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 - 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 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 - JOUR A1 - Ruiz-Hervias, J. A1 - Simbruner, Kai A1 - Cristobal-Beneyto, M. A1 - Perez-Gallego, D. A1 - Zencker, Uwe T1 - Failure mechanisms in unirradiated ZIRLO® cladding with radial hydrides N2 - The purpose of this paper is to investigate the relationship between hydride morphology, in particular the presence of radial hydrides (RHs), stress state and failure mechanisms associated with the ring compression test (RCT). Samples of ZIRLO®cladding were pre-hydrided and subjected to thermo-mechanical treatments to precipitate long radial hydrides. The results show that the reorientation treatment was very successful. A considerable fraction of RHs was generated, the radial hydride continuity factor being around 80 to 90% of the wall thickness. The samples with reoriented hydrides were tested using the RCT at room temperature. Macroscopic brittle failure was observed with sudden load drops for displacements around 0.5 mm, with a calculated “offset strain”between 0.5 and 1%. Crack nucleation occurs in RHs located in regions with the highest values of hoop stress. These locations are the inner diameter of cladding at the vertical plane of the sample (12 and 6 o’clock positions) and the outer diameter at the horizontal plane (3 and 9 o’clock positions). Noticeable load drops in the RCT are associated with unstable crack propagation events through the radial hydride network, the crack front reaching up to 90% of the wall thickness in some cases. The failure micro-mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth and coalescence in the Zr matrix, with ductile tearing patches connecting neighboring hydrides. The main conclusion is that radial hydride metrics is not the only parameter that determines cladding failure in the presence of RHs, but the interaction between the location and continuity of RHs and the stress normal to the hydride (the hoop stress in this case). Consequently, if a radial hydride is located at a position within the cladding where the hoop stress is small, a crack will not be initiated easily in the RCT. KW - ZIRLO® KW - Cladding Embrittlement KW - Radial Hydrides KW - Failure Mechanism KW - Ring compression test PY - 2021 DO - https://doi.org/10.1016/j.jnucmat.2020.152668 SN - 0022-3115 VL - 544 SP - 152668 PB - Elsevier B.V. AN - OPUS4-52000 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 the correct puncture bar length for the IAEA puncture bar drop test N2 - Among other mechanical tests the 1 meter drop onto a steel puncture bar shall be considered for accident safe packages for the transport of radioactive material. According with 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…”. The most damaging puncture bar length can be estimated by iterative processes in numerical simulations. On the one hand, a sufficient puncture bar length has to 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 - ASME Pressure Vessels and Piping Conference 2018 CY - Prague, Czech Republic DA - 15.07.2018 KW - Length of puncture bar KW - Mechanical assessment KW - Numerical simulation KW - Puncture bar test KW - Transport of radioactive materials PY - 2018 SP - PVP2018-84614, 1 EP - 7 AN - OPUS4-46538 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 AN - OPUS4-60672 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Völzke, Holger 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 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 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 AN - OPUS4-60324 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Sjöland, A. A1 - Christensen, P. A1 - Zetterström Evins, L. A1 - Bosbach, D. A1 - Duro, L. A1 - Farnan, I. A1 - Metz, V. A1 - Zencker, Uwe A1 - Ruiz-Hervias, J. A1 - Rodriguez-Villagra, N. A1 - Kiraly, M. A1 - Schillebeeckx, P. A1 - Rochman, D. A1 - Seidl, M. A1 - Dagan, R. A1 - Verwerft, M. A1 - Herranz Puebla, L. E. A1 - Hordynskyi, D. A1 - Feria, F. A1 - Vlassopoulos, E. T1 - Spent nuclear fuel management, characterisation, and dissolution behaviour: progress and achievement from SFC and DisCo N2 - SFC is a work package in Eurad that investigates issues related to the properties of the spent nuclear fuel in the back-end of the nuclear fuel cycle. Decay heat, nuclide inventory, and fuel integrity (mechanical and otherwise), and not least the related uncertainties, are among the primary focal points of SFC. These have very significant importance for the safety and operational aspect of the back-end. One consequence is the operation economy of the back-end, where deeper understanding and quantification allow for significant optimization, meaning that significant parts of the costs can be reduced. In this paper, SFC is described, and examples of results are presented at about half-time of the work package, which will finish in 2024. The DisCo project started in 2017 and finished in November 2021 and was funded under the Horizon 2020 Euratom program. It investigated if the properties of modern fuel types, namely doped fuel, and MOX, cause any significant difference in the dissolution behavior of the fuel matrix compared with standard fuels. Spent nuclear fuel experiments were complemented with studies on model materials as well as the development of models describing the solid state, the dissolution process, and reactive transport in the near field. This research has improved the understanding of processes occurring at the interface between spent nuclear fuel and aqueous solution, such as redox reactions. Overall, the results show that from a long-term fuel matrix dissolution point of view, there is no significant difference between MOX fuel, Cr+Al-doped fuel, and standard fuels. KW - Spent nuclear fuel management KW - Spent fuel characterization KW - Dissolution behaviour PY - 2023 UR - https://nbn-resolving.org/urn:nbn:de:kobv:b43-572461 DO - https://doi.org/10.1051/epjn/2022029 SN - 2491-9292 VL - 9 SP - 1 EP - 12 PB - EDP Sciences CY - Les Ulis Cedex, France AN - OPUS4-57246 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Völzke, Holger A1 - Zencker, Uwe A1 - Herz, A. A1 - Kreienmeyer, M. T1 - Design of a Drop Test Target with Reproducible Properties for Konrad Prototype Testing N2 - Beginning in 1976 the former iron ore mine Konrad was geologically investigated concerning its suitability as a location for a German final repository for low and intermediate level waste. After a comprehensive licensing procedure it was approved and is now being prepared and equipped for the planned start of storage in 2027. At the end of the 1980s, the requirements for waste containers for storage in this repository were defined for the first time on the basis of an incident analysis of the on-site handling procedures, beginning with the unloading of the packages after the arrival at the site up to the final positioning in the underground storage galleries. A 5-meter drop onto the rock ground of the mine was identified as the covering case for high mechanical requirements (ABK II container class). In contrast to the 9-meter drop according to the IAEA Regulations for the Safe Transport of Radioactive Material, the 5-meter drop is performed onto a hard but not essentially unyielding target. The container is typically not protected by an impact limiter. The requirements for a potential test facility are described in the regulations for the Konrad repository. The mechanical strength of the target is defined as a concrete strength equivalent to the identified properties of the rock ground of the mine. Since 1991 BAM has consistently used precast reinforced concrete slabs as target for drop tests in the framework of licensing procedures as well as in research projects. While the original design fulfilled the requirement for the integrity of the concrete slab in most cases, it failed when drop tests onto an edge of a container were performed. A redesigned concrete slab developed in a research project and suggested as a reference target in 2009 has been successfully used in Konrad licensing procedures since then. The paper gives a brief overview of the historical development and it describes BAM’s efforts and approaches to continuously guarantee a concrete slab of defined quality and to provide a test setup for valid drop tests from the applicants and the authorities view. T2 - PATRAM 2022, 20th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Juan-les-Pins, France DA - 11.06.2023 KW - Target KW - Drop test KW - Final repository Konrad PY - 2023 SP - 1 EP - 11 AN - OPUS4-58563 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 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 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 - Liu, Yung A1 - Völzke, Holger A1 - Zencker, Uwe A1 - Jaunich, Matthias T1 - Cooperation in Advanced Technologies Associated with the Back End of Nuclear Fuel Cycle N2 - A Memorandum of Understanding (MOU) between Argonne National Laboratory (Argonne) and the German Bundesanstalt für Materialforschung und -prüfung (BAM) was signed in October 2014. Its objectives are to promote cooperation among scientists and specialists at Argonne and BAM and establish a framework for collaboration in advanced technologies associated with the back end of the nuclear fuel cycle. Collaborative activities involving Argonne and BAM may be implemented through the promotion of joint research activities and scientific workshops and conferences; exchange of technical information; and visits by scientists, specialists, and graduate, postgraduate, and Ph.D. students. The MOU was renewed in January 2020 for another five years. Highlights of the progress and achievements are provided in identified topical areas for collaboration on ageing management guidance, storage, transportation and disposal R&D, joint conference activities, and conclude with the benefits of the MOU in fostering Argonne and BAM collaboration. T2 - 26th International Conference on Structural Mechanics in Reactor Technology - SMiRT 26 CY - Potsdam, Germany DA - 10.07.2022 KW - Nuclear fuel cycle KW - Back end KW - Spent fuel performance KW - Storage KW - Transportation PY - 2022 SP - 1 EP - 10 AN - OPUS4-55861 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - 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 -