria for SNF interim storage, transport and canister packaging) using several numerical and experimental approaches and methodologies is presented. This SOTA report is a result of the spent fuel characterisation (SFC) work package as part of the European Joint Programme on Radioactive Waste Management (EURAD), which offers an overview of the status of knowledge in the field of SNF characterisation and assessment during the pre-disposal phase. The document aims to focus on the current safety-significant gaps and related challenges, providing a direct link to the goals of the mandated actors of EURAD. The report is expected to be used by all EURAD colleagues in their national programmes as a key resource for knowledge management programmes and to contribute to demonstrating and documenting the state-of-the-art. KW - Radioactive waste management KW - Spent fuel characterisation KW - Extended interim storage KW - Predisposal PY - 2022 UR - https://www.ejp-eurad.eu/publications/eurad-d81-state-art-report SP - 1 EP - 112 PB - Agence Nationale pour la Gestion des Déchets Radioactifs (ANDRA) CY - Châtenay-Malabry AN - OPUS4-59154 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - EURAD WP 8 Task 3 Overview: Behaviour of nuclear fuel and cladding after discharge N2 - Task 3 of Work package 8 (Spent Fuel Characterization) of the European Joint Programme on Radioactive Waste Management (EURAD) investigates the behaviour of nuclear fuel and cladding after discharge. The aim of the work is to understand and describe the be-haviour of spent nuclear fuel (SNF), irradiated cladding, fuel/cladding chemical interaction (FCCI) and ageing under conditions of extended interim storage, transportation and em-placement in a final disposal system. BAM contributes to the project as partner and leads Task 3. The presentation gives an overview of the project status, main achievements in experimental work and modelling studies, deviations from the plan, delays and challenges ahead. T2 - EURAD Work Package 8 (SFC) Annual Meeting CY - Wettingen, Switzerland DA - 31.10.2023 KW - Nuclear fuel KW - Cladding KW - Discharge PY - 2023 AN - OPUS4-59148 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Spent Fuel Characterization - Current Activities at BAM N2 - The European Joint Programme on Radioactive Waste Management (EURAD) is working on Spent Fuel Characterization (SFC) in its work package (WP) 8. Inspired by the EURAD activities, the International Atomic Energy Agency (IAEA) established an international Coordinated Research Project (CRP) on SFC. The EURAD WP SFC participants are collaborating as a team on the IAEA CRP on SFC. The EURAD WP SFC project consists of four tasks. Task 3 investigates the behaviour of nuclear fuel and cladding after discharge. The aim of these activities is to understand and describe the evolution of the cladding-pellet system and its ageing under conditions of extended interim storage, transportation and emplacement in a final disposal system. At a Consultancy Meeting, BAM as contributor to Task 3 presented current results of the failure analysis of irradiated ZIRLO® claddings under conditions of the Ring Compression Test. T2 - IAEA Consultancy Meeting on the Coordinated Research Project on Spent Fuel Characterization CY - Online meeting DA - 12.09.2023 KW - Nuclear Fuel Cladding KW - Numerical Failure Analysis KW - Ring Compression Test KW - Spent Fuel Characterization KW - Extended Interim Storage PY - 2023 AN - OPUS4-58276 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -