TY - CONF A1 - Simbruner, Kai A1 - Zencker, Uwe A1 - Völzke, Holger T1 - Brittle failure analysis of spent fuel zirconium alloy claddings during long-term dry interim storage - Current approach at BAM N2 - Understanding the mechanical behaviour of fuel cladding is decisive for an extended dry interim storage period beyond the currently licensed period of 40 years in Germany and subsequent handling, transport, and disposal considering normal and accidental conditions. Since fuel cladding acts as the first barrier for spent fuel, its integrity should be demonstrated to enable safe and secure discharge and handling options after interim storage and subsequent transportation in preparation of final disposal. An established method for characterizing cladding material is the Ring Compression Test (RCT), where a small, cylindrical sample of the cladding tube is subjected to a compressive load. This test is a laboratory representation of a fuel rod load case and has shown a high susceptibility to failure under the occurrence of radial hydrides. Certain hydride morphologies and low temperature might even lead to brittle fracture at very small loads. The BAM research project BRUZL (Fracture mechanical analysis of spent fuel claddings during long-term dry interim storage) aims to study the results of quasi-static RCTs and to establish numerical models to gain an in-depth understanding of the stress state during the experiments. Using a fracture mechanical approach, cases of sudden failure during the RCT procedure are used to characterize the material behaviour and establish a failure criterion. As an integral part of the project BRUZL, quasi-static Ring Compression Tests have been performed to identify all experimental details, which might be helpful for numerical modelling. Unirradiated samples of the cladding material ZIRLO® have been subjected to hydrogen charging and a thermo-mechanical treatment for radial hydride reorientation. Sample preparation, testing procedures, and analysis results are presented. A numerical model has been established and an elastic-plastic material model was derived from as-received RCT samples by inverse finite element analyses. Cohesive zone modelling has been implemented to reproduce sudden load drops during RCT. The project is funded by the Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561. T2 - Kolloquium der Abteilung 3 CY - Online meeting DA - 05.11.2020 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2020 AN - OPUS4-51987 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Simbruner, Kai A1 - Zencker, Uwe A1 - Völzke, Holger T1 - Embrittlement of Spent Fuel Claddings - Micromechanically Motivated Cohesive Zone Modelling N2 - Since cladding acts as the first barrier for nuclear fuel, it is key to understand the degrading mechanisms leading to cladding embrittlement. Zirconium alloy based cladding samples subjected to simulated pre-storage drying conditions have shown high susceptibility to brittle failure under the occurrence of hoop stress and low temperature as expected after long-term interim storage. The stress state associated with a compressive pinch-load scenario is reproduced by the Ring Compression Test (RCT), which results in sudden failure at relatively small mechanical loads with high sensitivity to the presence of radial hydrides. As part of the BAM research project BRUZL (Fracture mechanics analysis of spent fuel claddings during long-term dry interim storage), static RCTs were conducted on non-irradiated, pre-hydrided ZIRLO® ring-shaped samples that were subjected to a radial hydride treatment. All samples failed in macroscopically brittle fashion with sudden load drops and severe crack propagation through almost the entire wall thickness. In a post-RCT examination of the fracture surfaces using scanning electron microscopy, radial hydrides could be identified as the weakest link of the structure. However, the microscopic failure mechanism is much more complex due to cracks initiating at different axial positions in depth direction of the ring, overlapping cracks, and arresting cracks. A considerable portion of the fracture surfaces appeared to have undergone plastic deformation of the zirconium matrix resulting in void growth and coalescence associated with ductile failure. The geometry of the ductile patches was measured, and the ductile surface fraction was determined. Based on these observations, a statistical model was developed that can reproduce the composition of the fracture surface as a one-dimensional projection. A micromechanically motivated cohesive zone model (CZM) was implemented in existing RCT finite element analysis (FEA) models with different constitutive laws for brittle hydrides and ductile matrix respectively. In this work, FEA results are presented for non-irradiated ZIRLO® and high-burnup M5® claddings. The impact of varying hydride-matrix combinations is discussed. Acknowledgement: The project was funded by the former Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561. T2 - 6th GRS Workshop on the Safety of Extended Dry Storage of Used Nuclear Fuel CY - Garching, Germany DA - 01.06.2022 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2022 AN - OPUS4-55023 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Simbruner, Kai A1 - Zencker, Uwe A1 - Völzke, Holger T1 - Embrittlement of Spent Fuel Claddings – Fracture Mechanical Evaluation and Failure Criteria N2 - The BAM research project BRUZL (Fracture mechanical analysis of spent fuel claddings during long-term dry interim storage) evaluates results of quasi-static Ring Compression Tests (RCT). Due to its simple realisation, the RCT is an established method for characterising the mechanical behaviour of fuel cladding, where a small tube sample is subjected to a compressive load and the relation between load and displacement is obtained. During tests at room temperature, unirradiated ZIRLO® samples with a distinct extend of radial hydrides have shown high susceptibility to sudden, brittle failure at relatively small loads. Hydride morphology and radial hydride metrics have been an integral part of the characterisation of degradation mechanisms of cladding. A software tool for the analysation of metallographic micrographs was developed. The program allows the automated calculation of hydride lengths, detection of continuous hydride structures, their statistical evaluation and visualisation. Numerical models have been established to reproduce experimental RCT results featuring an elastic-plastic material model, which was derived from inverse finite element analyses of as-received RCT samples. A cohesive zone model (CZM) is introduced to the model to reproduce load drops in the linear regime of the load vs. displacement curve. Using a traction-separation approach for the CZM, the cohesive parameters are fitted to match numerical and experimental results yielding the fracture mechanical material properties. Combining the three key aspects hydride morphology, fracture mechanics and RCT stress-state a failure criterion can be established as a result of the BRUZL project. The project is funded by the Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561. T2 - 5th GRS Workshop on Safety of Extended Dry Storage of Spent Nuclear Fuel CY - Online meeting DA - 09.06.2021 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2021 AN - OPUS4-53779 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 - Vlassopoulos, E. A1 - Caruso, S. A1 - Linnemann, Konrad A1 - Nasyrow, R. A1 - Gretter, R. A1 - Fongaro, L. A1 - Papaioannou, D. T1 - Mechanical integrity of spent nuclear fuel rods N2 - The properties of spent nuclear fuel (SNF) rods change significantly during their operation life in the reactor core. Further changes occur after their discharge mainly due to the heating-cooling processes and possible ageing associated with the cumulative effects of radioactive decay induce damage in the fuel. Such changes may affect the response of the SNF rods to mechanical solicitations corresponding to normal and accidental conditions. Research activities at JRC-KARLSRUHE aim at assessing the integrity of SNF rods and processes which might affect their mechanical properties during their interim storage, transport or other handling operations. JRC Hot Cell facilities have been fully adapted to fulfil the experimental goals. The number of experiments that can be performed, however, is limited and there is an acute need to model them, using this process to validate codes, to deeper understand and to extend the results gained at the JRC beyond the conditions that have been tested. For the experimental campaigns two devices for gravitational impact and 3-point bending tests were developed and installed in a hot cell. Segments of real SNF rods pressurized at their original pressures after discharge have been investigated. The setup is fully operational and new results are reported continuously. T2 - ANS Annual 2018 CY - Philadelphia, PA, USA DA - 17.06.2018 KW - Spent nuclear fuel KW - Mechanical testing KW - Hot cell testing PY - 2018 SP - 170 EP - 172 AN - OPUS4-44862 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger A1 - Jaunich, Matthias A1 - Zencker, Uwe A1 - Simbruner, Kai A1 - Nagelschmidt, Sven A1 - Herbrich, Uwe A1 - Keller, Christian A1 - Qiao, Linan T1 - Research at BAM on Metal Seal Performance and Fuel Rod Integrity N2 - The presentation provides an update on preliminary results from research projects in the area of long term performance of metal seals and fuel rod integrity as safety relevant components of spent fuel transport and storage casks for spent nuclear fuel. T2 - Extended Storage Collaboration Program (ESCP) Winter Meeting 2020 CY - Online meeting DA - 09.11.2020 KW - Metal seal KW - Safety KW - Interim storage KW - Spent nuclear fuel KW - Fuel rod PY - 2020 AN - OPUS4-51615 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger A1 - Wille, Frank A1 - Wolff, Dietmar T1 - Update on the German Spent Fuel Management Strategy N2 - The presentation provides an update on the German spent fuel Management strategy in the areas of transportation, dry interim storage and deep geological disposal. T2 - IAEA 18th TWG-NFCO CY - Online meeting DA - 07.09.2020 KW - Spent nuclear fuel KW - Final disposal KW - Safety KW - Transport and storage cask PY - 2020 AN - OPUS4-51616 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger A1 - Wolff, Dietmar A1 - Herold, C. T1 - Container requirements for high-level radioactive waste disposal in rock salt, claystone, and crystalline rock – Outcomes of the research project KoBrA N2 - Suitable containers for receiving high-level radioactive waste are a key elements of every repository system concerning different host rocks like rock salt, claystone and crystalline rock as being considered in Germany. The disposal container design is decisive regarding boundary conditions for the transport, emplacement techniques, and the operational and long-term safety assessment. The systematic derivation of the requirements to be placed on disposal containers forms the basis for a targeted, comprehensible and transparent development of containers that meet the requirements. The presentation summarizes the outcomes of the research project KoBrA. T2 - IGD-TP Symposium CY - Zurich, Switzerland DA - 20.09.2022 KW - Disposal KW - Container KW - Radioactive waste KW - Safety PY - 2022 AN - OPUS4-57021 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger T1 - Qualification of Nuclear Waste Packages - Needs and Challenges from BAM's Perspective N2 - This Presentation adresses the major needs and challenges of qualification processes for nuclear waste packages performed by BAM. Packages include those for high level radioactive waste like spent fuel as well as those for low and intermediate level waste to be disposed of in the Konrad repository. Beginning with the current German nuclear waste management policy design testing capabilities and experience at BAM is explained followed by illustrating the major strategic process to identify future challenges including specific R&D needs. T2 - 5th International Nuclear Decommissioning Summit CY - Berlin, Germany DA - 30.01.2018 KW - Radioactive KW - Container KW - Testing KW - Safety PY - 2018 AN - OPUS4-44750 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger A1 - Wolff, Dietmar A1 - Zencker, Uwe A1 - Orellana Pérez, Teresa T1 - Ongoing R&D work at BAM N2 - The presentation provides an update on the German nuclear waste management policy towards final disposal and related reasearch at BAM. Topics include long term investigations of metal seals, investigations of fuel cladding failure mechanisms, and the development of requirements and concepts for high level radioactive waste and spent fuel disposal containers. T2 - EPRI Extended Storage Collaboration Program (ESCP) International Subcommittee Meeting 2018 CY - Munich, Germany DA - 05.06.2018 KW - Radioactive waste KW - Metal seal KW - Disposal container KW - Fuel cladding PY - 2018 AN - OPUS4-45328 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - BOOK A1 - Völzke, Holger A1 - Quercetti, Thomas A1 - Ballheimer, Viktor A1 - Nehrig, Marko A1 - Wolff, Dietmar A1 - Wille, Frank T1 - Basic of Transport and Storage of Radioactive Materials N2 - Transport and storage of radioactive materials are performed in countries with policy of either closed or open nuclear fuel cycle. The related technologies have been established by accumulation of experiences and researches including demonstrative tests using full scale or scale models and analyses. Those are essential before commercialization, but are often costly and time consuming. Such demonstrative works should not be repeated meaninglessly, but can be shared through this kind of book and used by readers and the future generations to advance the technology effectively. This book systematically provides findings from lots of valuable researches on safety of transport and storage of radioactive materials under normal and accident conditions that have an impact on basis of safe regulations, designs, and operations. KW - Drop testing KW - Metal seals KW - Transport safety KW - Aging management KW - Package PY - 2018 SN - 978-981-3234-03-1 DO - https://doi.org/10.1142/10820 SP - 1 EP - 376 PB - Worlds Scientific Publishing Co Pte Ltd CY - Singapore AN - OPUS4-44867 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CHAP A1 - Völzke, Holger A1 - Wolff, Dietmar ED - Saegusa, T. ED - Sert, G. ED - Völzke, Holger ED - Wille, Frank T1 - Long-term metal seal performance N2 - For the long-term storage of spent nuclear fuel and other high-Level radioactive waste, dual purpose casks for the transportation and storage are widely used. Usually, these casks consist of a thick-walled monolithic cask body and a bolted double barrier lid System equipped with metal seals (consisting of an inner helical spring and two outer metal layers with a total cross-section diameter between 5 and 10 mm) to ensure the long-term safe confinement of the radioactive inventory. Based on analytical and experimental studies, such metallic Systems have been qualified of storage periods of up to four decades so far as long as proper manufacturing and assembling conditions are met. More than 20 years of Operation experience with a permanently increasing number of casks has shown full compliance with the expectations. In the meantime many countries have to face a significant delay of establishing a high-level radioactive waste repository what leads to Need for extending interim storage periods for probably 60, 80 or 100 years. For that reason the long-term performance of metal seals for such time periods is of specific internst to demonstrate their sufficient safety function and specific investigations have already been launched, e.g. by BAM, to gain more experimental data and to develop validated models to extrapolate the seal performance to longer periods of time and or requested temperature levels. KW - Metal Seal KW - Container KW - Safety KW - radioactive waste PY - 2018 SN - 978-981-3234-03-1 SP - Chapter 20, 293 EP - 303 PB - World Scientific CY - Singapore AN - OPUS4-47669 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Zaghdoudi, Maha A1 - Brandt, Guido A1 - Kömmling, Anja A1 - Jaunich, Matthias A1 - Wolff, Dietmar T1 - Coefficients of Friction in Dependence on Aging State of Elastomers – Experimental Identification and Numerical Simulation of the Experiment N2 - Elastomer seals are mounted as barrier seals in lid systems of containers designed for transport and disposal of negligible heat generating radioactive waste and as auxiliary seals in spent fuel storage and transportation casks (dual purpose casks (DPC)). When the behavior of mounted seals under normal and hypothetical accident conditions of disposal and transport is to be simulated, a comprehensive knowledge of their complex mechanical properties at every state of aging is necessary. In previous works, BAM’s efforts in experimental investigations on specimen artificially aged at different temperatures and times and the implementation of the found results in finite element material models were presented. Additionally, our approaches to reproduce the aging process itself and to extrapolate the results of artificially accelerated aging to longer times were presented. Numerical simulations have shown that the behavior of the seal during mounting and one-sided pressurizing and the resulting performance values such as leakage rate strongly depend on the coefficient of friction (COF) between flange and seal. The friction coefficient, in turn, depends on the aging state of the elastomer material as several publications suggest (see below). Dynamic COF between an exemplary ethylene propylene diene rubber (EPDM) material and a stainless steel ball were determined by using a self-designed linear oscillation tribometer. Unaged and artificially aged EPDM specimen stored for 30 days and 100 days at a temperature of 150 °C were tested. A stainless steel ball (d=10 mm) is brought in contact with the specimen’s surface und loaded by normal forces of 2.5 N, 5 N, 10 N and 20 N. During a reciprocating movement of the EPDM sheet, the horizontal force/friction force is continuously measured, and the COF can be derived. It is well known that friction is a complex phenomenon especially in soft materials. It cannot be excluded that the measured friction force is influenced by additional force components, resulting from the ball’s grooving through the elastomer’s surface. This force depends on the penetration depth of the ball and on the resistance of the elastomer in its different states of aging. The latter results from microstructural changes i.e., chain scission and additional crosslinking that occur during aging which in turn influence the softening or hardening of the material. A finite element (FE) ABAQUS® model was developed to reproduce the measurement process. It should help to better understand the physical mechanisms and to quantify the percentage of measured forces resulting from real friction on the one hand and forces resulting from unintended side effects that could falsify the result on the other hand. The behavior of the elastomer in its different states of aging is reproduced by a FE material model already presented in previous works of BAM. T2 - ASME 2022 Pressure Vessels & Piping Conference (PVP2022) CY - Las Vegas, NV, USA DA - 17.07.2022 KW - Radioactive Waste KW - Elastomers KW - Aging KW - Experiment KW - Seal Behavior KW - Numerical Simulation KW - Leakage Rate KW - Coefficient of Friction PY - 2022 AN - OPUS4-57095 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 AN - OPUS4-49013 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Kömmling, Anja A1 - Zaghdoudi, Maha A1 - Jaunich, Matthias A1 - Wolff, Dietmar T1 - Development and application of a finite element model representing the rapid partial release of elastomeric O-ring seals N2 - Due to their high versatility and recovery potential under several load conditions, seals made from elastomers are widely used as barrier seals in containers for low and intermediate level radioactive waste and as auxiliary seals in spent fuel storage and transportation casks (dual purpose casks (DPC)). In DPC, elastomer seals allow leakage rate measurements of metal barrier seals for demonstration of their proper assembly conditions and performance. For spent fuel and high level waste as well