TY - RPRT 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 - Wiss, Thierry A1 - Zencker, Uwe A1 - Žerovnik, Gasper T1 - EURAD Work Package 8 Deliverable 8.2 Updated State-of-the-Art Report N2 - The state-of-the-art (SOTA) report offers an overview of the status of knowledge in the area of spent nuclear fuel (SNF) characterisation and assessment during the pre-disposal phase using several numerical and experimental approaches and methodologies. A review on characterisation of SNF properties in terms of source term and inventory assessment (neutron, gamma-ray emission, decay heat, radionuclide inventory, and 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) is presented. This updated SOTA report provides the progress made in the spent fuel characterisation (SFC) work package as part of the European Joint Programme on Radioactive Waste Management (EURAD), documents the identified technical gaps, and provides recommendations for future work. 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 - 2024 UR - https://www.ejp-eurad.eu/publications/d82-updated-state-art-report SP - 1 EP - 182 AN - OPUS4-61218 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 behaviour of spent nuclear fuel (SNF), irradiated cladding, fuel/cladding chemical interaction (FCCI) and ageing under conditions of extended interim storage, transportation and emplacement 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 - Kalmar, Sweden DA - 19.09.2022 KW - Nuclear fuel KW - Cladding KW - Discharge PY - 2022 AN - OPUS4-56120 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - EURAD WP 8 (SFC) Task 3 - Update: Behaviour of nuclear fuel and cladding after discharge 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 the Second Research Coordination Meeting on Spent Fuel Characterization, BAM as leader of Task 3 reported on the status of the research work. T2 - IAEA 2nd Research Coordination Meeting of the Coordinated Research Project on Spent Fuel Characterization CY - Kalmar, Sweden DA - 20.09.2022 KW - Nuclear fuel KW - Cladding KW - Discharge PY - 2022 AN - OPUS4-56121 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Völzke, Holger T1 - Brittle Failure Limits of Spent Fuel Claddings Subjected to Long-Term Dry Interim Storage Conditions N2 - The mechanical properties of spent fuel claddings can be adversely affected under the conditions of long-term dry interim storage, so that the failure limits may be reached in case of mechanical loads during handling or transport after storage. Pre-storage drying and the early stage of interim storage can expose the cladding to higher temperatures and higher tensile hoop stresses than those associated with in-reactor operation and pool storage. During slow cooling of a cladding tube under internal pressure, radial hydrides may precipitate in zirconium-based cladding alloys. This can lead to embrittlement of the material and sudden failure of the cladding integrity under mechanical stress. In order to prevent brittle failure, numerical methods are being developed to predict the mechanical behaviour and identify limiting conditions. Experimental investigations, numerical analyses and evaluation methods are discussed. An established experimental method for characterising cladding materials is the Ring Compression Test (RCT). Some test results on irradiated cladding tubes after operation in pressure water reactors are publicly available. However, it is helpful to carry out studies on unirradiated surrogate claddings with similar material properties to reduce the effort associated with irradiated samples in hot cells and to perform material tests with a wider range of parameters. On the basis of such experimental data, load-displacement curves have been numerically analysed for a selection of cladding materials. Radial hydrides can cause a sample to break suddenly due to fracture even at low deformation. Noticeable load drops in the RCT are caused by 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 adjacent hydrides. The cohesive zone approach was used to simulate the failure process taking into account the radial hydride morphology. The developed method can adequately describe both the deformation and failure behaviour of irradiated as well as unirradiated claddings of zirconium-based alloys with radial hydrides under RCT conditions. Limiting conditions can be expressed in terms of fracture energy and cohesive strength. T2 - IAEA International Conference on the Management of Spent Fuel from Nuclear Power Plants - Meeting the Moment CY - Vienna, Austria DA - 10.06.2024 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2024 SP - 1 EP - 10 AN - OPUS4-60323 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Brittle Failure Limits of Spent Fuel Claddings Subjected to Long-Term Dry Interim Storage Conditions N2 - The mechanical properties of spent fuel claddings can be adversely affected under the conditions of long-term dry interim storage, so that the failure limits may be reached in case of mechanical loads during handling or transport after storage. Pre-storage drying and the early stage of interim storage can expose the cladding to higher temperatures and higher tensile hoop stresses than those associated with in-reactor operation and pool storage. During slow cooling of a cladding tube under internal pressure, radial hydrides may precipitate in zirconium-based cladding alloys. This can lead to embrittlement of the material and sudden failure of the cladding integrity under mechanical stress. In order to prevent brittle failure, numerical methods are being developed to predict the mechanical behaviour and identify limiting conditions. Experimental investigations, numerical analyses and evaluation methods are discussed. An established experimental method for characterising cladding materials is the Ring Compression Test (RCT). Some test results on irradiated cladding tubes after operation in pressure water reactors are publicly available. However, it is helpful to carry out studies on unirradiated surrogate claddings with similar material properties to reduce the effort associated with irradiated samples in hot cells and to perform material tests with a wider range of parameters. On the basis of such experimental data, load-displacement curves have been numerically analysed for a selection of cladding materials. Radial hydrides can cause a sample to break suddenly due to fracture even at low deformation. Noticeable load drops in the RCT are caused by 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 adjacent hydrides. The cohesive zone approach was used to simulate the failure process taking into account the radial hydride morphology. The developed method can adequately describe both the deformation and failure behaviour of irradiated as well as unirradiated claddings of zirconium-based alloys with radial hydrides under RCT conditions. Limiting conditions can be expressed in terms of fracture energy and cohesive strength. T2 - IAEA International Conference on the Management of Spent Fuel from Nuclear Power Plants - Meeting the Moment CY - Vienna, Austria DA - 10.06.2024 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2024 AN - OPUS4-60324 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Fracture Mechanics Analysis of Spent Fuel Claddings during Long-Term Dry Interim Storage 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 the Third Research Coordination Meeting on Spent Fuel Characterization, BAM as contributor to Task 3 reported on the activities and status of its research work. T2 - IAEA 3rd Research Coordination Meeting of the Coordinated Research Project on Spent Fuel Characterization CY - Cockermouth, UK DA - 01.07.2024 KW - Nuclear fuel KW - Cladding KW - Discharge PY - 2024 AN - OPUS4-60668 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 - Failure Analysis on Irradiated Claddings Subjected to Long-Term Dry Interim Storage N2 - Long-term dry interim storage may adversely affect the mechanical properties of spent fuel rods, possibly resulting in a reduced resilience during handling or transport after storage. The cladding is the first barrier for the spent fuel pellets. An established method for characterising the cladding material is the ring compression test (RCT), in which a small, cylindrical sample of the cladding tube is subjected to a compressive load. Radial hydrides may precipitate in zirconium-based alloys (Zircaloy) under pre-storage drying and during slow cooling, which result in embrittlement of the cladding material and eventually a possible sudden failure of cladding integrity under additional mechanical loads. The focus of the presented research is on the development of appropriate nu-merical methods for predicting the mechanical behaviour and identification of limiting conditions to prevent brittle fracture of Zircaloy claddings. A modelling approach based on cohesive zones is ex-plained which is able to reproduce the propagation of cracks initiated at radial hydrides in the zir-conium matrix. The developed methods are applied to defueled samples of cladding alloy ZIRLO®, which were subjected to a thermo-mechanical treatment to reorient existing circumferential hy-drides to radial hydrides. A selected sample showing sudden load drops during a quasi-static ring compression test is analysed by means of fracture mechanics for illustrative purposes. Based on the developed fracture mechanics approach, not only the deformation behaviour but also the fail-ure behaviour of irradiated as well as unirradiated Zircaloy claddings with radial hydrides under RCT loading conditions can be adequately described. T2 - 27th International Conference on Structural Mechanics in Reactor Technology - SMiRT 27 CY - Yokohama, Japan DA - 03.03.2024 KW - Cladding Embrittlement KW - Cohesive Zone Modelling KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2024 SP - 1 EP - 8 AN - OPUS4-60671 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Failure Analysis on Irradiated Claddings Subjected to Long-Term Dry Interim Storage N2 - Long-term dry interim storage may adversely affect the mechanical properties of spent fuel rods, possibly resulting in a reduced resilience during handling or transport after storage. The cladding is the first barrier for the spent fuel pellets. An established method for characterising the cladding material is the ring compression test (RCT), in which a small, cylindrical sample of the cladding tube is subjected to a compressive load. Radial hydrides may precipitate in zirconium-based alloys (Zircaloy) under pre-storage drying and during slow cooling, which result in embrittlement of the cladding material and eventually a possible sudden failure of cladding integrity under additional mechanical loads. The focus of the presented research is on the development of appropriate nu-merical methods for predicting the mechanical behaviour and identification of limiting conditions to prevent brittle fracture of Zircaloy claddings. A modelling approach based on cohesive zones is ex-plained which is able to reproduce the propagation of cracks initiated at radial hydrides in the zir-conium matrix. The developed methods are applied to defueled samples of cladding alloy