TY - CONF A1 - Qiao, Linan A1 - Zencker, Uwe A1 - Völzke, Holger ED - Brandt, A. M. ED - Olek, J. ED - Glinicki, M. A. ED - Leung, C.K.Y. ED - Lis, J. T1 - Crushable foam model with pressure and stress-triaxiality dependent damage mechanisms for damping concrete N2 - Damping concrete with high energy absorption capability consists of a cement matrix with embedded small polystyrene balls and is investigated experimentally and numerically under high static compression with and without clamping. A material model is derived which describes the fundamental effects of damping concrete like non-linear elastic-plastic behaviour, volume change, volume strain dependent hardening as well as shear failure with an adequate description of damage initiation and evolution. The suggested material model is validated by simulation of penetration tests. T2 - BMC-11, International Symposium on Brittle Matrix Composites CY - Warsaw, Poland DA - 28.09.2015 KW - Damping concrete KW - Impact load KW - Crushable foam material model KW - Damage PY - 2015 SN - 978-83-89687-96-8 SP - 269 EP - 277 PB - Inst Fundamental Technological Research, Polish Acad Sciences CY - Warschau AN - OPUS4-34613 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger T1 - Research activities at BAM related to extended spent fuel and HLW storage N2 - This presentation starts with a brief overview about the current situation of spent nuclear fuel and high level waste storage in Germany followed by a discussion of the expected timeline towards final disposal. Furthermore an overview is presented about research programmes for radioactive waste disposal and also extended interim storage issues due to the delay of the national repository siting procedure. In the second part of the presentation BAM research activities related to extended interim storage are addressed. This includes the status and preliminary results of proactive R&D projects concerning metal seals, degradation effects of polymers used as neutron shielding materials and finally elastomer seals. T2 - RIC 2016 - 28th Annual regulatory information conference CY - North Bethesda, Maryland, USA DA - 08.03.2016 KW - Spent Fuel KW - Extended interim storage KW - Radioactive waste KW - Degradation effects PY - 2016 UR - https://ric.nrc-gateway.gov/docs/abstracts/volzkeh-w22-hv.pdf AN - OPUS4-35614 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger T1 - Safety demonstrations with regard to extended interim storage N2 - This presentation addresses safety aspects concerning dual purpose casks for transportation and storage of spent fuel and high level radioactive waste. The long term performance of casks and their safety relevant components like sealed lid systems and the long term performance of cask internals, especially spent fuel assemblies are discussed. Transportation after interim storage is another key issue and due to the fact of delayed disposal projects current interim storage periods need to be extended causing additional safety demonstration needs for longer periods of time. Regarding this situation present BAM research activities in this area and their preliminary outcomes are presented. T2 - 5. RAM-Behältersicherheitstage CY - Berlin DA - 16.03.2016 KW - Interim storage KW - Radioactive waste KW - Spent fuel KW - Safety PY - 2016 AN - OPUS4-35615 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Erhard, Anton A1 - Völzke, Holger A1 - Droste, Bernhard A1 - Wolff, Dietmar A1 - Wille, Frank T1 - Nuclear power plant waste management strategy in Germany N2 - The shut-down of eight nuclear power plants and the phase out of the nuclear energy genera� tion in Germany, due to a decision of the German government after the Fukushima accident, has rapidly forced the need for transportation and storage of spent nuclear fuel (SNF) from shut-down reactors in this country. In accordance with the present strategy in Germany, which is characterized by searching for appropriate final repository areas, only the dry interim stor� age facilities are available for storage nuclear waste of the decommissioned plants as well as the operating plants. Therefore, the transportation and the storage casks must be available for both types of light water reactors. But for the time being this is the bottleneck. Unfortunately the prognoses are neglecting the fact, that such a high number of casks are needed in a very short space of time. This requires the acceleration of formal activeness in relation to the li� censes but also an increase of the production. In the meanwhile to get this process started, the only possibility is the storage in the pools at the plant site. Anyway, for the time being, this is one of the problems for dismantling the eight plants which have been shut down due to the decision of the Government, which is now fixed in the last amended German Atomic Energy Act dated 24th of February 2012. Starting to dismantle in the radioactive regions of the plants requires empty pools. For emptying the pools of damaged SNF special procedures regarding the transportation and storage are necessary. The transportation of damaged SNF requires special casks or at least special procedures and therefore also a special licenses. This paper describes the challenge for the management in transportation and storage of SNF in Germany after the shut-down of the nuclear power plants by reason of the Fukushima acci� dent. T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 PY - 2013 SP - 1 EP - 11 PB - Omnipress AN - OPUS4-30135 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger A1 - Wolff, Dietmar A1 - Probst, Ulrich A1 - Nagelschmidt, Sven A1 - Schulz, Sebastian T1 - Long-term performance of metal seals for transport and storage casks N2 - Dual purpose casks for the transportation and storage of spent nuclear fuel and other radioactive materials require very high leak-tightness of lid closure Systems under accident conditions as well as in the long-term to prevent activity release. For that purpose metal seals of specific types with an inner helical spring and outer metal liners are widely used and have shown their excellent performance if certain quality assurance requirements for fabrication and assembling are satisfied. Well defined surface roughness, clean and dry inert conditions are therefore essential. No seal failure in a loaded cask happened under these conditions until today. Nevertheless, the considered and licensed Operation period is limited and all safety assessments have been performed and approved for this period of time which is 40 years in Germany so far. But in the meantime longer storage periods might be necessary for the future and therefore additional material data will be required. BAM is involved in the qualification and evaluation procedures of those seals from the early beginning. Because long-term tests are always time consuming BAM has early decided to perform additional tests with specific test seal configurations to gain a better understanding of the long-term behavior with regard to seal pressure force, leakage rate and useable resilience which is safety relevant mainly in case of accidental mechanical loads inside a storage facility or during a subsequent transport. Main test parameters are the material of the outer seal jacket (silver or aluminum) and the temperature. This paper presents the BAM test program including an innovative test mock-up and most recent test results. Based on these data extrapolation models to extended time periods are discussed, and also future plans to continue tests and to investigate seal behavior for additional test parameters are explained. T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 PY - 2013 SP - 1 EP - 13 PB - Omnipress AN - OPUS4-30136 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Wolff, Dietmar A1 - Droste, Bernhard A1 - Völzke, Holger A1 - Baden, M. T1 - German approach and experience feedback of transport ability of SNF packages after interim storage N2 - In Germany the concept of dry interim storage of spent fuel in dual purpose metal casks is implemented, currently for periods of up to 40 years. The casks being used have an approved package design in accordance with the international transport regulations. The license for dry storage is granted on the German Atomic Energy Act with respect to the recently revised “Guidelines for dry interim storage of irradiated fuel assemblies and heat-generating radioactive waste in casks” by the German Waste management Commission (ESK). For transport on public routes between or after long term interim storage periods, it has to be ensured that the transport and storage casks fulfil the specifications of the transport approval or other sufficient properties which satisfy the proofs for the compliance of the safety objectives at that time. In recent years the validation period of transport approval certificates for manufactured, loaded and stored packages were discussed among authorities and applicants. A case dependant system of 3, 5 and 10 years was established. There are consequences for the safety cases in the Package Design Safety Report including evaluation of long term behavior of components and specific operating procedures of the package. Present research and knowledge concerning the long term behavior of transport and storage cask components have to be consulted as well as experiences from interim cask storage operations. Challenges in the safety assessment are e.g. the behavior of aged metal and elastomeric gaskets under IAEA test conditions to ensure that the results of drop tests can be transferred to the compliance of the safety objectives at the time of transport after the interim storage period. T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 KW - Storage KW - Transport KW - Lagerung KW - Verpackung KW - Radioactive material PY - 2013 SP - 1 EP - 9 PB - Omnipress AN - OPUS4-30159 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Wille, Frank A1 - Wolff, Dietmar A1 - Droste, Bernhard A1 - Völzke, Holger A1 - Baden, M. T1 - German approach and feedback on experience of transportability of SNF packages after interim storage N2 - In Germany, the concept of dry interim storage of spent fuel in dual purpose metal casks is implemented, currently for periods of up to 40 years. The casks being used have an approved package design in accordance with the international transport regulations. The license for dry storage is granted on the German Atomic Energy Act with respect to the recently revised 'Guidelines for dry interim storage of irradiated fuel assemblies and heat-generating radioactive waste in casks' by the German Waste management Commission. For transport on public routes between or after long term interim storage periods, it has to be ensured that the transport and storage casks fulfil the specifications of the transport approval or other sufficient properties, which satisfy the proofs for the compliance of the safety objectives at that time. In recent years, the validation period of transport approval certificates for manufactured, loaded and stored packages were discussed among authorities and applicants. A case dependent system of 3, 5 and 10 years was established. There are consequences for the safety cases in the Package Design Safety Report, including evaluation of long term behaviour of components and specific operating procedures of the package. The present research and knowledge concerning the long term behaviour of transport and storage cask components have to be consulted as well as experiences from interim cask storage operations. Challenges in the safety assessment are e.g. the behaviour of aged metal and elastomeric gaskets under IAEA test conditions to ensure that the results of drop tests can be transferred to the compliance of the safety objectives at the time of transport after the interim storage period. Assessment methods for the material compatibility, the behaviour of fuel assemblies and the aging behaviour of shielding parts are issues as well. This paper describes the state of the art technology in Germany, explains recent experience on transport preparation after interim storage and points out arising prospective challenges. KW - Radioactive material storage KW - Dry storage KW - Interim storage KW - Cask design KW - Radioactive material transport KW - Regulation KW - Operating procedures KW - Package KW - Safety KW - Storage KW - Transport KW - Lagerung KW - Spent fuel PY - 2014 DO - https://doi.org/10.1179/1746510914Y.0000000064 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 25 IS - 2 SP - 55 EP - 59 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-33834 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kasparek, Eva Maria A1 - Völzke, Holger T1 - Further research on cask impact limiter needed? N2 - Transport and storage casks for medium and high level radioactive waste are subjected to extreme heavy loads during the accidental drop scenarios prescribed by IAEA regulations and national storage acceptance criteria. Subsequently, considerable efforts have been made to optimize impact limiting structures aiming to reduce cask stresses and deformations. Though, their benefits can only fully be exploited, if the energy absorbing behavior is reliable simulated by numerical models which play an increasingly important role in safety assessments. BAM has had conducted the 5-years research project ENREA in order to overcome the problems caused by the use of simplified, not sufficiently validated or even defective computational concepts for damping materials. While major results of the just recently completed program will be presented, the focus is laid on the question whether these Outputs provide a solid foundation for approximating impact limiting structures in all relevant configurations. So far, material models have been implemented for spruce, Polyurethane foams and damping concrete based mainly on small and medium scale compression test results. The experimental program performed at BAM comprised likewise guided drop tests and, especially for concrete, penetration tests. Although the resulting numerical simulations showed good agreement with measured values, it turned out to be necessary to consider further effects with regard to e.g. multiaxial stress States, shearfailures of shock absorbing components and their interaction with enclosed or enclosing structures. Thus, the strengths and application constraints of the actual models will be shown as well as details about further research needs, if any. T2 - RAMTRANSPORT 2015 - International conference on the radioactive materials transport and storage CY - Oxford, UK DA - 19.05.2015 KW - Numerical model KW - Shock absorber KW - Damping concrete KW - Polyurethane foam KW - Radioactive waste storage cask PY - 2015 SP - 1 EP - 8 AN - OPUS4-33992 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger T1 - Extending interim storage of spent nuclear fuel: what are the needs and consequences? N2 - This presentation addresses the current German policy in nuclear waste management and illustrates significant changes throughout the last decades. With the reset of the high level waste repository siting process in 2013 the need for extending interim storage beyond initial license periods will become a future challenge. For that purpose knowledge needs to be enhanced and data gaps need to be closed timely. In addition, international collaboration in R&D and regulatory improvements are explained. T2 - NDC2016 7th Annual Nuclear Decommissioning Conference CY - Manchester, UK DA - 31.05.2016 KW - Nuclear waste management KW - Diposal KW - Storage cask KW - Safety KW - Extended storage PY - 2016 AN - OPUS4-36353 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger T1 - Integration across storage, transportation and disposal – important elements of the German program N2 - This presentation addresses the following issues. Unexpected changes in national nuclear policy have significantly affected spent fuel management strategies. The initial German spent fuel and HLW disposal concept has been delayed and finally “reset” by federal law in 2013. Repository site selection criteria and procedures will be defined by the Disposal Commission’s report end of June 2016. Final site selection should be finished by 2031. The consistent German concept of dry interim storage in dual purpose casks has proven to be a safe and secure spent fuel management strategy. The need for the extended interim storage of spent fuel and HLW casks towards disposal implicates additional challenges for the nuclear waste management strategy in Germany. Recently, governmental research programs have been adjusted to address technical and scientific issues also in the predisposal area. Interim storage, subsequent transportation, and final disposal are closely linked. Integrated approaches concerning waste package designs and operations are supposed to be beneficial for the establishment of efficient long-term spent fuel and HLW management strategies. T2 - Reset of U.S. Nuclear Waste Management Strategy and Policy Meeting CY - Washington, D.C., USA DA - 17.05.2016 KW - Nuclear waste management KW - spent fuel storage KW - extended storage KW - nuclear waste disposal PY - 2016 AN - OPUS4-36354 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger 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 - Zencker, Uwe A1 - Gaddampally, Mohan Reddy A1 - Völzke, Holger T1 - Fracture Mechanics Analysis of Spent Fuel Claddings during Long-Term Dry Interim Storage N2 - The prevention of brittle fracture of spent fuel claddings during long-term dry interim storage is based on experimental investigations, numerical analyses and assessment methods for predicting the mechanical behavior and determining limiting conditions. The ring compression test (RCT) is an established experimental method for characterizing cladding material. Test results for various high-burnup pressure water reactor zirconium-based fuel cladding alloys (e.g., ZIRLO®, M5®) are publicly available. To reduce the effort associated with irradiated samples in hot cells, it is helpful to perform studies on unirradiated surrogate cladding material. Based on such experimental data, load-displacement curves were numerically analyzed for selected cladding materials. In the presence of radial hydrides, a sample may suddenly fail by fracture even at small deformations. Noticeable load drops in the RCT occur associated to unstable crack propagation through the radial hydride network. The failure mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth, and coalescence in the zirconium matrix, with ductile tearing patches connecting neighboring hydrides. The failure process was simulated by cohesive zones controlled by the fracture energy and the cohesive strength. A modeling approach is presented in which the radial hydride morphology is taken into account. Based on the developed fracture mechanics approach with cohesive zone modeling, not only the deformation behavior but also the failure behavior of irradiated as well as unirradiated claddings with radial hydrides under RCT loading conditions can be adequately described. T2 - 20th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 2022) CY - Juan-les-Pins, France DA - 11.06.2023 KW - Extended Interim Storage KW - Nuclear Fuel Cladding KW - Numerical Failure Analysis KW - Ring Compression Test KW - Spent Fuel Characterization PY - 2023 SP - 1 EP - 8 AN - OPUS4-59146 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Völzke, Holger A1 - Zencker, Uwe A1 - Herz, A. A1 - Kreienmeyer, M. T1 - Design of a Drop Test Target with Reproducible Properties for Konrad Prototype Testing N2 - Beginning in 1976 the former iron ore mine Konrad was geologically investigated concerning its suitability as a location for a German final repository for low and intermediate level waste. After a comprehensive licensing procedure it was approved and is now being prepared and equipped for the planned start of storage in 2027. At the end of the 1980s, the requirements for waste containers for storage in this repository were defined for the first time on the basis of an incident analysis of the on-site handling procedures, beginning with the unloading of the packages after the arrival at the site up to the final positioning in the underground storage galleries. A 5-meter drop onto the rock ground of the mine was identified as the covering case for high mechanical requirements (ABK II container class). In contrast to the 9-meter drop according to the IAEA Regulations for the Safe Transport of Radioactive Material, the 5-meter drop is performed onto a hard but not essentially unyielding target. The container is typically not protected by an impact limiter. The requirements for a potential test facility are described in the regulations for the Konrad repository. The mechanical strength of the target is defined as a concrete strength equivalent to the identified properties of the rock ground of the mine. Since 1991 BAM has consistently used precast reinforced concrete slabs as target for drop tests in the framework of licensing procedures as well as in research projects. While the original design fulfilled the requirement for the integrity of the concrete slab in most cases, it failed when drop tests onto an edge of a container were performed. A redesigned concrete slab developed in a research project and suggested as a reference target in 2009 has been successfully used in Konrad licensing procedures since then. The paper gives a brief overview of the historical development and it describes BAM’s efforts and approaches to continuously guarantee a concrete slab of defined quality and to provide a test setup for valid drop tests from the applicants and the authorities view. T2 - PATRAM 2022, 20th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Juan-les-Pins, France DA - 11.06.2023 KW - Target KW - Drop test KW - Final repository Konrad PY - 2023 SP - 1 EP - 11 AN - OPUS4-58563 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger 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 - Zencker, Uwe A1 - Völzke, Holger T1 - Brittle Failure Limits of Spent Fuel Claddings Subjected to Long-Term Dry Interim Storage Conditions N2 - The mechanical properties of spent fuel claddings can be adversely affected under the conditions of long-term dry interim storage, so that the failure limits may be reached in case of mechanical loads during handling or transport after storage. Pre-storage drying and the early stage of interim storage can expose the cladding to higher temperatures and higher tensile hoop stresses than those associated with in-reactor operation and pool storage. During slow cooling of a cladding tube under internal pressure, radial hydrides may precipitate in zirconium-based cladding alloys. This can lead to embrittlement of the material and sudden failure of the cladding integrity under mechanical stress. In order to prevent brittle failure, numerical methods are being developed to predict the mechanical behaviour and identify limiting conditions. Experimental investigations, numerical analyses and evaluation methods are discussed. An established experimental method for characterising cladding materials is the Ring Compression Test (RCT). Some test results on irradiated cladding tubes after operation in pressure water reactors are publicly available. However, it is helpful to carry out studies on unirradiated surrogate claddings with similar material properties to reduce the effort associated with irradiated samples in hot cells and to perform material tests with a wider range of parameters. On the basis of such experimental data, load-displacement curves have been numerically analysed for a selection of cladding materials. Radial hydrides can cause a sample to break suddenly due to fracture even at low deformation. Noticeable load drops in the RCT are caused by unstable crack propagation through the radial hydride network. The failure mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth, and coalescence in the zirconium matrix, with ductile tearing patches connecting adjacent hydrides. The cohesive zone approach was used to simulate the failure process taking into account the radial hydride morphology. The developed method can adequately describe both the deformation and failure behaviour of irradiated as well as unirradiated claddings of zirconium-based alloys with radial hydrides under RCT conditions. Limiting conditions can be expressed in terms of fracture energy and cohesive strength. T2 - IAEA International Conference on the Management of Spent Fuel from Nuclear Power Plants - Meeting the Moment CY - Vienna, Austria DA - 10.06.2024 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2024 SP - 1 EP - 10 AN - OPUS4-60323 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - 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 - Nagelschmidt, Sven A1 - Probst, Ulrich A1 - Völzke, Holger A1 - Wolff, Dietmar T1 - Long-term investigations of metal seals for storage casks of radioctive materials N2 - The Bundesanstalt für Materialforschung und –prüfung (BAM) is a federal institute for materials research and testing in Germany and has been involved in the qualification and safety evaluation procedures of metal seals from the early beginning of the interim storage licensing procedures for radioactive materials, stored in dual purpose casks. Regarding this subject, BAM investigates the long-term behavior of metal seals under the influence of temperature using experimental data and analytical approaches. The development of numerical models is in progress as well. Systematic experimental investigations performed by BAM indicate a continuous decrease of the remaining seal force and the usable resilience considering the leak tightness. Hence, there is a fundamental interest of describing time and temperature dependency to gain predictable values for the long-term behavior and to achieve reliable results with help of short-term tests. The paper gives an overview about the sealing principle, test program and test results of metal seals of the type HELICOFLEX® HN200. The aging effect, respectively the long-term behavior in dependency of time and temperature, are introduced for two different outer liner materials, aluminum and silver. T2 - PVP ASME Conference CY - Vancouver, Canada DA - 17.07.2016 KW - Long-Term Behaviour KW - Metal Seal KW - Metal Gasket PY - 2016 SN - 978-0-7918-5045-9 VL - 7 SP - 63596-1 EP - 63596-5 AN - OPUS4-38039 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger A1 - Orellana Pérez, Teresa T1 - Latest developments in Germany concerning extended interim storage N2 - The presentation explains recent administrative developments in Germany concerning nuclear waste management and adresses national and European R&D funding concepts. National and international collaboration is explained followed by the latest update on long term metal seal investigations at BAM. Finally, fuel cladding failure mechanisms during long term dry storage and related R&D initiatives at BAM are explained. T2 - EPRI Extended Storage Collaboration Program (ESCP) International Subcommittee Meeting CY - Jeju, Korea DA - 15.09.2017 KW - Nuclear waste management KW - Long term storage KW - Spent fuel PY - 2017 AN - OPUS4-43793 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger T1 - Update on long-term investigations of metal seals at BAM and future perspectives of spent fuel management in Germany N2 - The presentation focuses on the latest developments of the German nuclear waste management policy including restructering of responsible organizations and the restart of the repository site selection process for high level radioactive waste. The second part illustrates latest results of long term investigations of metal seals at BAM with regard to extended interim storage periods. T2 - EPRI Extended Storage Collaboration Program (ESCP) Winter Meeting 2017 CY - Charlotte, NC, USA DA - 14.11.2017 KW - Spent fuel KW - Metal seal KW - Extended interim storage PY - 2017 AN - OPUS4-43796 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger 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 -