as for low and intermediate level waste, long time periods will be required before final disposal will be available. Therefore, the long-term behavior of elastomer seals can be an important issue when the cask’s suitability for an extended interim storage period has to be assessed. During the licensed interim storage duration, the requested leakage rate has to be maintained under any static and dynamic load conditions. A sufficient resilience, the capability of seals to compensate a (rapid) partial release was identified as an important end-of-lifetime criterion. In the framework of extensive aging and testing programs at BAM, specimen made from representative types of elastomer (EPDM, FKM) were aged for several years at different temperatures ranging from 23 °C to 150 °C. One aim of the experiments was to determine an end-of-lifetime criterion correlated to leakage as the point of seal failure. As the seals remained leak-tight under static conditions even after advanced material degradation, a more demanding leakage test set-up was constructed that allowed a rapid partial release of the seal. This device enables a release of the O-ring by approx. 0.2 mm from 25 % to 23 % compression in less than one second during the pressure rise measurement. If the resilience of the seal has decreased past a certain point, the seal cannot follow the decompression fast enough and a leak path can open. Using a finite element (FE) material model derived earlier from different material tests, this process was simulated with a 3D FE model, reproducing the essential characteristics of the test device. The FE-model enables the analyzation of the occurring global mechanisms if a leak path opens. The capability of the seals to follow the lifted flange and the subsequent run of compression force depending on the current state of aging are illustrated. The findings of the simulation are correlated and discussed with respect to the pressure rise measurements. T2 - ASME PVP 2019 CY - San Antonio, TX, USA DA - 14.07.2019 KW - Storage and transportation casks KW - Elastomeric seals KW - Rapid partial release KW - Leakage KW - Numerical simulation PY - 2019 AN - OPUS4-49017 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 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 - Transport of radioactive materials KW - Mechanical assessment KW - Puncture bar test KW - Length of puncture bar KW - Numerical simulation PY - 2018 AN - OPUS4-45785 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 - ASME 2023 Pressure Vessels & Piping Conference (PVP2023) CY - Atlanta, GA, USA DA - 16.07.2023 KW - Final repository Konrad KW - Drop test KW - Target PY - 2023 AN - OPUS4-58565 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 - Final repository Konrad KW - Drop test KW - Target PY - 2023 AN - OPUS4-58564 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Völzke, Holger A1 - Nieslony, Gregor T1 - Consequences of an Imperfectly Mounted Reinforcement Cage in a Cylindrical Concrete Container During Mechanical Specimen Tests N2 - In 2007 the license for the German Konrad repository for low and intermediate level radioactive waste was finally confirmed. Since then, the site is improved, prepared and equipped for the currently planned start of waste package emplacement in 2029. Bundesanstalt fuer Materialforschung und -pruefung is regularly contracted by the Bundesgesellschaft für Endlagerung as the responsible authority for waste product control, waste container design testing, and waste package acceptance. In this context BAM is mainly responsible for design testing of various container types including the evaluation of quality assurance measures for container manufacturing. In addition to general requirements concerning container design, the casks have to withstand specific mechanical load scenarios. Drop tests from different heights under consideration of the most damaging drop orientation and a stacking test are two examples for necessary safety demonstrations. If the containers are made from concrete the reinforcement has the function to absorb tensile forces the concrete is not able to withstand by its own. Because of manufacturing imperfections of cylindrical concrete containers manufactured already decades ago, the reinforcement cages are not always exactly, asymmetrically positioned when they were poured. This inspired BAM to perform various ABAQUS® finite-element (FE) simulations by using the example of a simplified cylindrical container design with generic dimensions. On the basis of the calculated tensile and compressive stress distributions suggestions are derived on how a cylindrical concrete container with an imperfectly placed reinforcement cage has to be oriented for a drop or stacking test to ensure the most damaging test scenario. T2 - RAMTRANS 2024 CY - London, United Kingdom DA - 14.05.2024 KW - Finite-element-analysis KW - Container design testing procedure KW - Reinforced concrete containers KW - Drop test KW - Stacking test KW - Imperfectly mounted reinforcement cage PY - 2024 AN - OPUS4-60120 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 -