ZIRLO®, which were subjected to a thermo-mechanical treatment to reorient existing circumferential hy-drides to radial hydrides. A selected sample showing sudden load drops during a quasi-static ring compression test is analysed by means of fracture mechanics for illustrative purposes. Based on the developed fracture mechanics approach, not only the deformation behaviour but also the fail-ure behaviour of irradiated as well as unirradiated Zircaloy claddings with radial hydrides under RCT loading conditions can be adequately described. T2 - 27th International Conference on Structural Mechanics in Reactor Technology - SMiRT 27 CY - Yokohama, Japan DA - 03.03.2024 KW - Cladding Embrittlement KW - Cohesive Zone Modelling KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2024 AN - OPUS4-60672 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Simbruner, Kai A1 - Billone, Michael C. A1 - Zencker, Uwe A1 - Liu, Yung Y. A1 - Völzke, Holger T1 - Brittle failure analysis and modeling of high-burnup PWR fuel cladding alloys N2 - The aim of this research is the development of methods for predicting mechanical behavior and identification of limiting conditions to prevent brittle failure of high-burnup (HBU) pressure water reactor (PWR) fuel cladding alloys. A finite element (FE) model of the ring compression test (RCT) was created to analyze the failure behavior of zirconium-based alloys with radial hydrides during the RCT. An elastic-plastic material model describes the zirconium alloy. The stress-strain curve needed for the elastic-plastic material model was derived by inverse finite element analyses. Cohesive zone modeling is used to reproduce sudden load drops during RCT loading. Based on the failure mechanism in non-irradiated ZIRLO® claddings, a micro-mechanical model was developed that distinguishes between brittle failure along hydrides and ductile failure of the zirconium matrix. Two different cohesive laws representing these types of failure are present in the same cohesive interface. The key differences between these constitutive laws are the cohesive strength, the stress at which damage initiates, and the cohesive energy, which is the damage energy dissipated by the cohesive zone. Statistically generated matrix-hydride distributions were mapped onto the cohesive elements and simulations with focus on the first load drop were performed. Computational results are in good agreement with the RCT results conducted on high-burnup M5® samples. It could be shown that crack initiation and propagation strongly depend on the specific configuration of hydrides and matrix material in the fracture area. KW - Cladding KW - Radial hydrides KW - Ring compression test KW - Cohesive zone model PY - 2024 DO - https://doi.org/10.1515/kern-2024-0109 SN - 2195-8580 SP - 1 EP - 9 PB - Walter de Gruyter GmbH AN - OPUS4-62409 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Simbruner, Kai A1 - Völzke, Holger T1 - Analysis of Zirconium Hydride Structures in Spent Fuel Claddings Subjected to Storage Conditions N2 - Pre-storage drying and the early stage of storage can subject spent fuel claddings to higher temperatures and higher tensile hoop stresses than those associated with in-reactor operation. Under these conditions, existing circumferential hydrides in zirconium-based alloys can be partly reoriented into radial ones that may result in embrittlement of the cladding material. As a consequence, the failure limits may be reduced under mechanical loads caused by handling or during transport after storage. The potential occurrence of brittle fracture strongly depends on the specific location, size, and orientation of the zirconium hydrides. The morphology of such hydride structures was experimentally investigated in unirradiated, hydrogen-loaded samples made of a zirconium-based alloy. The hydrogen was introduced into the cladding sample by means of a cathodic charging process. The loading process was followed by a thermal treatment to precipitate the hydrides in the circumferential direction and then a thermomechanical treatment to reorient the hydrides in the radial direction. Metallographic methods were used to describe and classify the hydride morphology. The resulting metallographic micrographs were evaluated using suitable radial hydride metrics. The determination of the maximum length of a hydride structure was carried out in two different ways, each with its own advantages and disadvantages. Both methods consider neighboring hydrides as continuous according to the definition of the Radial Hydride Continuity Factor (RHCF). The first option is to measure the Euclidean distance between the two most distant pixels of the hydride structure in the image analysis, whereby the actual maximum size of the hydride structure is determined. The second possibility is to measure the maximum size of a hydride or a hydride structure within a certain arc length in the radial direction of the cladding tube. This value is of particular interest because the ratio of the maximum radial size of a hydride structure to the wall thickness of the cladding tube is used for the safety assessment of spent fuel claddings with radial hydrides using the RHCF. T2 - 21st International Symposium on the Packaging and Transportation of Radioactive Materials - PATRAM 2025 CY - San Antonio, TX, USA DA - 27.07.2025 KW - Extended Interim Storage KW - Nuclear Fuel Cladding KW - Zirconium Hydride Structure KW - Ring Compression Test KW - Micrograph Analysis PY - 2025 SP - 1 EP - 8 CY - Indianapolis, IN, USA AN - OPUS4-63875 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -