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 - 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 - 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 - 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 - 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 - 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 - JOUR A1 - Nagel, Thomas A1 - Rühaak, Wolfram A1 - Amann, Florian A1 - Bracke, Guido A1 - Buske, Stefan A1 - Kowalski, Julia A1 - Reiche, Sönke A1 - Schäfer, Thorsten A1 - Scheytt, Traugott A1 - Stumpf, Thorsten A1 - Völzke, Holger A1 - Wellmann, Florian A1 - Kolditz, Olaf T1 - Deep geological disposal - Editorial to the topical collection N2 - The subsurface geological environment, historically a key source of fossil and mineral resources, has seen a significant broadening of its potential applications in recent years, with increasing emphasis on its use for energy production and storage. A critical factor in utilizing subsurface space is ensuring the sustainability of management practices, with particular attention to the potential environmental impacts that may result. The current Topical Collection on Deep Geological Disposal places greater emphasis on the geotechnical aspects of geological disposal, with a particular focus on the safe isolation of hazardous materials, including radioactive waste. The Topical Collection on Deep Geological Disposal covers both experimental (laboratory and in-situ) and theoretical studies of various geotechnical applications, such as geological disposal of radioactive or other materials, e.g. coal-based solid waste, carbon dioxide sequestration in fractured reservoirs, and brine solutions, but also aspects of uranium mining. The articles in the Topical Collections are arranged under the following headings: – Fundamental studies – Experimental studies – Numerical studies KW - Disposal KW - Radioactive Waste KW - Geology PY - 2025 UR - https://nbn-resolving.org/urn:nbn:de:kobv:b43-624955 DO - https://doi.org/10.1007/s12665-024-12075-2 SN - 1866-6299 VL - 84 IS - 3 SP - 1 EP - 6 PB - Springer Nature AN - OPUS4-62495 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 generical cylindrical concrete container during mechanical specimen tests N2 - By highest court decision in 2007 the issued license for the Konrad repository (a former iron ore mine) 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 by 2029. The Bundesanstalt fuer Materialforschung und -pruefung (BAM, Federal Institute for Materials Research and Testing) is regularly contracted by the Bundesgesellschaft für Endlagerung (BGE, Federal Company for Radioactive Waste Disposal) 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 quality assurance measures for container manufacturing. In the written Konrad disposal requirements all acceptance criteria for potential final disposal containers are defined. In addition to general requirements concerning container design, materials, dimensions and others, specific mechanical load scenarios the casks have to withstand are derived from an analysis of the on-site handling and emplacement procedures. Drop tests from different heights under consideration of the most damaging drop orientation and a stacking test are two examples for safety demonstrations that have to be performed during the container design testing procedure to qualify the respective container type. The containers, whether cubic or cylindrical containers, can be made from different materials like ductile cast iron, reinforced concrete or as welded construction from steel plates and profiles. In case of concrete containers 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 waste containers manufactured already decades ago, the reinforcement cages are not always exactly, axisymmetrically 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 and pure elastic material properties. As outcome the stress conditions in the container during the drop test impact under different drop positions and during the stacking test were investigated and are explained and illustrated in this contribution. Different reinforcement cage positions were calculated and compared. 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 - ASME 2024 Pressure Vessels & Piping Conference CY - Bellevue, Washington, USA DA - 28.07.2024 KW - Container design testing procedure KW - Reinforced concrete containers KW - Drop test KW - Stacking test KW - Imperfectly mounted reinforcement cage KW - Finite-element-analysis PY - 2024 SN - 978-0-7918-8851-3 DO - https://doi.org/10.1115/PVP2024-123797 VL - 5 SP - 1 EP - 8 PB - ASME CY - New York City, USA AN - OPUS4-62558 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger A1 - Liu, Y. T1 - Collaboration on aging management for extended storage of used fuel N2 - A Memorandum of Understanding (MOU) between the Federal Institute for Materials Research and Testing (BAM) and Argonne National Laboratory (Argonne) was signed in the fall of 2014. Its objectives are to promote cooperation among scientists and specialists at BAM and Argonne and establish a framework for collaboration in the field of advanced technologies associated with the back end of the nuclear fuel cycle. Collaborative activities involving the BAM and Argonne may be implemented through the promotion of joint research activities and scientific workshops and conferences; the exchange of technical information; and visits by scientists, specialists, and graduate, postgraduate, and Ph.D. students. The first Argonne/BAM workshop was held at Argonne in October 2014 to discuss aging management issues related to the long-term dry storage of spent fuel. Major topics for potential collaboration on the extended storage of spent nuclear fuel and its subsequent transportation were identified. T2 - IHLRWM 2015 - 15th International high-level radioactive waste management conference CY - Charleston, SC, USA DA - 12.04.2015 KW - Aging management KW - Fuel KW - Extended storage KW - Nuclear PY - 2015 SN - 978-0-89448-719-4 SP - 814 EP - 817 AN - OPUS4-33115 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Scheidemann, Robert A1 - Kasparek, Eva Maria A1 - Qiao, Linan A1 - Müller, Karsten A1 - Droste, Bernhard A1 - Völzke, Holger T1 - Dynamic penetration tests on shock absorbing damping concrete N2 - Mechanical loading conditions of transport and storage casks for radioactive materials in accidental scenarios are highly affected by the behavior of both: the impact limiters and the footing materials. To minimize potential damages during the handling of casks, a so called damping concrete is frequently used for the footings in interim nuclear facilities. It obtains its shock absorbing properties through admixing of polymer cells. For a comprehensive mechanical evaluation of casks, advanced material models are also needed for damping concrete. In order to characterize the mechanical properties and to develop numerical material models, penetration tests were carried out at different test facilities of BAM. The tests contain dynamic penetration tests on mortared specimens with a size of 2,400 x 2,400 x 500 mm³. For these model-sized penetration tests indenters with different geometries and diameters were used. Subsequently a cylindrical cast-iron indenter with a diameter of 1.100 mm was dropped of 5 m height on a realistic damping concrete footing. T2 - WM2014 Conference CY - Phoenix, Arizona, USA DA - 02.03.2014 KW - Drop test KW - Dynamic KW - Penetration KW - Damping concrete PY - 2014 SN - 978-0-9836186-3-8 SP - Paper 14166, 1 EP - 9 AN - OPUS4-31948 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger T1 - The German aging management approach for dry spent fuel storage in dual purpose casks N2 - Since the decision by the German government to face out nuclear electricity generation the total amount of spent nuclear fuel and high level wastes from reprocessing is limited and well determined. In addition the siting and licensing procedure to establish a final repository has been ruled by a new law in the mid of 2013 and further delays are very likely until a deep geological repository may start its operation. In the meantime dry interim storage in dual purpose casks being permanently certified for interim storage as well as transportation is the established technical solution. Several on-site as well former centralized facilities are operated successfully for many years but storage licenses are generally limited to 40 years and future lifetime extensions are predictable. Permanent aging management for storage facilities and casks is necessary to demonstrate compliance with safety requirements and furthermore to gain relevant data and information about the technical conditions of the facilities and their components for future lifetime extensions. For that reason procedures and measures are currently improved and the approach is explained in this paper. In addition, the current status and latest experiences concerning periodic safety inspections and aging management measures are discussed. T2 - PSAM 12 - Probabilistic safety assessment and management CY - Honolulu, Hawaii, USA DA - 22.06.2014 KW - Interim storage KW - Dual purpose casks KW - Spent nuclear fuel KW - Aging management PY - 2014 SP - T06 - Doc.183, 1-7 AN - OPUS4-32250 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR 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. However, 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 behaviour 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 aluminium) 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 behaviour for additional test parameters are explained. KW - Metal Seal KW - Spent Fuel KW - Storage KW - Ageing KW - Cask KW - Leak tightness PY - 2014 DO - https://doi.org/10.1179/1746510914Y.0000000057 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 25 IS - 1 SP - 8 EP - 15 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-32327 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Zencker, Uwe A1 - Wolff, Dietmar A1 - Völzke, Holger A1 - Herbrich, Uwe A1 - Nagelschmidt, Sven T1 - Numerical analysis of cask accident scenarios in storage facilities N2 - Existing spent nuclear fuel (SF) and high active waste (HAW) management policies and practices worldwide are the result of past presumptions that sufficient reprocessing and/or disposal capacity would be available in the near term. Consequently, in the past many countries have developed specific solutions for different periods of time due to their individual national nuclear policies. In Germany the concept of dry interim storage in dual purpose metal casks before disposal is being pursued for SF and HAW management and transport and storage licenses have been issued accordingly. The current operation licenses for existing storage facilities have been granted for a storage period of up to 40 years. This concept has demonstrated its suitability for over 20 years so far. Relevant safety requirements have been assessed for the short-term as well as for the long-term for site-specific operational and accidental storage conditions. But in the meantime significant delays in the national repository siting procedure occurred which will make extended storage periods necessary in the future. T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 KW - Numerical analysis KW - Interim storage KW - Cask accident scenario PY - 2013 SP - Session H, Paper 202, 1 EP - 9 PB - Omnipress AN - OPUS4-30227 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Scheidemann, Robert A1 - Kasparek, Eva Maria A1 - Müller, Karsten A1 - Droste, Bernhard A1 - Völzke, Holger T1 - Model-sized and full-scale dynamic penetration tests on damping concrete N2 - Mechanical loading conditions of transport and storage casks for radioactive materials in accidental scenarios are highly affected by the behavior of both: the impact limiters and the footing materials. To minimize potential damages during the handling of casks, a so called damping concrete is frequently used for the footings in interim nuclear facilities. It obtains its shock absorbing properties through admixing of polymer cells. For a comprehensive mechanical evaluation of casks, advanced material models are also needed for damping concrete. In order to characterize the mechanical properties and to develop numerical material models, penetration tests were carried out at different test facilities of BAM. The tests contain static and dynamic penetration tests on cubic specimen with an edge length of 100 mm as well as mortared specimen with a size of 240 x 240 x 50 cm³. Indenters with different geometries and diameters were used for these model-sized penetration tests. Subsequently a full-scale cylindrical cast-iron indenter with a diameter of 110 cm was dropped of 5 m height on a realistic damping concrete footing. T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 KW - Drop test KW - Dynamic KW - Penetration KW - Damping concrete PY - 2013 SP - 1 EP - 10 PB - Omnipress AN - OPUS4-30233 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Botsch, W. A1 - Smalian, S. A1 - Hinterding, P. A1 - Völzke, Holger A1 - Wolff, Dietmar A1 - Kasparek, Eva Maria T1 - Dual purpose cask safety evaluation - Expertise in dry storage N2 - In Germany dual purpose casks for spent nuclear fuel (SF) or high level radioactive waste (HLW) are used for safe transportation and interim storage. Key safety issues in both fields are the safe enclosure of the radioactive material, the safe removal of decay heat, securing nuclear criticality safety, limitation of radiation exposure to acceptable levels and keeping it as low as reasonable achievable (ALARA principle). Whereas these safety requirements during transportation are ensured by the Type-B(U)-design testing and approval procedure on basis of internationally agreed IAEA (International Atomic Energy Agency) requirements storage licenses are issued on national and site specific safety analyses. This paper presents experiences from recent interim storage safety evaluation and licensing procedures in Germany on basis of a Type-B(U) certified cask designs concerning specific differences from Operation and accident conditions inside the storage facility. The focus is laid on the interaction of cask and building structures with regard to shielding, heat removal and accident analyses including aircraft crash. Basic safety is assured by thick-walled metal casks with monitored double lid Systems. They also reduce radiation to levels where workers can safely operate and maintain the casks inside the storage facility. The storage building e. g. provides additional shielding and protection against extemal hazards depending on the building construction. In addition, the storage building helps to reduce radiation levels at the boundary of the storage site well below regulatory limits given by the German Radiation Protection Ordinance (StrSchV). T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 KW - Dual purpose cask KW - Long-term storage PY - 2013 SP - 1 EP - 8 PB - Omnipress AN - OPUS4-30235 LA - eng 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 - Weber, Mike A1 - Zencker, Uwe A1 - Wolff, Dietmar A1 - Völzke, Holger A1 - Herbrich, Uwe A1 - Nagelschmidt, Sven T1 - Numerical analysis of cask accident scenarios in storage facilities N2 - Mechanical drop test scenarios for Type B (U) packages according to the IAEA regulations have to be carried out onto the so-called “unyielding target” (usually with cask impact limiters) and onto the puncture bar respectively. They are predefined and do not require any further investigation of scenarios that really could happen on transportation routes. Cask accident scenarios in the framework of approval procedures for interim storage sites are derived from a detailed analysis of the handling procedures necessary from arrival of cask at the site to its storing position. In that case, casks are usually handled without impact limiters. Dependent on possible drop heights, drop positions and floor properties, conservative cask accident scenarios are derived for further safety proofs. According to the mechanical assessment concept of the considered approval procedure numerical calculations have to be provided by the applicant to demonstrate mechanical cask safety. Stresses and strains in the cask body as well as in the lid System have to be identified and assessed. Using the example of a 3-mvertical-drop of a transport and storage cask for spent fuel elements onto the floor construction made of damping concrete covered by screed, BAM developed a finite element model. The finite element code ABAQUS/Explicit™ was used. Results of experimental investigations are not available. Therefore parameter studies are necessary to identify the sensitivity of the finite element model to significant Parameters and to verify the finite element models according to the requirements of the Guidelines for the Numerical Safety Analyses for the Approval of Transport and Storage Casks for Radioactive Materials (BAM GGR-008). The paper describes the modeling of the material behavior and attachment of bottom side cask components. Questions concerning the modeling of a crack length limiting reinforcement in the screed layer are discussed. The influence of the mesh density of the screed layer and its strength is considered as well. Finally, the developed finite element model can be used for a numerical safety assessment. It can help to understand the complex mechanisms of the interaction between the cask components and floor construction. T2 - WM2014 Conference CY - Phoenix, Arizona, USA DA - 02.03.2014 KW - Numerical analysis KW - Interim storage KW - Cask accident scenario PY - 2014 SN - 978-0-9836186-3-8 SP - Paper 14541, 1 EP - 12 AN - OPUS4-31516 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wolff, Dietmar A1 - Völzke, Holger A1 - Wille, Frank A1 - Droste, Bernhard T1 - Extended storage after long-term storage N2 - Existing spent nuclear fuel (SF) and high active waste (HAW) management policies and practices worldwide are the result of past presumptions that sufficient reprocessing and/or disposal capacity would be available in the near term. Consequently, in the past many countries have developed specific solutions for different periods of time due to their individual national nuclear policies. In Germany the concept of dry interim storage in dual purpose metal casks before disposal is being pursued for SF and HAW management and transport and storage licenses have been issued accordingly. The current operation licenses for existing storage facilities have been granted for a storage period of up to 40 years. This concept has demonstrated its suitability for over 20 years so far. Relevant safety requirements haven been assessed for the short-term as well as for the long-term for site-specific operational and accidental storage conditions. But in the meantime significant delays in the national repository siting procedure occured which will make extended storage periods necessary in the future. This paper describes the current situation in Germany with regard to dry cask storage and focuses on current perspectives considering regulatory, technical, and scientific aspects for storage license renewal. Since there is one case of a storage license limitation to only 20 years due to administrative reasons, first experience is currently gathered in case of an expiring storage license. Subsequent license options have been pursued intensively including the extension of the initial storage license as well as shipping all casks to another storage facility considering an extended storage period at that storage facility as well. All safety relevant aspects have to be reviewed on basis of the current state-of-the-art which might be different from the initial safety demonstrations. That includes new safety assessment standards as well as improved knowledge base. Major issues are e.g. improved accident scenario analyses, assessment methods, and consideration of aging effects from previous operation periods. Inspection programs with respect, e.g. to bolts and seals to verify leak-tightness and lid screw pre-stress have been initiated to demonstrate proper cask conditions for extended storage as well as transportation to another storage facility. T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 KW - Extended storage KW - Long-term storage KW - Dual purpose cask PY - 2013 SP - Paper 184, 1 EP - 12 PB - Omnipress AN - OPUS4-30242 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger A1 - Wolff, Dietmar T1 - Update on the German approach of storing spent nuclear fuel in dual purpose casks until final disposal N2 - The German approach of storing spent nuclear fuel in dual purpose casks until final disposal is based on individually licensed storage sites concerning specific nuclear inventories and dual purpose cask types. The currently approved license period is maximum 40 years. With nuclear policy changes including the phase-out decision of 2011 and a new approach for the HLW repository siting process set up 2013 by law also the spent fuel and HLW management strategy has to be adapted. This includes the timely development of disposal container concepts as part of the overall repository siting process and the need for longer interim storage periods until final disposal. Other aspects affect new regulations about the financing of the nuclear waste management by the industry and the public sector, the restructuring of responsible authorities in 2016/2017 and the adaption of R&D funding programs addressing also pre-disposal (or extended interim storage) issues. BAM as a Federal Research Institute has been responsible for or in charge of container safety Evaluation for transport and storage casks for radioactive waste for several decades and started addressing potential Extended interim storage issues already several years ago. This included participation in international projects at the IAEA or in the U.S. Extended Storage Collaboration Program (ESCP) as well as long term laboratory tests with metal seals, elastomeric seals and polymers for neutron shielding purposes. Test results of BAM are published regularly and gained data provide valuable information about thermo-mechanical and radiological degradation effects as well as the long term Performance of respective cask components. Regarding the reliable extrapolation and prediction of the long term Performance of cask materials and components for time periods of up to 80 or even 100 years the continuation and extension of laboratory tests and the development of analytical and numerical simulation models and tools is essential and under way. 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 - Cask KW - Safety KW - Disposal PY - 2017 SP - 189 EP - 194 AN - OPUS4-40411 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 - 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 - Dittrich, Matthias A1 - Kasparek, Eva-Maria A1 - Völzke, Holger T1 - Legacy waste containers for low and intermediate level radioactive waste in Germany - Actual situation and challenges –17070 N2 - Low and intermediate level radio active waste has been produced in Germany for several decades. While a part of the waste had been deposited in the Morleben repository and the Asse mine which have been operated for just a limited period of time, the major amount is conditioned and packaged in temporarily stored waste containers. Those containers are designated for final disposal in the KONRAD repository being currently under construction and expected to start operation in 2022. As a consequence, an assessment of the waste container documentation and safety is strongly needed to receive an approval for the KONRAD repository. This requalification is challenging because those old containers have to meet the same safety requirements as new containers. T2 - WasteManagement Symposia 2017 CY - Phoenix, USA DA - 05.03.2017 KW - Radioactive waste container KW - Konrad KW - Repository PY - 2017 UR - https://www.xcdsystem.com/wmsym/2017/028.html#17070 SP - Paper 17070, 1 EP - 14 AN - OPUS4-40326 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger A1 - Wolff, Dietmar T1 - Spent Nuclear Fuel and HLW Storage towards Disposal - Challenges and Perspectives of an Integrated Approach - N2 - This paper describes and discusses the major challenges of spent fuel management in Germany after the phase-out of nuclear electricity generation was decided in 2011 and a new repository siting procedure was implemented in 2013. Consequences from those decisions which were legally founded by amendments of the German Atomic Energy Act (AtG) result in the need to transfer all remaining spent fuel from limited reactor operation (last reactor shutdown until the end of 2022) into casks for subsequent dry interim storage on-site. Storage licenses are generally issued site-specific considering specific dual purpose casks (DPC) and their inventories and they are generally limited to 40 years so far. But the need for extending the interim storage period in the future has become obvious. Even though, this may not be an issue to be solved already today questions about additional safety demonstrations will arise as soon as licenses need to be extended. Certainly, these questions will ask for reliable data about the long term performance and safety of structures, systems, and components, e. g. the long term performance of cask components and materials like bolted closure systems including metal seals, or fuel rod behaviour concerning cladding materials under stress and temperature conditions. In case of dual purpose casks for storage and transportation this includes aspects on how to demonstrate transportability during or after several decades of interim storage. Long term investigations often require plenty of time and therefore need to be initiated timely. T2 - PATRAM 2016 The 18th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Kobe, Japan DA - 18.09.2016 KW - Storage KW - Spent Fuel KW - Disposal PY - 2016 UR - http://www.patram2016.org/ SN - 978-4-88898-270-2 SP - Paper 5006 AN - OPUS4-40300 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger A1 - Wolff, Dietmar T1 - Spent Nuclear Fuel and HLW Storage towards Disposal - Challenges and Perspectives of an Integrated Approach - N2 - This paper describes and discusses the major challenges of spent fuel management in Germany after the phase-out of nuclear electricity generation was decided in 2011 and a new repository siting procedure was implemented in 2013. Consequences from those decisions which were legally founded by amendments of the German Atomic Energy Act (AtG) result in the need to transfer all remaining spent fuel from limited reactor operation (last reactor shutdown until the end of 2022) into casks for subsequent dry interim storage on-site. Storage licenses are generally issued site-specific considering specific dual purpose casks (DPC) and their inventories and they are generally limited to 40 years so far. But the need for extending the interim storage period in the future has become obvious. Even though, this may not be an issue to be solved already today questions about additional safety demonstrations will arise as soon as licenses need to be extended. Certainly, these questions will ask for reliable data about the long term performance and safety of structures, systems, and components, e. g. the long term performance of cask components and materials like bolted closure systems including metal seals, or fuel rod behaviour concerning cladding materials under stress and temperature conditions. In case of dual purpose casks for storage and transportation this includes aspects on how to demonstrate transportability during or after several decades of interim storage. Long term investigations often require plenty of time and therefore need to be initiated timely. T2 - PATRAM 2016 The 18th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Kobe, Japan DA - 18.09.2016 KW - Storage KW - Spent Fuel KW - Disposal PY - 2016 AN - OPUS4-40301 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger A1 - Wolff, Dietmar T1 - Update on the German approach of storing spent nuclear fuel in dual purpose casks until final disposal N2 - The safe and secure long term storage of spent nuclear fuel until its final disposal in a deep geological repository is a corner stone of the German nuclear waste management strategy. So far interim storage licenses are limited to 40 years concerning the initial concept of having a repository available until 2035. But in the meantime the exploration of the designated Gorleben salt dome was finally stopped and a reset of the complete site selection process was concluded by the German Federal parliament and subsequently fixed by law. A 33-member “Commission on the storage of highly radioactive materials” was established in 2014, to elaborate a consent based siting process and to develop criteria for the safe disposal of heat generating waste. The disposal commission agreed and published its final report in 2016. As a major consequence, the necessary timeframe for establishing the final repository is going to be exceeded significantly and with that, an extension of the current dry interim storage in dual purpose casks becomes necessary. This paper explains the major actions to be taken in Germany to address all relevant issues for a future extension of storage licenses beyond the initial timeframe of 40 years. T2 - ANS Conference International High-Level Radioactive Waste Management (IHLRWM) 2017 CY - Charlotte, NC, USA DA - 09.04.2017 KW - Spent Fuel KW - Storage KW - Disposal PY - 2017 AN - OPUS4-40303 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger T1 - Preserving competence in nuclear technology - BAM's contributions N2 - With regard to the nuclear phase-out in Germany preserving competence in nuclear technology is a major challenge. This presentation explains how BAM addresses this issue within its specific tasks, expertise and long term experience on nuclear waste package design testing and safety evaluation. T2 - 48th Annual Meeting on Nuclear Technology (AMNT 2017) CY - Berlin, Germany DA - 16.05.2017 KW - Nuclear Technology KW - Safety KW - Waste Packages KW - Design Testing PY - 2017 AN - OPUS4-40306 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger T1 - Perspectives on extending interim storage of spent nuclear fuel (beyond the initial license period) N2 - The presentation adddresses the consequences and challenges of the German nuclear phase-out in conjunction with the restart of siting process for a high-level radioactive waste repository. Concerning spent fuel this will lead to extended interim storage periods for the dry storage casks beyond the initial licensing period of 40 years. For extending storage periods additional safety assessments and data about the long term performance of materials, components and casks will be needed and BAM has already been lauching specific R&D projects addressing identified data gaps. T2 - 8th Annual Nuclear Decommissioning Conference Europe CY - Manchester, UK DA - 24.05.2017 KW - Spent Fuel KW - Storage KW - Disposal PY - 2017 AN - OPUS4-40358 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Scheidemann, Robert A1 - Weber, Mike A1 - Völzke, Holger A1 - Kasparek, Eva Maria ED - Gomes, J.F.Silva ED - Meguid, S.A. T1 - Temperature dependendency of dynamically loaded polyurethane foam N2 - Polyurethane foam used as impact limiter material undergoes high plastic deformations, whereat the resulting stress-strain relations strongly depend on loading speed and temperature. This paper discusses the efforts necessary to develop a reliable numerical foam simulation model focussing on generation and implementation of temperature- dependent yield curves T2 - 6th International Conference on Mechanics and Materials in Design CY - Ponta Delgada, Azores, Portugal DA - 26.07.2015 KW - Impact limiter KW - Polyurethane foam KW - Finite element model KW - Strain rate KW - Temperature dependency PY - 2015 SN - 978-989-98832-3-9 SP - 1043 EP - 1046 PB - INEGI-FEUP AN - OPUS4-38889 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kasparek, Eva A1 - Völzke, Holger T1 - Non-Heat Generating Legacy Waste - New Strategies for Older Vessels N2 - The German repository “Konrad” is approved for the final storage of approximately 300.000 m³ of non-heat generating radioactive waste considering also the expected residues from operation and decommissioning of current power plants. According to the latest survey as constituted end of 2014, a large proportion of this volume already exists, of which 100.288 m³ are fully loaded packages. As prerequisite for their “Konrad” acceptance, the vessels have to be designed against operational and accidental conditions of the site and be manufactured according to adequate quality assurance programs. The respective safety assessment performed by BAM on behalf of the Federal Office for Radiation Protection is a major challenge especially for those casks that had been produced and loaded already some decades ago. Typically, they were manufactured in big numbers over a longer period of time involving a lot of constructional and management changes. This fact in combination with the previously rather low documentation standards often causes relevant uncertainties about the specific physical and geometrical characteristics of the cask materials and components as well as about details concerning their whole assemblage. This paper highlights current assessment strategies for such older vessels including cubic steel sheet containers as well as cylindrical thick-walled casks. As compliance with each “Konrad”-requirement can hardly be shown independently, only integrated approaches promises success: First, a broader concept of data collection has to be implemented that allows closing documentation gaps by linking dispersed and diverse data, e.g. delivery slips and handwritten fabrication records, by including plausibility considerations as well as random inspection programs. Thereby, the relevant properties of all applied casks have to be identified and their potential range should be quantified as accurately as possible. This knowledge is necessary to define one or more representative configurations of the cask as the base for design testing. The actual investigation program must enable to track the influence of deviations and determine safety margins especially in the likely case that not the most conservative setting of all cask features can be tested. The paper evaluates the respective ongoing efforts of industrial applicants and summarizes BAM experiences and associated research projects. T2 - 18th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Kobe, Japan DA - 18.09.2016 KW - Requalification KW - Low and medium level radioactive waste KW - Final disposal container KW - Legacy waste KW - Konrad repository PY - 2016 SP - Paper 1014, 1 EP - 9 AN - OPUS4-38892 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger T1 - Issues on Aging of Spent Fuel Storage Systems N2 - The Bundesanstalt für Materialforschung und –prüfung (BAM) and Sandia National Laboratories (SNL) entered into a Memorandum of Understanding (MOU) in September 2012 to foster technical collaborations in the areas associated with the backend of the commercial nuclear fuel cycle. Specifically, the focus is on packaging, transportation, and storage of commercial spent nuclear fuel. The institutes meet about twice each year, alternating between institutes. This provides the opportunity for staff members from the host organization more exposure to technical issues that are of concern internationally and to collaborate with technical experts working on similar problems. Since 2012, the focus of the meetings has been on technical issues associated with extended dry storage and subsequent transportation of commercial spent fuel. Topics range from hydride effects on cladding integrity, spent fuel response during Normal Conditions of Transport (NCT), finite element analyses of fuel and cask response to accident conditions, bolt and seal behavior over extended periods of time, and corrosion associated with bolts, metallic seals, and stainless steel canisters. This MOU has provided an effective leverage for technical collaboration. For example, SNL is funding (through DOE), Savannah River National Laboratories (SRNL) to look at bolt and seal degradation issues. SRNL has an MOU with BAM to collaborate on bolt and seal degradation during extended storage. Likewise, Sandia and BAM are collaborating with the EC Joint Research Center on an International Nuclear Energy Research Initiative (I-NERI) to investigate spent fuel behavior when subjected to mechanical loadings. This important work will provide insight into failure mechanisms, as well as spent fuel release fractions, given a breach of the cladding wall. This presentation provides an overview of high ranked technical issues associated with extended storage and subsequent transportation, as well as the work underway at BAM and SNL that are addressing these issues. T2 - 7th US/German Workshop on Salt Repository Research, Design, and Operation CY - Washington, DC, USA DA - 07.09.2016 KW - Spent Fuel KW - Ageing KW - Storage PY - 2016 AN - OPUS4-37318 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 -