TY - RPRT A1 - Zencker, Uwe A1 - Aguado, Carlos A1 - Arkoma, Asko A1 - Bertsch, Johannes A1 - Cristóbal Beneyto, Miguel A1 - Feria Marquez, Francisco A1 - Herranz, Luis Enrique A1 - Ivanov, Ivan A1 - Király, Márton A1 - Nasyrow, Ramil A1 - Papaioannou, Dimitrios A1 - Ruiz-Hervias, Jesus A1 - Vlassopoulos, Efstathios T1 - Thermo-mechanical-chemical properties of unirradiated and irradiated samples of spent nuclear fuel rod segments and cladding N2 - The report addresses the experimental characterisation, thermo-mechanical modelling and performance of unirradiated and irradiated samples of spent nuclear fuel rod segments and cladding. Hydrogen is relocated in nuclear fuel claddings by diffusion along temperature, stress, and concentration gradients, and towards a liner if existent. The precipitation of hydrides during cool-down at the beginning of the storage is influenced by mechanical stress. A strong accumulation of hydrides and a disadvantageous hydride alignment in relation to the stress state can affect the mechanical properties and compromise the fuel cladding integrity. Neutron radiography is used to determine spatially resolved hydrogen and hydride distributions in claddings. A liner in duplex claddings attracts hydrogen, which will then no longer be available to form detrimental hydrides in the bulk part of the cladding. Individual three-point bending and gravitational impact tests on filled pressurised spent nuclear fuel rod segments were conducted in hot-cell facilities. The material behaviour under three-point bending in the presence of hydrides and at different temperatures was systematically investigated in quasi-static laboratory tests on unirradiated rod segments with and without surrogate pellets until rupture. Ring compression tests were carried out on irradiated and unirradiated samples at different temperatures and both with and without radial hydrides. With radial hydrides, brittle failure is possible even at low deformation and simultaneously low temperature. Crack initiation will occur on the radial hydride with the highest fracture mechanics load, which depends on the hydride size in the regions with the highest hoop stresses. Due to the high hydrogen content used in creep tests conducted, precipitation hardening of the hydrides in the zirconium alloy significantly reduces the creep rate, as it strengthens the material and inhibits the dislocation sliding. The use of finite element analyses on mechanical tests to determine the cladding properties using a numerical optimisation method based on a large number of automatically generated sensitivity calculations is demonstrated. The brittle fracture of cladding samples with radial hydrides in the ring compression test can be described with a cohesive zone model if the cohesive parameters are selected appropriately and the hydride morphology is specified. Progress was made in the verification, validation and enhancement of spent fuel performance codes by the compilation of a representative validation database (related to rod internal pressure, fission gas release, rod void volume, and in-clad hydrogen distribution), the enhancement of CIEMAT’s FRAPCON-xt code for rod internal pressure prediction at high burnup, the enhancement of CIEMAT’s in-clad hydrogen performance subroutine HYDCLAD with more phenomenological modelling, the extension of INL’s BISON code with a cladding creep law for dry storage, and a benchmark with FRAPCON-xt and BISON simulating in-reactor and dry storage scenarios. KW - Spent Nuclear Fuel KW - Cladding KW - Experimental Characterisation KW - Modelling KW - Fuel Performance Code PY - 2024 SP - 1 EP - 102 PB - European Joint Programme EURAD AN - OPUS4-61008 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - 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 - RPRT A1 - Zencker, Uwe A1 - Simbruner, Kai A1 - Völzke, Holger T1 - Entwicklung eines bruchmechanischen Berechnungsansatzes zur Beschreibung des Festigkeitsverhaltens von Brennstabhüllrohren bei längerfristiger Zwischenlagerung (BRUZL) N2 - Es werden Methoden zur sicherheitstechnischen Bewertung des mechanischen Verhaltens von Brennstabhüllrohren entwickelt, um sprödes Versagen in Belastungsszenarien nach längerfristiger trockener Zwischenlagerung zu verhindern. Unbestrahlte Brennstabhüllrohrsegmente aus der Zirkoniumlegierung ZIRLO® wurden hydriert und thermomechanischen Behandlungen unterzogen, um Hüllrohrmaterial durch Ausscheidung radialer Hydride zu verspröden. Die Proben mit radialen Hydriden wurden im Ring Compression Test (RCT) bei Raumtemperatur untersucht. Bereits bei kleinen Probenverformungen wurden abrupte Kraftabfälle mit instabiler Rissausbreitung durch das Netzwerk radialer Hydride beobachtet. Die Risslängen erreichten in einigen Fällen bis zu 90 % der Wanddicke. Der mikromechanische Versagensmechanismus ist Quasi-Spaltbruch in den Hydriden und Porenbildung, -wachstum und -vereinigung in der Zirkoniummatrix an den duktilen Brücken zwischen benachbarten Hydriden. Es wurde ein Finite-Elemente-Modell des Ring Compression Tests erstellt, um das Versagensverhalten zirkoniumbasierter Legierungen mit radialen Hydriden im RCT zu analysieren. Die Fließkurve für das elastisch-plastische Materialmodell des Matrixmaterials wurde durch eine inverse Finite-Elemente-Analyse ermittelt. Kohäsivzonenmodelle wurden benutzt, um die plötzlichen Kraftabfälle zu reproduzieren. Es wurde ein mikromechanisches Modell mit statistisch generierten Matrix-Hydrid-Verteilungen entwickelt, das zwischen sprödem Versagen entlang der Hydride und duktilem Versagen der Zirkoniummatrix in der Kohäsivzone unterscheidet. Für das jeweilige Materialverhalten wird ein lokales zweiparametriges Versagenskriterium auf Grundlage der Kohäsionsfestigkeit und der Separationsenergie bei einer festgelegten Form des Kohäsivgesetzes vorgeschlagen. Geeignete Kohäsivparameter lassen sich mit einer inversen Finite-Elemente-Analyse des Versagensverhaltens von hydrierten Proben im Ring Compression Test bestimmen. In den Simulationen lag der Schwerpunkt auf dem ersten Kraftabfall. Die Berechnungsergebnisse stimmen gut mit den RCT-Ergebnissen überein. Es konnte gezeigt werden, dass die Rissinitiierung und -ausbreitung stark von der speziellen Anordnung der radialen Hydride und Matrixbrücken in der Bruchzone abhängt. Die numerische Modellbildung wurde an Versuchen mit bestrahlten Proben aus der Zirkoniumlegierung M5® validiert. KW - Zwischenlagerung KW - Brennstabhüllrohr KW - Versagensmechanismus KW - Kohäsivzonenmodell KW - Ring Compression Test PY - 2022 SP - 1 EP - 106 PB - Bundesanstalt für Materialforschung und -prüfung (BAM) CY - Berlin AN - OPUS4-56853 LA - deu 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 - 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 - Orellana Pérez, Teresa A1 - Völzke, Holger A1 - Wolff, Dietmar A1 - Zencker, Uwe T1 - Relevance of cladding failure mechanisms due to zirconium hydride precipitation under extended dry storage conditions in Germany N2 - The German nuclear waste management strategy consists of dry interim storage of spent fuel inside dual purpose casks and its subsequent direct final disposal in a deep geological repository. Interim storage is limited to 40 years so far, but will have to be extended until repository site selection and operation is concluded, potentially up to 100 years. In this context, research on the long term performance of safety relevant cask components and fuel rod integrity during such extended periods of time is of outmost importance. The barrier function of the fuel cladding depends on its integrity under operational and accidental conditions along with storage and subsequent transportation and is mainly affected over time by altering of the cladding microstructure and by thermo-mechanical conditions during reactor operation and subsequent wet and dry storage. This paper focuses on failure mechanisms regarding cladding embrittlement caused by physical phenomena concerning hydrogen dissolution and precipitation of hydrides in the zirconium matrix. Within this context, potential cladding integrity issues under extended interim storage conditions followed by transportation are emphasized. Delayed hydride cracking (DHC) is a well-known time-dependent temperature-activated phenomenon failure mechanism of the fuel cladding during reactor operation. The phenomenon consists of hydrogen diffusion and hydride precipitation at the tip of an initial crack and the propagation of the crack to an approximate distance equal to the length of the hydride. This failure mechanism is temperature-activated phenomenon and enhances along with thermal cycling of the fuel cladding, e. g. during reactor shutdowns. DHC shows limited relevance under quasi static storage conditions with continuously lowering temperatures. However, the literature points that DHC may occur when the temperature during storage decreases to a certain level after some decades and in combination with stress concentrations in the cladding material. When spent fuel is transferred to dry storage by vacuum drying, the cladding temperature raises up to peak temperatures of ~370°C. Under these conditions, hydrides dissolute into the zirconium matrix up to the solubility limit defined by the peak temperature and this amount of hydrogen can precipitate in a radial re-oriented fashion due to the temperature decrease during storage. Cladding materials with radially precipitated hydrides show significant embrittlement. Radially oriented hydrides are perpendicularly oriented to hoop stresses representing the most critical configuration. Brittle failure of the cladding then occurs by crack propagation through radially oriented hydrides and finally through the zirconium matrix. The susceptibility to radial hydride precipitation depends on cladding material type and microstructure, hydrogen content, pre-drying hydride distribution, irradiation conditions, and temperature and stress histories during drying and storage operations. This paper addresses major considerations concerning spent fuel cladding embrittlement due to hydride precipitation with regard to the boundary conditions of dry spent fuel storage in Germany. Analytical, numerical, and experimental approaches are to be discussed by BAM in order to identify the specific needs for future R&D work in that area with the purpose to provide the necessary data base for proper safety demonstration and evaluation along with future extended storage licensing procedures. T2 - 2017 Water Reactor Fuel Performance Meeting CY - Jeju, Korea DA - 10.09.2017 KW - Cladding KW - Hydride precipitation KW - Spent fuel PY - 2017 SP - 1 EP - 8 AN - OPUS4-44749 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Qiao, Linan A1 - Völzke, Holger T1 - Influence of impact angle and real target properties on drop test results of cubic containers N2 - Drop test scenarios with cubic containers without impact limiters at interim storage sites or in a final repository have been investigated by numerical simulations. An ideally flat drop is impossible to conduct as a free fall of a container even under laboratory conditions. Dynamic stresses and strains inside the container structure are sensitive to the impact angle. Even very small impact angles cause remarkable changes in the experimental or numerical results when a flat bottom or wall of a container hits a flat target. For drop tests with transport packages the International Atomic Energy Agency (IAEA) regulations define an essentially unyielding target. In contrast, potential accident scenarios for storage containers are derived from site-specific safety analyses or acceptance criteria in Germany. Each interim storage site or repository has a yielding or so-called real target with individual structural and material properties. The real target acts as a kind of impact limiter. A more conservative container design is required if the impact limiting effect of the target is not considered. T2 - ASME 2017 Pressure Vessels and Piping Conference CY - Waikoloa, Hawaii, USA DA - 16.07.2017 KW - Finite element method KW - Simulation KW - Drop test KW - Impact angle KW - Real target KW - Yielding target PY - 2017 SN - 978-0-7918-5802-8 DO - https://doi.org/10.1115/PVP2017-65731 VL - 7 SP - Article UNSP V007T07A039, 1 EP - 9 AN - OPUS4-43631 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Influence of impact angle and real target properties on drop test results of cubic containers N2 - Drop test scenarios with cubic containers without impact limiters at interim storage sites or in a final repository have been investigated by numerical simulations. An ideally flat drop is impossible to conduct as a free fall of a container even under laboratory conditions. Dynamic stresses and strains inside the container structure are sensitive to the impact angle. Even very small impact angles cause remarkable changes in the experimental or numerical results when a flat bottom or wall of a container hits a flat target. For drop tests with transport packages the International Atomic Energy Agency (IAEA) regulations define an essentially unyielding target. In contrast, potential accident scenarios for storage containers are derived from site-specific safety analyses or acceptance criteria in Germany. Each interim storage site or repository has a yielding or so-called real target with individual structural and material properties. The real target acts as a kind of impact limiter. A more conservative container design is required if the impact limiting effect of the target is not considered. T2 - ASME 2017 Pressure Vessels and Piping Conference CY - Waikoloa, Hawaii, USA DA - 16.07.2017 KW - Finite element method KW - Simulation KW - Drop test KW - Impact angle KW - Real target KW - Yielding target PY - 2017 AN - OPUS4-43632 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 T1 - Structural and Thermal Analyses of Storage Casks T2 - Workshop on Aging Management Issues Related to Long-Term Dry Spent Nuclear Fuel (SNF) Storage CY - Argonne, USA DA - 2014-10-15 PY - 2014 AN - OPUS4-31873 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Numerical Analyses of Crack-like Defects under Impact Loads T2 - Workshop on Aging Management Issues Related to Long-Term Dry Spent Nuclear Fuel (SNF) Storage CY - Argonne USA DA - 2014-10-15 PY - 2014 AN - OPUS4-31875 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Behälterspezifische Sicherheitsbewertung für auslegungsbestimmende Handhabungsstörfälle T2 - 4. RAM-Behältersicherheitstage 2014, BAM CY - Berlin, Germany DA - 2014-03-26 PY - 2014 AN - OPUS4-30711 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Qiao, Linan A1 - Zencker, Uwe A1 - Kasparek, Eva Maria A1 - Völzke, Holger ED - Topping, B.H.V. ED - Iványi, P. T1 - Simulation of damping concrete under severe impact loads using a crushable foam model with damage mechanisms N2 - The deformation and damage behaviour of damping concrete under impact loading conditions is investigated experimentally and numerically. The material model is based on the assumption of crushable foam with volumetric or isotropic hardening combined with ductile and shear damage criteria. Model parameters are determined in static and dynamic compression tests of confined cubic specimens. The derived material model is validated by numerical simulation of penetration tests. The static and dynamic penetration of indenters into uniform as well as assembled bricks made of damping concrete is discussed. Finally, the successful calculation of a large-scale drop test with a heavy cask-like test object onto a realistic damping concrete foundation is demonstrated. T2 - 12th International conference on computational structures technology CY - Naples, Italy DA - 02.09.2014 KW - Damping concrete KW - Impact load KW - Crushable foam material model KW - Damage PY - 2014 SN - 978-1-905088-61-4 DO - https://doi.org/10.4203/ccp.106.233 SN - 1759-3433 SP - Paper 233, 1 EP - 16 PB - Civil Comp Press AN - OPUS4-31480 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Protz, Christian A1 - Zencker, Uwe A1 - Liebich, R. ED - Onate, E. ED - Oliver, J. ED - Huerta, A. T1 - Drop tests and dynamic finite element analyses of steel sheet containers for final disposal of radioactive waste N2 - Within a safety assessment, Containers for radioactive waste have to withstand drop tests at defined conditions. Alternatively to prototype drop tests, numerical methods can be applied, if they are suitable and sufficiently verified. This paper describes the development of a finite element (FE) model of a thin-walled Steel sheet Container used to investigate dynamic load scenarios due to impact events. Experimental and numerical analyses were performed for different drop Orientations. The results are compared to prove the suitability of the FE model. T2 - WCCM XI - 11th World congress on computational mechanics CY - Barcelona, Spain DA - 20.07.2014 KW - Explicit dynamic FEM KW - Impact KW - Drop test KW - Steel sheet container PY - 2014 SP - 1 EP - 12 AN - OPUS4-31137 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Reliability of Cask Designs under Mechanical Loads in Storage Facilities T2 - 12th International Probabilistic Safety Assessment and Management Conference (PSAM12) CY - Honolulu, HI, USA DA - 2014-06-22 PY - 2014 AN - OPUS4-31108 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Orellana Pérez, Teresa A1 - Völzke, Holger A1 - Wolff, Dietmar A1 - Zencker, Uwe T1 - Cladding considerations at BAM regarding extended storage spent fuel N2 - With the clear expectation of the need for extending spent fuel storage in Germany beyond the initial license period of maximum 40 years additional material data and knowledge about the long term performance and integrity of casks and internals like fuel rods will be required to meet major safety goals and regulatory requirements. Based on a broad knowledge base about fuel cladding behavior, degradation mechanisms and related potential failure mechanisms, e. g. induced by reoriented Hydride precipitation, as illustrated in this paper, a specific gap analyses for the German situation regarding fuel conditions and cladding materials should identify the specific knowledge gaps to be addressed and closed in the near future. This is of specific interest as all spent fuel assemblies are going to be transferred to dry cask storage with the next decade and as a consequence of the German nuclear phase-out decision. Within several decades, BAM as a Federal materials research institute gathered a broad range of expertise in the safety evaluation of materials, components and casks under operational and accidental thermal and mechanical conditions. This paper elucidates selected Basic considerations regarding technical and scientific knowledge about the long term aging performance and potential failure mechanisms of fuel cladding materials. Under consideration of the national R&D funding policy which has been adapted to the extended interim storage or pre-disposal needs and the competences of BAM, its future R&D strategy is explained and some R&D objectives have already been identified. BAM will continue and expand ist engagement in the area of investigating fuel cladding material behavior in the long term and developing validated simulation and analyses tools in collaboration with national and international partners. R&D outcomes shall improve the knowledge-base to close identified knowledge gaps in the area of demonstrating fuel cladding integrity and to provide crucial information with regard to future safety assessments for the extended interim spent fuel storage. T2 - 16th International High -Level Radioactive Waste Management Conference (IHLRWM 2017) CY - Charlotte, NC, USA DA - 09.04.2017 KW - Spent fuel KW - Storage KW - Fuel cladding PY - 2017 SP - 718 EP - 725 AN - OPUS4-40412 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Protz, Christian A1 - Zencker, Uwe A1 - Liebich, R. T1 - Explicit finite element analyses of drop tests with thin-walles steel sheet containers for the Konrad repository N2 - Alternatively to experimental drop tests, the mechanical safety analyses of containers for final disposal of radioactive waste with negligible heat generation in the German Konrad repository may be carried out by numerical simulations within the safety assessment procedure. In the past, safety assessments for thin-walled steel sheet containers have been done exclusively by prototype tests and unfavorable drop scenarios were determined by engineering judgment. So far, reliable numerical simulations do not exist. Therefore, a research project was started to develop numerical simulation approaches for drop test analyses and to determine existing safety margins. Comparisons of experimental and numerical results confirm that the Finite Element (FE) model represents the general mechanical behavior of the steel sheet container sufficiently. Simulations have been used to determine an unfavorable drop scenario resulting in large deformation and damage. This paper presents the investigations carried out as well as the further development of the FE model in terms of damage mechanics. T2 - ASME 2015 Pressure vessels & piping conference - PVP2015 CY - Boston, Massachusetts, USA DA - 19.07.2015 KW - Explicit dynamic FEM KW - Impact KW - Drop test KW - Steel sheet container KW - Ductile damage KW - Damage mechanics PY - 2015 SN - 978-0-7918-5702-1 SP - PVP2015-45522, 1-10 AN - OPUS4-33805 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Qiao, Linan A1 - Völzke, Holger T1 - Dynamic finite element analysis of cask handling accidents at storage sites N2 - The safety assessment of casks for radioactive material at interim storage facilities or in final repositories includes the investigation of possible handling accidents if clearly defined test conditions are not available from the regulations. Specific handling accidents usually are the drop of a cask onto the transport vehicle or the floor as well as the collision with the wall of the storage building or another cask. For such load cases an experimental demonstration of cask safety would be difficult. Therefore, numerical analyses of the entire load scenario are preferred. The lessons learnt from dynamic finite element analyses of accident scenarios with thick-walled cubical containers or cylindrical casks are presented. The dependency of calculation results on initial and boundary conditions, material models, and contact conditions is discussed. Parameter sets used should be verified by numerical simulation of experimentally investigated similar test scenarios. On the other hand, decisions have to be made whether a parameter or property is modeled in a realistic or conservative manner. For example, a very small variation of the initial impact angle of a container can cause significantly different stresses and strains. In sophisticated cases an investigation of simpler limit load scenarios could be advantageous instead of analyzing a very complicated load scenario. T2 - ASME 2015 Pressure vessels & piping conference - PVP2015 CY - Boston, Massachusetts, USA DA - 19.07.2015 KW - Finite element method KW - Simulation KW - Cask handling accident KW - Interim storage PY - 2015 SN - 978-0-7918-5702-1 SP - Paper 45606, 1 EP - 10 AN - OPUS4-34937 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Qiao, Linan A1 - Völzke, Holger T1 - Strategies for numerical modelling of metal gaskets in transport and storage casks N2 - Metal gaskets in the lid system of transport and storage casks for radioactive material have to guarantee leak-tightness and safe enclosure of the radioactive inventory under normal and accident conditions during transport, in case of accidents and for the longterm interim storage. For safety assessments by nurnerical simulation of the thermomechanical behaviour of a lid system with metal gaskets, the finite element method offers three options to use: i) special gasket elements, ii) complex three-dimensional modeHing with solid elements, and iii) a simplified axisymmetric approach. Gasket elements can be adjusted by many parameters, but they give only a global representation of the experimentally observed gasket behaviour. For calculations of the entire cask including the lid system with gaskets, nonetheless this approach is recommended. To investigate the hardly measurable impact behaviour in detail or to extrapolate the long-term behaviour, a local modeHing of all parts of a gasket and their interactions with effects like elastic-plastic deformation, creep, relaxation, and friction is necessary. The three-dimensional model can describe the change of contact area between outer jacket of the gasket and flange dependent on the load conditions, what overcomes an essential limitation of special gasket elements. To simplify the problern of investigating the underlying physical effects, an axisymmetric lid system can be modelled with axisymmetric finite elements. Usually an approximation is only necessary for the helical spring of a metal gasket. This paper explains basic ideas for an adequate finite element simulation of cask lid systems with metal gaskets and their thermo-mechanical behaviour under specific load scenarios. T2 - RAMTRANSPORT 2015 - International conference on the radioactive materials transport and storage CY - Oxford, UK DA - 19.05.2015 KW - Metal gasket KW - Finite element method KW - Simulation PY - 2015 SP - Paper RAM15-27, 1-9 AN - OPUS4-34938 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Strategies for Numerical Modelling of Metal Gaskets in Transport and Storage T2 - Radioactive Materials Transport and Storage Conference and Exhibition (RAMTRANSPORT 2015) CY - Oxford, United Kingdom DA - 2015-05-19 PY - 2015 AN - OPUS4-33398 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Quercetti, Thomas A1 - Wieser, Günter A1 - Völzke, Holger A1 - Droste, Bernhard ED - Droste, B. ED - Kowalewsky, H. T1 - Mechanical impact assessment of cubic waste containers depending on target construction T2 - PATRAM 98 - 12th International conference on the packaging and transportation of radioactive materials CY - Paris, France DA - 1998-05-10 PY - 1998 VL - 3 SP - 1152 EP - 1159 PB - Nuclear Technology Publishing CY - Paris, France AN - OPUS4-21556 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Völzke, Holger A1 - Wieser, Günter A1 - Zencker, Uwe A1 - Droste, Bernhard T1 - Mechanical and thermal safety analyses and demonstrations for cubic DCI multipurpose containers PY - 1997 SN - 0957-476X VL - 8 IS - 3-4 SP - 247 EP - 252 PB - Nuclear Technology Publ. CY - Ashford, Kent AN - OPUS4-21664 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger A1 - Wieser, Günter A1 - Zencker, Uwe A1 - Droste, Bernhard T1 - Mechanical and thermal safety analyses and demonstrations for cubic DCI - multi-purpose containers T2 - 4th International conference on transportation for the nuclear industry CY - Bournemouth, Dorset, UK DA - 1997-05-13 PY - 1997 IS - Session 2, Paper 5 SP - 1 EP - 13 AN - OPUS4-21665 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Völzke, Holger A1 - Droste, Bernhard T1 - Entwicklung von Beurteilungsmethoden für Transport- und Lagerbehälter mit erhöhten metallischen Reststoffanteilen - Weiterführende Untersuchungen (EBER II) T2 - VI. Stilllegungskolloquium Hannover und 5. Statusbericht "Stilllegung und Rückbau kerntechnischer Anlagen" CY - Hannover, Deutschland DA - 2000-04-13 PY - 2000 SP - 315 EP - 326 AN - OPUS4-21591 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Völzke, Holger A1 - Droste, Bernhard T1 - Entwicklung von Beurteilungsmethoden für Transport- und Lagerbehälter mit erhöhten metallischen Reststoffanteilen (EBER) - FEM-Berechnungen zum Aufprall kubischer Gußcontainer T2 - V. Stilllegungskolloquium Hannover und 4. Statusbericht "Stilllegung und Rückbau kerntechnischer Anlagen" CY - Hannover, Deutschland DA - 1997-06-24 PY - 1997 SP - 264 EP - 275 AN - OPUS4-21659 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - GEN A1 - Völzke, Holger A1 - Zencker, Uwe A1 - Wieser, Günter A1 - Droste, Bernhard T1 - Testing and licensing of radioactive waste containers for disposal in Germany T2 - DisTec `98 - International conference on radioactive waste disposal CY - Hamburg, Germany DA - 1998-09-09 PY - 1998 VL - 2 SP - 431 EP - 436 PB - KONTEC CY - Hamburg AN - OPUS4-21608 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Wille, Frank A1 - Droste, Bernhard A1 - Müller, Karsten A1 - Zencker, Uwe T1 - Mechanical design assessment approaches of actual spent fuel and HLW transport package designs KW - Package design KW - Spent fuel KW - High level waste KW - Design computations KW - Testing KW - Design testing KW - Transport packages KW - Numerical analysis KW - Mechanical analysis PY - 2010 DO - https://doi.org/10.1179/174650910X12913738452796 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 21 IS - 4 SP - 203 EP - 207 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-23052 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Numerical safety assessment of a transport and storage cask for radioactive materials without impact limiters by the 0.3 m drop test onto an unyielding target T2 - 9th International Conference on Computational Structures Technology CST 2008 CY - Athens, Greece DA - 2008-09-02 PY - 2008 AN - OPUS4-17939 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Numerische Werkstoff- und Bauteilsimulation T2 - 2. RAM-Behältersicherheitstage 2010, BAM CY - Berlin, Germany DA - 2010-03-18 PY - 2010 AN - OPUS4-21067 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Droste, Bernhard A1 - Müller, Karsten A1 - Zencker, Uwe T1 - Recent experiences in mechanical design assessment of spent fuel and HLW casks by competent authority in Germany - 10093 T2 - WM2010 Conference CY - Phoenix, Arizona, USA DA - 2010-03-07 KW - Design testing KW - Transport packages KW - Numerical analysis PY - 2010 SP - 1 EP - 10 AN - OPUS4-21106 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Droste, Bernhard A1 - Zencker, Uwe A1 - Komann, Steffen T1 - Reflektion der verkehrsrechtlichen Bauartprüfung des CASTOR HAW 28M aus Sicht der zuständigen Behörde BAM T2 - Jahrestagung Kerntechnik 2010 CY - Berlin, Deutschland DA - 2010-05-04 KW - Bauartprüfung KW - Fallprüfungen KW - Transportbehälter KW - FEM PY - 2010 SP - 1 EP - 5 CY - Berlin AN - OPUS4-21241 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Bletzer, Claus W. A1 - Zencker, Uwe A1 - Völzke, Holger A1 - Nehrig, Marko T1 - Strategien zur Bestimmung von Temperaturen von Transport- und Lagerbehältern für radioaktive Stoffe T2 - Jahrestagung Kerntechnik 2010 CY - Berlin, Deutschland DA - 2010-05-04 KW - Temperaturprofil KW - Transportbehälter KW - Radioaktive Stoffe PY - 2010 IS - Section 5 / Paper 514 SP - 1 EP - 6 CY - Berlin AN - OPUS4-22604 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Qiao, Linan A1 - Zencker, Uwe A1 - Musolff, André A1 - Komann, Steffen T1 - Dynamic finite element analyses of a spent fuel transport and storage cask with impact limiters by 9 meter drop tests N2 - The 9 meter drop onto an unyielding target is one of the important mechanical tests within the safety assessment of transport casks for radioactive material. In general, the cask is equipped with impact limiters to reduce the dynamic load on the cask body by absorbing a major part of the kinetic energy. The impact limiters are often made of wood or aluminium. In this study an elastic-plastic material model with volume change was used to describe the stress-strain behaviour of wood found in crush tests. For aluminium, an elastic-incremental plastic material model with Cowper-Symonds parameters for strain rate depending material hardening was used to model the adiabatic stress-strain relations measured at specimens at constant ambient temperature. Hereafter simulations with a sophisticated finite element model were carried out and compared with different drop tests. Four drop tests of a half-scale cask model equipped with wood and aluminium impact limiters with different drop positions were selected to investigate the impact limiter behaviour during a 9 meter drop test. All drop tests were simulated with the same FE mesh but under different boundary and initial conditions. T2 - SIMULIA Customer conference CY - Barcelona, Spain DA - 17.05.2011 KW - Cask KW - Drop test KW - Impact limiter KW - Wood KW - Aluminium KW - Dynamics PY - 2011 SP - 932 EP - 944 AN - OPUS4-23706 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Qiao, Linan A1 - Zencker, Uwe A1 - Wille, Frank A1 - Musolff, André T1 - Numerical simulation of 9 meter drop of a transport and storage cask with aluminium impact limiter N2 - For the purpose of numerical simulation of 9 meter drop of a transport and storage cask with aluminium impact limiter, an elastic-incremental plastic material model with strain rate hardening acc. to Cowper-Symonds is used for the development of isothermal as well as adiabatic stress-strain relations of aluminium from the compression test at constant ambient temperature. After that, two different simulation strategies are compared. At first, the drop test is calculated fully coupled, i.e. with isothermal stress-strain relations and possible heat generation in the material. Then the drop test is recalculated in a very simplified manner with adiabatic stress-strain relations from the compression test in an isothermal simulation. Both calculation strategies show similar results in the investigated load scenario. T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Drop test KW - Spent fuel transport cask KW - Finite element calculation KW - Dynamic simulation KW - Impact limiter KW - Aluminium PY - 2010 SP - 1-8 (Monday-T16-117) AN - OPUS4-23707 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kasparek, Eva Maria A1 - Scheidemann, Robert A1 - Zencker, Uwe A1 - Wolff, Dietmar A1 - Völzke, Holger T1 - Effect of dynamic loading on compressional behavior of damping concrete N2 - In drop test scenarios related to assessing and licensing the storage procedure of spent fuel and high active waste, the casks under examination are generally not equipped with impact limiters. Hence, the extent of mechanical stresses in case of an assumed handling accident is largely affected by the ground properties of the reception hall floor in the specific storage facility. Unlike conventional brittle foundation materials, damping concrete performs quite well in such applications as it features high stiffness as well as high energy absorption due to the filler pore volume. However, its damping ability is not sufficiently exploited in current finite element (FE) calculations due to a lack of advanced material models for simulating its impact response. An implementation of qualified concepts that account for plastic, strain rate dependent behavior requires additional information that has to be provided by systematic test series. BAM recently started a research project to generate such data, subsequently to develop and to improve numerical methods for the analysis of impact limiters and damping foundation material and thus to optimize safety assessment tools for the design of transport and storage casks. A major part of this research concerns dynamic compression tests of variably shaped specimens conducted at a servo hydraulic 1MN impact testing machine as well as at a BAM facility for guided drop tests. This presentation focuses 100mm damping concrete cubes deformed vertically at constant rates under different constraint conditions. For example, a special fitting jig was constructed to subject the specimens to multi-axial loading. Thereby a deformation of 60% could be applied. Simulation was conducted by FE code ABAQUS™ based on material models “Concrete damaged plasticity” and “Crushable foam” which both allow defining rate sensitive nonlinear stress-strain relations in compression beyond the classic metal plasticity approach. T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Damping concrete KW - Strain rate sensitivity KW - Numerical simulation KW - Material model PY - 2010 SP - 1-8 (Thursday-T40-114) AN - OPUS4-23708 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kasparek, Eva Maria A1 - Völzke, Holger A1 - Scheidemann, Robert A1 - Zencker, Uwe A1 - Wolff, Dietmar T1 - Experimentelle und numerische Untersuchungen von stoßdämpfenden Werkstoffen für Behälter zur Entsorgung radioaktiver Stoffe N2 - Die mechanische Beanspruchung von Transport- und Lagerbehältern für radioaktive Stoffe gemäß den durch die IAEA vorgegebenen Prüfszenarien oder bei Unfällen, die während der Handhabung in einer kerntechnischen Anlage auftreten können, wird maßgeblich durch das Design der Stoßdämpfer und die Eigenschaften des jeweiligen Untergrundes beeinflusst. Da der Nachweis der Behälterintegrität und -dichtheit häufig an numerische Simulationen gekoppelt ist, sind für die sicherheitstechnische Auslegung der Konstruktion zuverlässige Informationen über das Energieabsorptionsvermögen der Werkstoffe und eine geeignete Umsetzung dieser Daten in FEM - Programme erforderlich. Die im Rahmen von Falluntersuchungen bei der BAM festgestellten, teilweise erheblichen Abweichungen zwischen experimentellen und berechneten Größen sind entsprechend darauf zurückzuführen, dass die zur Verfügung stehenden numerischen Verfahren das komplexe Verhalten der stoßdämpfenden Strukturen nicht ausreichend präzise wiedergeben. Hiervon besonders betroffen sind Bauteile aus Holz und Polyurethan-Schaum (PU-Schaum), die zur Konstruktion von Behälterstoßdämpfern für die Transportkonfiguration eingesetzt werden, sowie Dämpferbeton, der in ausgewählten Bereichen im Fundament von Standort-Zwischenlagern eingebaut wird. Um die zugehörigen Materialmodelle weiter zu entwickeln und die dafür erforderliche Datenbasis zu schaffen, hat die BAM vor knapp zwei Jahren das vom BMBF geförderte Forschungsvorhaben ENREA (Entwicklung rechnerischer Analysemethoden für stoßdämpfende Strukturen beim Anprall oder Absturz von Abfallgebinden) gestartet. Wesentlicher Bestandteil ist ein systematisches Versuchsprogramm, bei dem die drei genannten Werkstoffe statisch und dynamisch hohen Druckbelastungen ausgesetzt werden. Relevante Versuchsparameter sind, neben der Dehnrate, die Temperatur und Lagerungsbedingungen sowie die Orientierung des Werkstoffes hinsichtlich der Belastungsrichtung. Dieser Beitrag präsentiert die wichtigsten Ergebnisse aus der ersten Versuchsphase, die durch weggesteuerte Versuche mit konstanten Prüfgeschwindigkeiten von 0,02 mm/s und 3000 mm/s gekennzeichnet ist. Darüber hinaus wird die Verwertung dieser Daten zur Anpassung von Materialmodellen auf der Grundlage von Nachrechnungen gezeigt, wobei die Möglichkeiten und Grenzen der verwendeten Konzepte Umrissen werden. Aufgrund des Umfangs, den die Versuche mit Holzwerkstoffen einnehmen, und des erforderlichen Aufwandes für die numerische Modellierung werden die entsprechenden Maßnahmen und Ergebnisse in einem eigenständigen Beitrag auf der KONTEC vorgestellt. T2 - KONTEC 2011 - 10. Internationales Symposium "Konditionierung radioaktiver Betriebs- und Stilllegungsabfälle" CY - Dresden, Germany DA - 06.04.2011 KW - Stoßdämpfende Werkstoffe KW - Materialmodellierung KW - Versuchsprogramm KW - Stoßbelastung KW - Dehnratenabhängigkeit KW - Temperaturabhängigkeit KW - Polyurethan Schaum KW - Dämpferbeton PY - 2011 SP - 453 EP - 458 AN - OPUS4-23709 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Kasparek, Eva Maria A1 - Zencker, Uwe A1 - Scheidemann, Robert A1 - Völzke, Holger A1 - Müller, Karsten T1 - Numerical and experimental studies of polyurethane foam under impact loading N2 - Proper predictions of the behaviour of shock absorber materials are of utmost importance in safety assessments for licensing casks for transport and storage of highly active waste. After having identified significant discrepancies between numerical results and the actual response of polyurethane foam limiters subjected to accidental scenarios, a new research project ENREA was established by BAM. A major objective is to enhance and to develop advanced material models intended to simulate limiters under impact loading. They should account for all major factors influencing the load–deformation relationship like temperature, strain rate and specific stress state. The corresponding test program, applicable plasticity models, the overall parameter identification strategy based on local and global optimization techniques as well as experimental and numerical results are presented here in particular for closed cell foams. KW - Polyurethane foam KW - Limiter KW - Impact loading KW - Strain rate KW - Plasticity model KW - Parameter identification PY - 2011 DO - https://doi.org/10.1016/j.commatsci.2010.11.025 SN - 0927-0256 VL - 50 IS - 4 SP - 1353 EP - 1358 PB - Elsevier CY - Amsterdam AN - OPUS4-23151 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Droste, Bernhard A1 - Müller, Karsten A1 - Zencker, Uwe T1 - Mechanical design assessment approaches of actual spent fuel and HLW transport package designs T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Design testing KW - Transport packages KW - Numerical analysis PY - 2010 SP - 1 EP - 8(?) AN - OPUS4-23590 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Qiao, Linan A1 - Zencker, Uwe A1 - Wieser, Günter A1 - Völzke, Holger T1 - Numerische Analyse eines ungedämpften Behälterabsturzes aus 0,3 m Höhe T2 - Jahrestagung Kerntechnik CY - Hamburg, Germany DA - 2008-05-27 KW - Duktiles Gusseisen KW - Behälterdesign KW - Sicherheitsanalyse KW - Dynamische Beanspruchungsbedingungen KW - Finite-Element-Berechnung PY - 2008 SP - 380 EP - 383 PB - INFORUM-Verl. u. Verwaltungsges. CY - Berlin AN - OPUS4-18314 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - GEN A1 - Qiao, Linan A1 - Zencker, Uwe A1 - Wieser, Günter A1 - Völzke, Holger ED - Topping, B. H. V. ED - M. Papadrakakis, T1 - Numerical safety assessment of a transport and storage cask for radioactive materials without impact limiters by the 0.3m drop test onto an unyielding target T2 - 9th International Conference on Computational Structures Technology (CST2008) CY - Athens, Greece DA - 2008-09-02 KW - Impact KW - Simulation KW - Cask KW - Safety KW - Drop KW - Foundation KW - Unyielding KW - Target PY - 2008 SN - 978-1-905088-23-2 IS - Paper 21 SP - 1 EP - 10 PB - Civil Comp Press CY - Kippen AN - OPUS4-18340 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Qiao, Linan A1 - Wieser, Günter A1 - Zencker, Uwe A1 - Droste, Bernhard T1 - Dynamic finite element analyses of a HLW transport cask with polyurethane impact limiters in 9m drop tests T2 - PATRAM 2007, 15th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Miami, Florida, USA DA - 2007-10-21 KW - Ductile iron KW - Cask design KW - Safety assessment KW - Impact limiter KW - Dynamic loading conditions PY - 2007 IS - Session E, Paper #60 SP - 1 EP - 7 PB - Institute of Nuclear Materials Management AN - OPUS4-18400 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Qiao, Linan A1 - Weber, Mike A1 - Kovacs, Oliver A1 - Droste, Bernhard T1 - Application limits of low-ductile cast iron for radioactive waste containers T2 - PATRAM 2007, 15th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Miami, Florida, USA DA - 2007-10-21 KW - Ductile iron KW - Container design KW - Safety assessment KW - Fracture mechanics KW - Dynamic loading conditions PY - 2007 IS - Session B, Paper #97 SP - 1 EP - 8 PB - Institute of Nuclear Materials Management AN - OPUS4-18399 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Entwicklung von Beurteilungsmethoden für Transport- und Lagerbehälter mit erhöhten metallischen Reststoffanteilen (EBER III) T2 - 9. Intern. Symp. "Konditionierung radioaktiver Betriebs- und Stilllegungsabfälle" - KONTEC CY - Dresden, Germany DA - 2009-04-15 PY - 2009 AN - OPUS4-19251 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Beurteilungsmethoden für Transport- und Lagerbehälter mit erhöhten metallischen Reststoffanteilen - Behälterbruchsicherheitsuntersuchungen und Entwicklung eines KONRAD-Referenzprüfstandfundamentes T2 - 9. Intern. Symposium "Konditionierung radioaktiver Betriebs- und Stilllegungsabfälle" einschließlich 9. Statusbericht des BMBF "Stilllegung und Rückbau kerntechnischer Anlagen" (KONTEC 2009) CY - Dresden, Germany DA - 2009-04-15 PY - 2009 AN - OPUS4-19300 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Methoden der numerischen Nachweisführung für mechanische und thermodynamische Beanspruchungen T2 - 1. RAM-Behältersicherheitstage 2009 CY - Berlin-Horstwalde, Germany DA - 2009-03-26 PY - 2009 AN - OPUS4-19175 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Weber, Mike A1 - Qiao, Linan A1 - Menzel, Martin A1 - Droste, Bernhard T1 - Beurteilungsmethoden für Transport- und Lagerbehälter mit erhöhten metallischen Reststoffanteilen - Behälterbruchsicherheitsuntersuchungen und Entwicklung eines KONRAD-Referenzprüfstandfundamentes T2 - 9. Internationales Symposium "Konditionierung radioaktiver Betriebs- und Stilllegungsabfälle" (KONTEC 2009) CY - Dresden DA - 2009-04-15 KW - Duktiles Gusseisen KW - Behälterdesign KW - Sicherheitstechnische Bewertung KW - Bruchmechanik KW - Dynamische Belastungsbedingungen KW - KONRAD KW - Referenzprüfstandfundament PY - 2009 SP - 646 EP - 652 PB - KONTEC CY - Hamburg AN - OPUS4-19396 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -