TY - CONF A1 - Weber, Mike A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Zencker, Uwe T1 - Numerical approach to determine a package dependent bar length for the iaea pin drop test N2 - The Federal Institute for Materials Research and Testing (BAM) is assessing the mechanical and thermal safety performance of packages for the transport of radioactive materials. Drop testing and numerical calculations are usually part of the safety case concepts, where BAM is performing the regulatory tests at their own test facility site. Among other mechanical tests the 1 meter drop onto a steel puncture bar shall be considered for accident safe packages. According to the IAEA regulations “the bar shall be of solid mild steel of circular section, 15.0 ± 0.5 cm in diameter and 20 cm long, unless a longer bar would cause greater damage…”. Particularly with regard to the German transport- and storage cask designs, often made from ductile cast iron, an accurate determination of the puncture bar length to guarantee a load impact covering the worst case scenario can be imperative. If the fracture mechanical proof for the cask material shall be provided by a test, small deviations in the concentrated load applied can be decisive for the question if the cask fails or not. The most damaging puncture bar length can be estimated by iterative procedure in numerical simulations. On the one hand, a sufficient puncture bar length shall guarantee that shock absorbers or other attachments do not prevent or reduce the local load application to the package, on the other hand, a longer and thus less stiff bar causes a smaller maximum contact force. The contrary influence of increasing puncture bar length and increasing effective drop height shall be taken into account if a shock absorber is directly placed in the target area. The paper presents a numerical approach to identify the bar length that causes maximum damage to the package. Using the example of two typical package masses the sensitivity of contact forces and puncture bar deformations to the initial length are calculated and assessed with regard to the international IAEA package safety requirements. T2 - PATRAM 2019 CY - New Orleans, LA, USA DA - 04.08.2019 KW - IAEA KW - 1-m-punch-bar-drop-test KW - Numerical approach KW - Bar length KW - Finite element analysis PY - 2019 UR - https://www.inmm.org/INMM-Resources/Proceedings-Presentations/PATRAM-Proceedings.aspx SP - 1 EP - 10 AN - OPUS4-49016 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Kömmling, Anja A1 - Zaghdoudi, Maha A1 - Jaunich, Matthias A1 - Wolff, Dietmar T1 - Development and application of a finite element model representing the rapid partial release of elastomeric O-ring seals N2 - Due to their high versatility and recovery potential under several load conditions, seals made from elastomers are widely used as barrier seals in containers for low and intermediate level radioactive waste and as auxiliary seals in spent fuel storage and transportation casks (dual purpose casks (DPC)). In DPC, elastomer seals allow leakage rate measurements of metal barrier seals for demonstration of their proper assembly conditions and performance. For spent fuel and high level waste as well as for low and intermediate level waste, long time periods will be required before final disposal will be available. Therefore, the long-term behavior of elastomer seals can be an important issue when the cask’s suitability for an extended interim storage period has to be assessed. During the licensed interim storage duration, the requested leakage rate has to be maintained under any static and dynamic load conditions. A sufficient resilience, the capability of seals to compensate a (rapid) partial release was identified as an important end-of-lifetime criterion. In the framework of extensive aging and testing programs at BAM, specimen made from representative types of elastomer (EPDM, FKM) were aged for several years at different temperatures ranging from 23 °C to 150 °C. One aim of the experiments was to determine an end-of-lifetime criterion correlated to leakage as the point of seal failure. As the seals remained leak-tight under static conditions even after advanced material degradation, a more demanding leakage test set-up was constructed that allowed a rapid partial release of the seal. This device enables a release of the O-ring by approx. 0.2 mm from 25 % to 23 % compression in less than one second during the pressure rise measurement. If the resilience of the seal has decreased past a certain point, the seal cannot follow the decompression fast enough and a leak path can open. Using a finite element (FE) material model derived earlier from different material tests, this process was simulated with a 3D FE model, reproducing the essential characteristics of the test device. The FE-model enables the analyzation of the occurring global mechanisms if a leak path opens. The capability of the seals to follow the lifted flange and the subsequent run of compression force depending on the current state of aging are illustrated. The findings of the simulation are correlated and discussed with respect to the pressure rise measurements. T2 - ASME PVP 2019 CY - San Antonio, TX, USA DA - 14.07.2019 KW - Storage and transportation casks KW - Elastomeric seals KW - Rapid partial release KW - Leakage KW - Numerical simulation PY - 2019 AN - OPUS4-49017 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Zaghdoudi, Maha A1 - Kömmling, Anja A1 - Jaunich, Matthias A1 - Wolff, Dietmar T1 - Scission, cross-linking, and physical relaxation during thermal degradation of elastomers N2 - Elastomers are susceptible to chemical ageing, i.e., scission and cross-linking, at high temperatures. This thermally driven ageing process affects their mechanical properties and leads to limited operating time. Continuous and intermittent stress Relaxation measurements were conducted on ethylene propylene diene rubber (EPDM) and hydrogenated nitrile butadiene rubber (HNBR) samples for different ageing times and an ageing temperature of 125 °C. The contributions of chain scission and cross-linking were analysed for both materials at different ageing states, elucidating the respective ageing mechanisms. Furthermore, compression set experiments were performed under various test conditions. Adopting the two-network model, compression set values were calculated and compared to the measured data. The additional effect of physical processes to scission and cross-linking during a long-term thermal exposure is quantified through the compression set analysis. The characteristic times relative to the degradation processes are also determined. KW - Ageing KW - Scission KW - Cross-linking KW - Compression set KW - Physical relaxation PY - 2019 UR - https://nbn-resolving.org/urn:nbn:de:kobv:b43-486091 DO - https://doi.org/10.3390/polym11081280 SN - 2073-4360 VL - 11 IS - 8 SP - 1280, 1 EP - 12 PB - MDPI AN - OPUS4-48609 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Kömmling, Anja A1 - Jaunich, Matthias A1 - Pourmand, P. A1 - Wolff, Dietmar A1 - Hedenqvist, M. T1 - Analysis of O-ring seal failure under static conditions and determination of end-of-lifetime criterion N2 - Determining a suitable and reliable end-of-lifetime criterion for O-ring seals is an important issue for long-term seal applications. Therefore, seal failure of ethylene propylene diene rubber (EPDM) and hydrogenated nitrile butadiene rubber (HNBR) O-rings aged in the compressed state at 125 °C and at 150 °C for up to 1.5 years was analyzed and investigated under static conditions, using both non-lubricated and lubricated seals. Changes of the material properties were analyzed with dynamic-mechanical analysis and permeability experiments. Indenter modulus measurements were used to investigate DLO effects. It became clear that O-rings can remain leak-tight under static conditions even when material properties have already degraded considerably, especially when adhesion effects are encountered. As a feasible and reliable end-of-lifetime criterion for O-ring seals under static conditions should include a safety margin for slight dimensional changes, a modified leakage test involving a small and rapid partial decompression of the seal was introduced that enabled determining a more realistic but still conservative end-of-lifetime criterion for an EPDM seal. KW - EPDM KW - HNBR KW - Seal failure KW - Leak-tightness KW - DLO KW - Oxygen permability KW - DMA KW - Indenter modulus PY - 2019 UR - https://nbn-resolving.org/urn:nbn:de:kobv:b43-486125 DO - https://doi.org/10.3390/polym11081251 SN - 2073-4360 VL - 11 IS - 8 SP - 1251, 1 EP - 19 PB - MDPI CY - Basel, CH AN - OPUS4-48612 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin A1 - Quercetti, Thomas A1 - Wille, Frank T1 - Experimental investigation of heat flux into a fire reference package and determination of propane gas fire characteristics N2 - In order to investigate the heat fluxes over a wide range of surface temperatures during a propane gas fire test and to get information about local fire impact a fire reference package, representing the outer geometry of a specific type of transport cask for radioactive waste, was designed. A closed steel sheet cylinder with a wall thickness of 10 mm was chosen as fire reference package. The cylinder was filled with refractory insulation material and instrumented with thermocouples distributed all over the cylinder. The local steel sheet temperatures measured allow the determination of local as well as global heat fluxes as a function of time and surface temperature. With this fire reference package three open-air propane gas fire tests were performed at BAM’s open air fire test stand. The flame exposure time period was changed for the different fire tests. Furthermore, the wind conditions changed between and during the tests. Test stand parameters like wind shield location and propane gas volume flow were chosen constant for the three tests. The test results were used to determine the changes of heat flux into the fire reference package in relation to the package surface temperature. This data also allows the calculation of local characteristics of the propane gas fire as there are the flame temperature, the fire convection coefficient and the radiation exchange coefficient in a first approach. The recently conducted tests provide an initial picture of local fire characteristics of the propane gas fire test facility. The test shows that the propane gas fire covers the IAEA-fire over a wide range of surface temperatures with the chosen test stand parameters. T2 - Workshop BAM – Sandia National Laboratories CY - Berlin, Germany DA - 01.07.2019 KW - Thermal testing KW - Fire testing KW - Propane gas fire KW - IAEA KW - Convection KW - Radiation KW - Slug KW - Calorimeter PY - 2019 AN - OPUS4-48597 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Schönfelder, Thorsten A1 - Müller, Lars A1 - Komann, Steffen A1 - Wille, Frank ED - Schönfelder, Thorsten T1 - Design assessment of a dual purpose cask for damaged spent nuclear fuel N2 - German package design approvals were granted recently for dual purpose casks (DPC) intended for loading with encapsulated damaged spent nuclear fuel (DSNF). Comprehensive assessment procedures were carried out by the authority BAM with respect to the mechanical and thermal package design, the activity release of radioactive material and quality assurance aspects for manufacturing and operation of each packaging. The objective of each procedure was to verify the Package Design Safety Report (PDSR) and the relevant guidelines fulfils the requirements of the IAEA regulations. Previous approvals of German SNF package designs consider mainly standard fuel assemblies with defined specifications and properties for transport and interim storage. Due to the nuclear power phase-out in Germany all kinds of SNF, e.g. damaged spent fuel rods shall be packed in DPC now. Therefore specific requirements shall be considered in accordance with international experiences including IAEA technical reports. The main requirement for DSNF is a tight encapsulation with specific defined properties under transport and storage conditions. Due to the interim storage period of currently up to 40 years the encapsulation with DSNF in the casks shall also be long term durable. Thus specific loading and drying procedures are necessary and had to be qualified during the approval process. BAM assessed these drying procedures and could confirm the long-term behaviour of the encapsulation and the suitability of the drying equipment. This special equipment was qualified in a “cold handling”. In addition, it was shown that the behaviour of the test equipment used in the qualification process was comparable with the original equipment, e.g. test fuel rods or test encapsulation. In the development of the drying process, experience was obtained in how to put the requirements of the IAEA regulations and related IAEA technical reports into practice. The paper gives an overview of approval assessment and testing experience made by BAM and point out the main resulting requirements on drying processes for these kinds of encapsulations with DSNF. T2 - Proceedings of the 19th International Symposium on the Packaging and Transportation of Radioactive Materials PATRAM 2019 CY - New Orleans, USA DA - 04.08.2019 KW - Assessment KW - Dual purpose cask KW - Spent nuclear fuel PY - 2019 SP - Paper 1204, 1 EP - 10 AN - OPUS4-48685 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wolff, Dietmar T1 - Ageing management issues for extended SF and HLW dry storage in Germany – Challenges, approach, examples N2 - In this presentation, the history and actual situation of the German nuclear waste management policy of spent fuel (SF) and high-level waste (HLW) is reported. Beside the description of the restructuration of responsible authorities and organizations in Germany, the principle of dry interim storage of SF and HLW in dual purpose casks is explained. The need for extended interim storage towards disposal implicates additional challenges for the nuclear waste management strategy in Germany. (Extended)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 SF and HLW management strategies. Knowledge management and staff recruiting, education and training during phase out and beyond nuclear power plant operation are major issues. T2 - IAEA - Second Coordinate Research Meeting and Consultancy Meeting on Ageing Management Programmes for Spent Fuel Dry Storage Systems CY - Lemont, IL, USA DA - 29.04.2019 KW - Ageing Management KW - Extended Storage KW - Spent Nuclear Fuel PY - 2019 AN - OPUS4-48514 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Qiao, Linan A1 - Keller, Christian A1 - Nagelschmidt, Sven A1 - Wolff, Dietmar T1 - Three-dimensional finite element analysis of O-ring metal seals considering manufacture tolerances N2 - Metal seals are widely used in various industrial branches with severe working conditions (e.g. high pressure, high temperature, corrosion, or radioactive radiation). For example, O-ring metal seals are applied in the closure lid system of transport and storage casks for radioactive materials to guarantee an approved specified leak-tightness and the safe enclosure of the radioactive inventory. Within safety assessments of those casks under normal or accident conditions during transport and long-term interim storage for several decades, numerical simulations of the thermo-mechanical behavior of metal seals by using finite element (FE) analyses are suitable and effective. In general, finite element codes provide the possibility to construct a complex three-dimensional (3D) modelling of metal seals with solid elements, cf., where all components of the metal seals are modeled in detail. This modelling is complex but allows the consideration of underlying physical effects such as elastic-plastic deformation, thermal expansion, creep/relaxation, friction and possible local damage. Therefore, this approach permits the investigation and understanding of the complex behavior of metal seals in detail which can hardly be measured for all seal components. In a prior work, the complex 3D modelling approach was applied to investigate the influence of each seal component on the global seal force by considering varying material properties and different seal diameters. In the present contribution, the influence of manufacturing tolerances is discussed. At first, the results of prior work are summarized. After that, the manufacturing tolerances of the studied O-ring metal seal types are described and the corresponding FE model with all individual components is introduced. Finally, the influence of varying manufacturing tolerances of each seal component on the global seal force is analyzed and discussed. T2 - The 2019 EMI International Conference CY - Lyon, France DA - 3. July 2019 KW - Finite element method KW - Metal seals KW - Material properties KW - Manufacturing tolerances PY - 2019 AN - OPUS4-48448 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Musolff, Andre A1 - Quercetti, Thomas A1 - Müller, Karsten A1 - Bartholmai, Matthias T1 - Analysis of RAM Packages Drop Testing under Kinematic Aspects N2 - BAM is the German Federal Institute for Materials Research and Testing and the competent authority for mechanical and thermal safety assessment of transport packages for spent fuel and high level waste. In context with safety assessment of RAM packages BAM performed numerous drop tests in the last decades. The tests were mostly accompanied by extensive and various measurement techniques especially by instrumented measurements with strain gages and accelerometers. The procedure of drop testing and the resulting measurement analysis are the main methods to evaluate the safety against mechanical test conditions. Measurement techniques are dedicated to answer questions in regard to the structural integrity of a RAM package, the mechanical behaviour of the prototype as well as of its content under impact conditions. Test results like deceleration-time functions constitute a main basis for the validation of assumptions in the safety analysis and for the evaluation of numerical calculations. In this context the adequate selection of accelerometers and measurement systems for the performance of drop tests is important. Therefore it is not only necessary to find suitable positions for the accelerometers at the test specimens, but also to consider technical boundary conditions as e.g. temperature. Accelerometers are widely used for the measuring of motion i.e. velocity or the displacement of the rigid cask body, vibration and shock events. Acceleration measurements as well as their analysis are often very complex and extensive also because they are in turn embedded in complex drop test experiments having to consider difficult boundary conditions as for example very low specimen temperatures, large drop heights and sophisticated drop orientations of the specimen. In every case special instruments and adequate technical equipment is required to accelerations under these and transient shock conditions which are characterized in our case by impact times in the range of a few milliseconds up to perhaps 100 Milliseconds naturally depending on container design and drop test conditions as drop height and target. The paper gives an overview of drop tests under kinematic aspects performed with RAM packages. Furthermore, experimental advancements of accelerometer instrumentation within drop testing, e.g. the characteristics and possibilities of accelerometers, behavior of accelerometers and various influence factors are shown. T2 - International High-Level Radioactive Waste Management 2019 (IHLRWM 2019) CY - Knoxville, TN, USA DA - 14.04.2019 KW - Accelerometer KW - Radioactive KW - Packages KW - Drop test KW - Kineatic aspects KW - Measurement PY - 2019 AN - OPUS4-48039 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Nagelschmidt, Sven A1 - Herbrich, Uwe A1 - Qiao, Linan T1 - Applicability of developed calculation models to predict long-term behavior of metal seals – Necessary scope of tests N2 - The long-term sealing behavior of metal seals, assembled in the lid system of casks for transportation and storage of radioactive materials, has been investigated. For that purpose, phenomenological models describing the time- and temperature dependent behavior have been introduced by BAM since 2016, e.g. by using the time-temperature superposition principle. Results have shown that these models describe the relaxation behavior adequately and are suitable for extrapolations. In this work, the applicability of these models is checked by analyzing the necessary scope of tests, which must be carried out to get sufficient information about the long-term behavior of metal seals based on short-term tests. T2 - Waste Management Conference CY - Phoenix, Arizona, USA DA - 03.03.2019 KW - Metal seal KW - Time-temperature superposition principle KW - Transportation and storage cask KW - Long-term interim storage PY - 2019 SP - 19311, 1 EP - 6 AN - OPUS4-48353 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Reichardt, Adrian A1 - Schubert, Sven T1 - Assessment of ageing effects at BAM N2 - In Germany, spent nuclear fuel and vitrified high active waste is stored in dual purpose casks (DPC) at interim storage facilities. In order to ensure the transportability of the DPC to a final repository in future, the maintenance of the package design approval is realized. Therefore, the assessment of possible ageing effects during interim storage is necessary to ensure an evaluation of the transportability. The lecture presents BAM's current ageing evaluation concept in the field of trasnport law. T2 - IRSN-BAM Symposium - Safety of Transport and Storage Packages CY - Online meeting DA - 19.11.2020 KW - Ageing KW - DPC KW - Interim storage KW - Ageing management PY - 2020 AN - OPUS4-51718 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin A1 - Quercetti, Thomas A1 - Wille, Frank T1 - Outcomes of Three Large-Scale Fire Reference Tests Conducted in Propane Gas Fire Test Facility N2 - Packages for the transport of high-level radioactive material are designed to withstand severe accidents. Hypothetical severe accident conditions are defined in the IAEA Regulations for the Safe Transport of Radioactive Materials. One of these accident conditions is the thermal test, mainly consisting of a 30 minute fully engulfing 800°C pool fire or an equally severe fire test. The heat fluxes into the package depend substantially on the fire characteristics and the surface temperature of the package. Fire tests can be performed at BAM on a propane gas fire test facility. In order to investigate the heat fluxes over a wide range of surface temperatures in this test facility a fire reference package was designed for multiple use. The package represented the outer geometry of a specific transport cask for radioactive waste. The fire reference package is a closed steel sheet cylinder with a wall thickness of 10 mm, a length of 1500 mm and a diameter of 1050 mm. The package was instrumented with thermocouples and filled with heat resistant insulation material. Three open-air fire tests were performed in the BAM propane gas fire test facility. The flames exposure time period varied slightly for the fire tests. The wind direction as well as the wind speed were measured and changed between and during the tests. Test stand parameters such as wind shield location and propane gas volume flow were chosen constant for the three tests. The locally measured fire reference package steel sheet temperatures were used for the calculation of heat fluxes as function of time and surface temperature. The measured temperatures allowed further calculations. In a first approach effective fire characteristics of the propane gas fire, including the flame temperature, the fire convection coefficient and a radiation exchange coefficient mathematically describing the determined average heat flux over the surface temperature were calculated. T2 - Abteilungskolloquium 3.3 CY - Online meeting DA - 03.12.2020 KW - Fire KW - Heat flux KW - Convection coefficient KW - Adiabatic surface temperature KW - IAEA PY - 2020 AN - OPUS4-51778 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Müller, Lars A1 - Scheidemann, Robert A1 - Schönfelder, Thorsten A1 - Komann, Steffen A1 - Wille, Frank T1 - Drop tests assessment of internal shock absorbers for packages loaded with encapsulations for damaged spent nuclear fuel N2 - Damaged spent nuclear fuel (DSNF) can be loaded in German dual-purpose casks (DPC) for transport and interim storage. Encapsulations are needed to guarantee a safe handling and a tight closure, separated from the package enclosure. These encapsulations shall be durable and leak-tight for a long storage period, because they are usually not accessible within periodical inspections of the DPC. Due to the general design of DPCs for standard fuel assemblies, specific requirements have to be considered for the design of encapsulations for DSNF to ensure the loading in existing package designs. Especially the primary lid system of a DPC is designed for maximum loads due to the internal impact of the content during drop test conditions. The main difference of encapsulations for damaged spent nuclear fuel is that they have usually a much higher stiffness than standard fuel assemblies. Therefore the design of an internal shock absorber, e.g. at the head of an encapsulation is required to reduce mechanical loads to the primary lid system during impacts. BAM as part of the German competent authority system is responsible for the safety assessment of the mechanical and thermal package design, the release of radioactive material and the quality assurance of package manufacturing and operation. Concerning the mechanical design of the encapsulation BAM was involved in the comprehensive assessment procedure during the package design approval process. An internal shock absorber was developed by the package designer with numerical analyses and experimental drop tests. Experimental drop tests are needed to cover limiting parameters regarding, e.g. temperature and wall thickness of the shock absorbing element to enable a detailed specification of the whole load-deformation behavior of the encapsulation shock absorber. The paper gives an overview of the assessment work by BAM and points out the main findings which are relevant for an acceptable design of internal shock absorbers. The physical drop tests were planned on the basis of pre-investigations of the applicant concerning shape, dimension and material properties. In advance of the final drop tests the possible internal impact behavior had to be analyzed and the setup of the test facility had to be validated. The planning, performance and evaluation of the final drop tests were witnessed and assessed by BAM. In conclusion it could be approved that the German encapsulation system for damaged spent nuclear fuel with shock absorbing components can be handled similar to standard fuel assemblies in existing package designs. T2 - Pressure Vessels & Piping Conference 2020 CY - Online meeting DA - 03.08.2020 KW - Encapsulations for damaged spent nuclear fuel KW - Drop tests KW - Internal shock absorber KW - Design assessment of RAM packages PY - 2020 SP - 1 EP - 9 AN - OPUS4-51546 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Müller, Lars A1 - Scheidemann, Robert A1 - Schönfelder, Thorsten A1 - Komann, Steffen A1 - Wille, Frank T1 - Drop tests assessment of internal shock absorbers for packages loaded with encapsulations for damaged spent nuclear fuel N2 - Damaged spent nuclear fuel (DSNF) can be loaded in German dual-purpose casks (DPC) for transport and interim storage. Encapsulations are needed to guarantee a safe handling and a tight closure, separated from the package enclosure. These encapsulations shall be durable and leak-tight for a long storage period, because they are usually not accessible within periodical inspections of the DPC. Due to the general design of DPCs for standard fuel assemblies, specific requirements have to be considered for the design of encapsulations for DSNF to ensure the loading in existing package designs. Especially the primary lid system of a DPC is designed for maximum loads due to the internal impact of the content during drop test conditions. The main difference of encapsulations for damaged spent nuclear fuel is that they have usually a much higher stiffness than standard fuel assemblies. Therefore the design of an internal shock absorber, e.g. at the head of an encapsulation is required to reduce mechanical loads to the primary lid system during impacts. BAM as part of the German competent authority system is responsible for the safety assessment of the mechanical and thermal package design, the release of radioactive material and the quality assurance of package manufacturing and operation. Concerning the mechanical design of the encapsulation BAM was involved in the comprehensive assessment procedure during the package design approval process. An internal shock absorber was developed by the package designer with numerical analyses and experimental drop tests. Experimental drop tests are needed to cover limiting parameters regarding, e.g. temperature and wall thickness of the shock absorbing element to enable a detailed specification of the whole load-deformation behavior of the encapsulation shock absorber. The paper gives an overview of the assessment work by BAM and points out the main findings which are relevant for an acceptable design of internal shock absorbers. The physical drop tests were planned on the basis of pre-investigations of the applicant concerning shape, dimension and material properties. In advance of the final drop tests the possible internal impact behavior had to be analyzed and the setup of the test facility had to be validated. The planning, performance and evaluation of the final drop tests were witnessed and assessed by BAM. In conclusion it could be approved that the German encapsulation system for damaged spent nuclear fuel with shock absorbing components can be handled similar to standard fuel assemblies in existing package designs. T2 - Pressure Vessels & Piping Conference 2020 CY - Online meeting DA - 03.08.2020 KW - Drop tests KW - Internal shock absorber KW - Design assessment of RAM packages KW - Encapsulations for damaged spent nuclear fuel PY - 2020 AN - OPUS4-51550 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin A1 - Erenberg, Marina A1 - Nehrig, Marko A1 - Bletzer, Claus A1 - Musolff, André A1 - Wille, Frank T1 - Heat Flux from Wood Filled Impact Limiter under Fire Conditions N2 - Packages for the transport of high-level radioactive material must withstand severe hypothetical accidents. Regulatory test conditions shall cover these severe accident conditions and consist of mechanical tests and a following thermal test. To withstand the mechanical tests heavy weight packages are often designed with impact limiters consisting of wood encapsulated in steel sheets. The thermal test is defined precisely in the IAEA-regulations as a 30 minute fully engulfing 800 °C fire. After the fire phase a pre-damaged impact limiter might continue burning or smouldering and influence the cask thermal behaviour with its energy release. The energy transferred from the impact limiter to the cask is of importance for the safety of transport packages. A full-scale fire test with an impact limiter of 2.3 m in diameter and filled with spruce wood was designed and performed. The impact limiter continued burning for 3 days. Energy transfer and temperature measurements were performed. T2 - 9th International Scientific Conference - wood & fire safety 2020 CY - Online meeting DA - 02.11.2020 KW - Fre KW - Smouldering KW - Wood PY - 2020 AN - OPUS4-51517 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zaghdoudi, Maha A1 - Weber, Mike A1 - Kömmling, Anja A1 - Jaunich, Matthias A1 - Wolff, Dietmar T1 - Numerical Modelling of Compression Stress Relaxation and Compression Set of Elastomer O-Ring During Aging N2 - Elastomer seals are used in many industrial applications due to their excellent mechanical properties at a wide range of temperatures. Their high versatility and recovery potential under several load conditions make them well suitable for the application in containers designed for transport, storage and/or disposal of radioactive materials. In containers for low and intermediate level radioactive waste, elastomer seals are used as barrier seals, and as auxiliary seals in storage and transportation casks (dual purpose casks) for heat generating radioactive waste, such as spent fuel and high-level waste. While a seal exchange at defined intervals is typical in many conventional applications, it is impossible or at least hard to perform when principles of minimization of radiation exposure have to be considered and prohibit an unnecessary cask handling. An extensive knowledge of the change of the elastomer’s properties during aging and the availability of reliable end-of-lifetime criteria to guarantee the permanent safe enclosure of the radioactive material is mandatory. As BAM is involved in most of the national cask licensing procedures and in the evaluation of cask-related long-term safety issues, great efforts have been already made and are still planned to scientifically support this task. Compression stress relaxation and compression set were identified as key indicators of elastomer long-term performance and quantitatively investigated in comprehensive test programs. Among other representative types of elastomers, specimens made from ethylene propylene diene rubber (EPDM) were tested before, during and after aging to capture the most important of their complex mechanical properties. In the presented study, exemplary results were used to simulate the compression stress relaxation and the compression set of elastomer O-rings during aging. Regarding the influence of temperature, the time-temperature superposition principle is applied in the relaxation analysis of elastomer O-rings. The proposed model is implemented in the commercial finite element software ABAQUS/Standard® [1] with a sequential temperature displacement coupling. Numerical results match the experimental compression stress relaxation measurements well. The prediction of compression set values after long-term aging shows a relatively good agreement with the experimental results. Nevertheless, all input parameters derived from the specimen tests, additional assumptions concerning boundary conditions and modeling strategy are discussed with regard to the identified slight discrepancies. The possibility to extend the finite element model to represent the O-ring seal’s ability to recover after a (fast) partial release is taken into account. T2 - ASME 2020 Pressure Vessels & Piping Conference (PVP2020) CY - Online meeting DA - 03.08.2020 KW - Compression Set KW - O-ring KW - Simulation KW - Sequential analysis KW - Ageing KW - stress relaxation PY - 2020 SN - 978-0-7918-8388-4 DO - https://doi.org/10.1115/PVP2020-21270 SP - PVP2020-21270 PB - ASME AN - OPUS4-51490 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin A1 - Quercetti, Thomas A1 - Wille, Frank T1 - Fire reference package - Results of large-scale fire tests N2 - Packages for the transport of high-level radioactive material are designed to withstand severe accidents. Hypothetical severe accident conditions are defined in the IAEA Regulations for the Safe Transport of Radioactive Materials. One of these accident conditions is the thermal test, mainly consisting of a 30 minute fully engulfing 800°C pool fire or an equally severe fire test. The heat fluxes into the package depend substantially on the fire characteristics and the surface temperature of the package. Fire tests can be performed at BAM on a propane gas fire test facility. In order to investigate the heat fluxes over a wide range of surface temperatures in this test facility a fire reference package was designed for multiple use. The package represented the outer geometry of a specific transport cask for radioactive waste. The fire reference package is a closed steel sheet cylinder with a wall thickness of 10 mm, a length of 1500 mm and a diameter of 1050 mm. The package was instrumented with thermocouples and filled with heat resistant insulation material. Three open-air fire tests were performed in the BAM propane gas fire test facility. The flames exposure time period varied slightly for the fire tests. The wind direction as well as the wind speed were measured and changed between and during the tests. Test stand parameters such as wind shield location and propane gas volume flow were chosen constant for the three tests. The locally measured fire reference package steel sheet temperatures were used for the calculation of heat fluxes as function of time and surface temperature. The measured temperatures allowed further calculations. In a first approach effective fire characteristics of the propane gas fire, including the flame temperature, the fire convection coefficient and a radiation exchange coefficient mathematically describing the determined average heat flux over the surface temperature were calculated. T2 - IRSN-BAM Symposium - Safety of Transport and Storage Packages CY - Online meeting DA - 19.11.2020 KW - Fire reference KW - Calorimetric test KW - Propane KW - Heat flux KW - Propane gas fire test facility PY - 2020 AN - OPUS4-51686 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Linnemann, Konrad A1 - Quercetti, Thomas A1 - Sterthaus, Jens A1 - Ballheimer, Viktor T1 - Prestressed bolt connections under lateral displacements N2 - The containment system of transport packages for spent nuclear fuel and high-level waste usually includes bolted lids with metal gaskets. The specified transport condition imply high loading on the lids and the bolt connections of the package. The response of the lid systems on these load conditions is generally investigated by drop tests or numerically. BAM has started a research project to get a better understanding about the behavior of prestressed bolt connection under loadings typical for drop tests. T2 - IRSN-BAM Symposium Safety of Transport and Storage Packages CY - Online meeting DA - 19.11.2020 KW - Transport packages KW - Pprestressed bolt connection KW - Experimental investigation KW - Finite element method PY - 2020 AN - OPUS4-51697 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias A1 - Chen, Hefeng T1 - Numerical Analysis for an Electro‐Magneto‐Mechanical Phenomenon with High‐Order Accurate Methods N2 - This paper establishes an axisymmetric model for a levitation device. Therein, the Maxwell equations are combined with the balance of linear momentum. Different possible formulations to describe the MAXWELL equations are presented and compared and discussed in the example. A high order finite element discretization using GALERKIN's method in space and the generalized NEWMARK‐α method in time are developed for the electro‐magneto‐mechanical approach. Several studies on spatial and temporal discretization with respect to convergence will be investigated. In addition, the boundary influences and the domain size with respect to the levitation device are also examined. T2 - International Association for Applied Mathematics and Mechanics CY - Kassel, Germany DA - 16.03.2020 KW - High-Order Methods KW - Electro-Magneto-Mechanical KW - Levitation Device PY - 2021 UR - https://nbn-resolving.org/urn:nbn:de:kobv:b43-520720 DO - https://doi.org/10.1002/pamm.202000018 VL - 20 IS - 1 SP - e202000018 PB - Wiley‐VCH GmbH CY - Online AN - OPUS4-52072 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zaghdoudi, Maha A1 - Weber, Mike A1 - Kömmling, Anja A1 - Jaunich, Matthias A1 - Wolff, Dietmar T1 - Numerical modelling of compression stress relaxation and compression set of elastomer O-rings during aging N2 - Elastomer seals are used in many industrial applications due to their excellent mechanical properties at a wide range of temperatures. Their high versatility and recovery potential under several load conditions make them well suitable for the application in containers designed for transport, storage and/or disposal of radioactive materials. In containers for low and intermediate level radioactive waste, elastomer seals are used as barrier seals, and as auxiliary seals in storage and transportation casks (dual purpose casks) for heat generating radioactive waste, such as spent fuel and high-level waste. While a seal exchange at defined intervals is typical in many conventional applications, it is impossible or at least hard to perform when principles of minimization of radiation exposure have to be considered and prohibit an unnecessary cask handling. An extensive knowledge of the change of the elastomer’s properties during aging and the availability of reliable end-of-lifetime criteria to guarantee the permanent safe enclosure of the radioactive material is mandatory. As BAM is involved in most of the national cask licensing procedures and in the evaluation of cask-related long-term safety issues, great efforts have been already made and are still planned to scientifically support this task. Compression stress relaxation and compression set were identified as key indicators of elastomer long-term performance and quantitatively investigated in comprehensive test programs. Among other representative types of elastomers, specimens made from ethylene propylene diene rubber (EPDM) were tested before, during and after aging to capture the most important of their complex mechanical properties. In the presented study, exemplary results were used to simulate the compression stress relaxation and the compression set of elastomer O-rings during aging. Regarding the influence of temperature, the time temperature superposition principle is applied in the relaxation analysis of elastomer O-rings. The proposed model is implemented in the commercial finite element software ABAQUS/Standard® [1] with a sequential temperature displacement coupling. Numerical results match the experimental compression stress relaxation measurements well. The prediction of compression set values after long-term aging shows a relatively good agreement with the experimental results. Nevertheless, all input parameters derived from the specimen tests, additional assumptions concerning boundary conditions and modeling strategy are discussed with regard to the identified slight discrepancies. The possibility to extend the finite element model to represent the O-ring seal’s ability to recover after a (fast) partial release is taken into account T2 - ASME 2020 Pressure Vessels & Piping Conference (PVP2020) CY - Online meeting DA - 03.08.2020 KW - Stress relaxation KW - Ageing KW - Compression Set KW - O-ring KW - Sequential analysis KW - Simulation PY - 2020 AN - OPUS4-52135 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kömmling, Anja A1 - Jaunich, Matthias A1 - Wolff, Dietmar T1 - Aging of elastomer seals for nuclear waste containers – Methods and lifetime prediction N2 - At BAM, which is a federal institute for materials research and testing in Germany, it is one of our tasks to evaluate the safety of casks designed for transport and/or storage of radioactive material. This includes assessment of elastomeric seals applied in the casks. One of the aims is to identify a suitable method for estimating the service lifetime of the seals with regard to the requirements for long-term safety (40 years and more) of the containers. Therefore, we started an accelerated ageing programme with selected rubbers often used for seals (HNBR, EPDM and FKM) which are aged at four different temperatures (75 °C, 100 °C, 125 °C and 150 °C) for up to five years. In order to assess sealability, O-rings were aged in compression by 25 % (corresponding to the compression during service) between plates as well as in flanges that allow leakage rate measurements. For comparison, uncompressed O-rings were aged as well. In order to understand the underlying ageing mechanisms in each material, material properties such as hardness, density and tensile properties were examined. Additionally, compression set (CS), which represents the recovery behaviour of a seal after release from compression, is measured. For obtaining results closely related to practical conditions, O-rings with a full-scale cord diameter of 10 mm were aged. However, this set-up can lead to heterogeneous ageing caused by diffusion-limited oxidation (DLO) effects especially for HNBR, resulting in distorted bulk properties such as compression set. However, if DLO-affected data is excluded, extrapolations of CS data are possible using time-temperature shifts. For selecting an appropriate end-of-lifetime criterion, leakage rate measurements were performed, since leakage rate is the only characteristic directly correlated to the performance of the sealing system. A significant increase in leakage rate was considered as the end of the lifetime. However, the O-rings remained leak tight under static conditions even though material properties had already degraded strongly. For this reason, a modified, more demanding leakage test involving a fast small decompression of the seal was developed that allowed determining a more conservative end-of-lifetime criterion with a safety margin for EPDM seals. FKM, which is a very heat and oxidation resistant material, exhibited only little degradation, even though it had the smallest activation energy. T2 - Polymers in nuclear applications CY - Espoo, Finland DA - 27.11.2019 KW - rubber KW - leakage KW - degradation KW - HNBR KW - EPDM KW - FKM PY - 2019 UR - https://www.energiforsk.se/media/27280/agingseals_kommling_bam.pdf AN - OPUS4-49918 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias A1 - Kömmling, Anja A1 - Zaghdoudi, Maha A1 - Wolff, Dietmar T1 - A closer look at elastomer O-ring seals - Interesting aspects of these typical machine elements and ways to evaluate their performance - N2 - For many sealing applications rubber O-rings are applied due to their special properties and easy use. In many applications as e.g. in aviation, traffic or process technology the material can be exposed to a variety of conditions during operation as e.g. low temperatures, elevated temperatures and different media. Therefore, it is important to know the material behaviour and the seal performance under those conditions. Additionally ageing of seals is an important factor which has an influence on the long term reliability. In this contribution we would like to discuss different aspects affecting the seal performance, lifetime estimation and address open questions which should be addressed to come to a deeper understanding of these seals commonly understood as typical machine elements. Results of running investigations at the Bundesanstalt für Materialforschung und -prüfung (BAM), Division 3.4 “Safety of Storage Containers” are presented and discussed. T2 - Technomer 2019 CY - Chemnitz, Germany DA - 07.11.2019 KW - Seal KW - Machine element KW - Rubber PY - 2019 AN - OPUS4-49981 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CHAP A1 - Droste, Bernhard A1 - Wolff, Dietmar A1 - Bevilacqua, A. A1 - Reiche, I. A1 - Harvey, J. A1 - Hirose, M. A1 - Kumano, Y. A1 - McConnell, P. A1 - Saegusa, T. A1 - Einziger, R. ED - González-Espartero, A. T1 - Methodology for a Safety Case of a Dual Purpose Cask for Storage and Transport of Spent Fuel N2 - Spent nuclear fuel which is generated in the operation of nuclear reactors needs to be safely managed following its removal from the reactor core. On-site power reactor storage pools were designed on the assumption that after a short period of time spent nuclear fuel would be removed for reprocessing and disposal or further storage elsewhere. The amount of highly radioactive spent fuel that needs to be stored over longer periods of time is growing and additional storage capacity is required. One of the widely used options for additional storage capacity is the use of casks for dry storage of spent fuel. Among various existing dry storage concepts, several Member States are utilizing a concept of dual purpose casks (DPCs). This publication provides practical advice on the structure and contents of a DPC integrated safety case with reference to existing IAEA requirements relevant to the licensing and use of transport and storage casks for spent fuel. KW - Dual Purpose Cask KW - Spent Fuel KW - Storage KW - Transport PY - 2020 UR - https://www.iaea.org/publications/10966/methodology-for-a-safety-case-of-a-dual-purpose-cask-for-storage-and-transport-of-spent-fuel SN - 978-92-0-131620-2 SN - 1011–4289 SP - 1 EP - 109 PB - IAEA Publishing Section CY - Vienna, Austria AN - OPUS4-51962 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Ruiz-Hervias, J. A1 - Simbruner, Kai A1 - Cristobal-Beneyto, M. A1 - Perez-Gallego, D. A1 - Zencker, Uwe T1 - Failure mechanisms in unirradiated ZIRLO® cladding with radial hydrides N2 - The purpose of this paper is to investigate the relationship between hydride morphology, in particular the presence of radial hydrides (RHs), stress state and failure mechanisms associated with the ring compression test (RCT). Samples of ZIRLO®cladding were pre-hydrided and subjected to thermo-mechanical treatments to precipitate long radial hydrides. The results show that the reorientation treatment was very successful. A considerable fraction of RHs was generated, the radial hydride continuity factor being around 80 to 90% of the wall thickness. The samples with reoriented hydrides were tested using the RCT at room temperature. Macroscopic brittle failure was observed with sudden load drops for displacements around 0.5 mm, with a calculated “offset strain”between 0.5 and 1%. Crack nucleation occurs in RHs located in regions with the highest values of hoop stress. These locations are the inner diameter of cladding at the vertical plane of the sample (12 and 6 o’clock positions) and the outer diameter at the horizontal plane (3 and 9 o’clock positions). Noticeable load drops in the RCT are associated with unstable crack propagation events through the radial hydride network, the crack front reaching up to 90% of the wall thickness in some cases. The failure micro-mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth and coalescence in the Zr matrix, with ductile tearing patches connecting neighboring hydrides. The main conclusion is that radial hydride metrics is not the only parameter that determines cladding failure in the presence of RHs, but the interaction between the location and continuity of RHs and the stress normal to the hydride (the hoop stress in this case). Consequently, if a radial hydride is located at a position within the cladding where the hoop stress is small, a crack will not be initiated easily in the RCT. KW - ZIRLO® KW - Cladding Embrittlement KW - Radial Hydrides KW - Failure Mechanism KW - Ring compression test PY - 2021 DO - https://doi.org/10.1016/j.jnucmat.2020.152668 SN - 0022-3115 VL - 544 SP - 152668 PB - Elsevier B.V. AN - OPUS4-52000 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CHAP A1 - Kömmling, Anja A1 - Wolff, Dietmar A1 - Jaunich, Matthias ED - van Breugel, K. ED - Koleva, D. ED - Beek, T. T1 - Investigation of long-term behaviour of elastomeric seals for transport and storage packages N2 - Elastomers are widely used as the main sealing materials in Containers for low- and intermediate-level radioactive waste and as an additional component fo metal seals in spent-fuel and high-level waste containers. According to appropriate guidelines and regulations, safe enclosure of the radioactive container contents has to be guaranteed for lengthy storage periods of at least 40 years. Therefore, the understanding of seal ageing behaviour is of high importance and has to be considered with regard to possible dynamic events taking place during transport after storage. An accelerated ageing approach for compressed seals is presented, as well as some first results. KW - O Ring KW - Ageing KW - Rubber seal PY - 2018 SN - 978-3-319-70192-9 DO - https://doi.org/10.1007/978-3-319-70194-3_2 SP - 17 EP - 25 PB - Springer AN - OPUS4-43248 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Grelle, Tobias A1 - Probst, Ulrich A1 - Jaunich, Matthias A1 - Skrotzki, Birgit A1 - Wolff, Dietmar T1 - Creep investigations on aluminum seals for application in radioactive waste containers N2 - In Germany spent nuclear fuel and high level radioactive waste is stored in interim storage containers with double lid systems. Those lids are equipped with metal seals (e.g. Helicoflex®) that ensure the safe enclosure of the inventory. The used metal seals consist of three components as can be seen in the cross-sectional view in Figure 1. The innermost part is a helical spring that is surrounded by an inner jacket made of stainless steel. The outer jacket that is made of a softer material which in case of assembly in the aforementioned storage containers is silver or aluminum (i.e. Al 99.5). During application the seal is compressed and due to the restoring force of the helical spring, the outer jacket is plastically deformed and adapts to the sealing surface. Hence, leakage paths are closed and the sealing function is generated. In Germany the above-mentioned containers are licensed for up to 40 years of interim storage, which in case extended storage becomes necessary before a final repository is available will have to be extended to even longer periods. Therefore, the evaluation of the long-term behavior of the seals is necessary, taking into account storage conditions, decay heat and possible mechanical loads as well. At Bundesanstalt für Materialforschung und –prüfung (BAM) long-term investigations are being conducted in which seals are assembled in test flanges and aged at temperatures ranging from room temperature to 150°C for accelerated aging. The aged seals are tested semi-annually (after the first 6 months in which the seals are tested more frequently) regarding the sealing performance, the remaining seal force, and the useable resilience upon decompression. Results of these investigations have been published over the past years (e.g. Grelle, Wolff, Probst, Jaunich, & Völzke, 2017; Völzke, Wolff, Probst, Nagelschmidt, & Schulz, 2014). It was found that the seal force and the useable resilience decrease with time and temperature, which is in agreement with the result of other studies (Sassoulas et al., 2006; Wataru et al., 2016) as well. Geometry change of the outer jacket has been identified as the main reason for this seal behavior. At the prevailing operating temperatures and stresses the aluminum is subjected to creep deformation leading to a thinning of the outer jacket. Since the seal groove depth remains unchanged the helical spring expands, which in turn leads to a decrease of the generated spring and seal force. Although the main reason for the change of seal parameters over time and temperature is known, a detailed characterization of the seal behavior and a reliable prediction of the parameter development for aging times that exceed the experimental time frame have not been possible, yet. For deeper understanding of the aging processes, an Investigation program, which is covered in this contribution, is conducted at Bundesanstalt für Materialforschung und –prüfung (BAM) that focusses on the behavior of the aluminum jacket and its influence on the long-term sealing performance. The program investigates properties of material samples as well as the behavior of the seal as a component. Original sheet material of the same aluminum that is used for manufacturing of the seals is investigated in compression creep tests. For this, a DMA (dynamic mechanical analysis) machine is employed (here used for static tests) that allows for a measurement of the specimens deformation under forces of up to 500 N. The advantage of this method is that the original material can be tested in the same shape as used for the seals which is 0.5 mm thick sheet material. For investigation of tensile creep standard specimens are used, that were machined from surrogate material of the same composition and annealing condition. Furthermore, aluminum seals that are cut into smaller segments are assembled in flanges and placed in heating chambers at temperatures ranging from 23°C to 150°C. After different periods of time from 3 days to 300 days the segments are taken out of the flanges and are investigated, thus giving information on different states of aging. Measurements of the development of the seal contact width and the aluminum jacket thickness are done with an optical microscope. Further investigations on the segments will include metallography and hardness measurements. From the detailed material and component behavior including the results of the long-term seal force and useable resilience investigations a better understanding of the overall seal behavior can be gained. The aim is to contribute to the development of material models and analytical approaches for the prediction of the sealing behavior in dependence of time and temperature. T2 - International Conference on Aluminum Alloys CY - Montreal, Canada DA - 17.06.2018 KW - Metal seal KW - Creep KW - Long-term behavior PY - 2018 SN - 978-1-926872-41-4 SP - 1 EP - 2 AN - OPUS4-45844 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Linnemann, Konrad A1 - Ballheimer, Viktor A1 - Rolle, Annette A1 - Wille, Frank T1 - Considerations on spent fuel behavior for transport after extended storage N2 - Packages for the transport of spent nuclear fuel shall meet the International Atomic Energy Agency regulations to ensure safety under different transport conditions. The physical state of spent fuel and the fuel rod cladding as well as the geometric configuration of fuel assemblies are important inputs for the evaluation of package capabilities under these conditions. In this paper, the mechanical behavior of high burn-up spent fuel assemblies (> approx. 50 GWd/tHM, value averaged over the fuel assembly) under transport conditions is analyzed with regard to the assumptions which are used in the Containment and criticality safety analysis. In view of the complexity of the interactions between the fuel rods as well as between the fuel assemblies, basket, and cask containment, the exact mechanical analysis of such phenomena is nearly impossible. Additionally, the gaps in information concerning the material properties of cladding and pellet behavior, especially for the high burn-up fuel, make the analysis more complicated. Considerations and knowledge gaps for the transport after extended interim storage are issues of growing interest. In this context, practical approaches are discussed based on the experience of BAM within the safety assessment of packages approved for transport of spent nuclear fuel. KW - Transport packages for radioactive material KW - Spent nuclear fuel PY - 2018 VL - 83 IS - 6 SP - 488 EP - 494 PB - Hanser AN - OPUS4-47359 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Jaunich, Matthias T1 - Editorial N2 - Dear Readers, a new year started for all of us (as I write these lines, possibly it was already some time ago when these lines are finally available) and this is the time for the so-called New Year’s resolutions. Setting goals and changing our behavior as well in a private as in a professional context. For Scientists the resolutions may comprise tasks like finalizing a longago started publication, picking up the loose ends of an application for funding or just bringing the running projects in a greater structure. KW - Editorial KW - Polymer Testing PY - 2019 DO - https://doi.org/10.1016/j.polymertesting.2019.01.010 SN - 0142-9418 VL - 74 SP - A1 PB - Elsevier AN - OPUS4-47406 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Kömmling, Anja A1 - von der Ehe, Kerstin A1 - Wolff, Dietmar A1 - Jaunich, Matthias T1 - Effect of high-dose gamma irradiation on (U)HMWPE neutron shielding materials N2 - High and ultra-high molecular weight polyethylenes were gamma-irradiated with doses up to 600 kGy. The changes in the material properties were analysed using DSC, DMA, IR spectroscopy, as well as measurements of density and insoluble content. The irradiation led to an increase of the degree of crystallinity because of chain scissions during irradiation, leading to shorter and thus more mobile chains. Both the plateau value of the shear modulus G′ and the insoluble content increased with Irradiation dose, indicating the formation of additional crosslinks. Furthermore, IR spectroscopy revealed irradiation induced oxidation and the formation of double bonds, indicating that some of the hydrogen atoms responsible for the neutron shielding capability have been released. KW - Ultra high molecular weight KW - Polyethylene KW - Gamma irradiation KW - Crosslinking KW - Oxidation PY - 2018 DO - https://doi.org/10.1016/j.radphyschem.2017.02.014 SN - 0969-806X VL - 142 SP - 29 EP - 33 PB - Elsevier AN - OPUS4-42941 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias A1 - Goral, Milan A1 - Grelle, Tobias A1 - Wolff, Dietmar T1 - Low temperature performance of rubber seals - Influence of crystallite formation on seal performance N2 - Rubbers are widely used as sealing materials in various applications. In many fields the function of seal materials at low temperatures is necessary. Therefore, the understanding of failure mechanisms that lead to leakage at low temperatures is of high importance. It is known that rubbers possess strongly temperature dependent material properties. At low temperatures high changes in the material properties are observed which can be caused by the rubber-glass transition (abbr. glass transition) or for some rubbers also by partial crystallisation at lower temperatures. During continuous cooling the material changes during rubber-glass transition from rubber-like entropy-elastic behaviour to stiff energy-elastic behaviour. In case of (thermal-) crystallisation the formed crystallites act as additional (physical) crosslinks within the material but the molecules in the amorphous phase can still behave like a rubber. Hence, rubbers are normally used above rubber-glass transition and (thermal-) crystallization temperature but the minimum working temperature limit is heavily depending on the application conditions. Therefore, the lower operation temperature limit of rubber seals is under investigation. In the past we have investigated material properties and sealing behaviour of rubber seals at low temperatures. As the initial work was dedicated to purely static sealing conditions we started to investigate the effect of a small partial release of rubber seals at low temperatures to study the influence of dynamic events on leak tightness. In addition, other groups report on the influence of counterpart surface and adhesion. In this contribution we focus on the difference between purely amorphous elastomers and elastomers which become partially crystalline at low temperatures. The influence of partial crystallization on the low temperature performance of rubber seals is discussed. T2 - Fall Rubber Colloquium 2018 CY - Hannover, Germany DA - 06.11.2018 KW - Seal KW - Low temperature KW - Rubber KW - Partial crystalline PY - 2018 AN - OPUS4-46553 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Rolle, Annette A1 - Ballheimer, Viktor A1 - Neumeyer, Tino A1 - Wille, Frank T1 - Application of leakage rates measured on scaled cask or component models to the package containment safety assessment N2 - The containment systems of transport casks for spent fuel and high radioactive waste usually include bolted lids with metallic or elastomeric seals. The mechanical and thermal loadings associated with the routine, normal and accident conditions of transport can have a significant effect on the leak tightness of such containment system. Scaled cask models are often used for providing the required mechanical and thermal tests series. Leak tests have been conducted on those models. It is also common practice to use scaled component tests to investigate the influence of deformations or displacements of the lids and the seals on the standard leakage rate as well as to study the temperature and time depending alteration of the seals. In this paper questions of the transferability of scaled test results to the full size design of the containment system will be discussed. T2 - ASME 2018 Pressure Vessels and Piping Conference CY - Prag, Czech Republic DA - 15.08 2018 KW - Containment KW - Spent fuel KW - Leakage rate PY - 2018 VL - 7 SP - PVP2018-84089, 1 EP - 8 AN - OPUS4-46533 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Zencker, Uwe T1 - Numerical approach to determine the correct puncture bar length for the IAEA puncture bar drop test N2 - Among other mechanical tests the 1 meter drop onto a steel puncture bar shall be considered for accident safe packages for the transport of radioactive material. According with the IAEA regulations “the bar shall be of solid mild steel of circular section, 15.0 ± 0.5 cm in diameter and 20 cm long, unless a longer bar would cause greater damage…”. The most damaging puncture bar length can be estimated by iterative processes in numerical simulations. On the one hand, a sufficient puncture bar length has to guarantee that shock absorbers or other attachments do not prevent or reduce the local load application to the package, on the other hand, a longer and thus less stiff bar causes a smaller maximum contact force. The contrary influence of increasing puncture bar length and increasing effective drop height shall be taken into account if a shock absorber is directly placed in the target area. The paper presents a numerical approach to identify the bar length that causes maximum damage to the package. Using the example of two typical package masses the sensitivity of contact forces and puncture bar deformations to the initial length are calculated and assessed with regard to the international IAEA package safety requirements. T2 - ASME Pressure Vessels and Piping Conference 2018 CY - Prague, Czech Republic DA - 15.07.2018 KW - Length of puncture bar KW - Mechanical assessment KW - Numerical simulation KW - Puncture bar test KW - Transport of radioactive materials PY - 2018 SP - PVP2018-84614, 1 EP - 7 AN - OPUS4-46538 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Grelle, Tobias A1 - Probst, Ulrich A1 - Jaunich, Matthias A1 - Skrotzki, Birgit A1 - Wolff, Dietmar T1 - Creep Investigations on Aluminum Seals for Application in Radioactive Waste Containers N2 - In Germany spent nuclear fuel (SNF) and high level radioactive waste (HLW) are stored in interim storage containers with double lid systems. Those lids are equipped with metal seals (e.g. Helicoflex®) that ensure the safe enclosure of the inventory. Being licensed for up to 40 years of interim storage the evaluation of the long-term behavior of the seals is necessary, taking into account storage conditions, decay heat and possible mechanical loads. T2 - International Conference on Aluminum Alloys CY - Montreal, Canada DA - 17.06.2018 KW - Metal seal KW - Creep KW - Long-term behavior PY - 2018 AN - OPUS4-45843 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin A1 - Erenberg, Marina A1 - Bletzer, Claus Wilhelm A1 - Musolff, André A1 - Nehrig, Marko A1 - Wille, Frank T1 - Investigations of the burning behavior of transport package impact limiters and thermal effects onto the cask N2 - Accident safe packages for the transport of spent nuclear fuel and high-level waste shall fulfil international IAEA safety requirements. Compliance is shown by consecutive mechanical and thermal testing. Additional numerical analysis are usually part of the safety evaluation. For damage protection some package designs are equipped with wood filled impact limiters encapsulated by steel sheets. The safety of these packages is established in compliance with IAEA regulations. Cumulative mechanical and fire tests are conducted to achieve safety standards and to prevent loss of containment. Mechanical reliability is proven by drop tests. Drop testing might cause significant damage of the impact limiter steel sheets and might enable sufficient oxygen supply to the impact limiter during the fire test to ignite the wood filling. The boundary conditions of the fire test are precisely described in the IAEA regulatory. During the test the impact limiter will be subjected to a 30 minute enduring fire phase. Subsequent to the fire phase any burning of the specimen has to extinguish naturally and no artificial cooling is allowed. At BAM a large-scale fire test with a real size impact limiter and a wood volume of about 3m³ was conducted to investigate the burning behaviour of wood filled impact limiters in steel sheet encapsulation. The impact limiter was equipped with extensive temperature monitoring equipment. Until today burning of such impact limiters is not sufficiently considered in transport package design and more investigation is necessary to explore the consequences of the impacting fire. The objective of the large scale test was to find out whether a self-sustaining smouldering or even a flaming fire inside the impact limiter was initiated and what impact on the cask is resulting. The amount of energy, transferred from the impact limiter into the cask is of particular importance for the safety of heavy weight packages. With the intention of heat flux quantification a new approach was made and a test bench was designed. A first computational simulation of transport package temperatures taking into account the results of the conducted fire test was performed. T2 - IRSN Conference on Safe Transport of Radioactive Material CY - Fontenay aux Roses, France DA - 13.11.2018 KW - Impact limiter KW - Shock absorber KW - Smoldering KW - Smouldering KW - Burning KW - Thermal testing KW - BAM TTS KW - Combustion KW - Fire KW - Energy release KW - Thermal simulation KW - Heat emission KW - Radioactive KW - Transport KW - IAEA KW - Wood KW - Spruce wood KW - Lid temperature PY - 2018 AN - OPUS4-46882 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Linnemann, Konrad A1 - Ballheimer, Viktor A1 - Rolle, Annette T1 - Spent fuel behavior under Transport Conditions N2 - German packages for the transport of spent nuclear fuel are assessed with respect to specific transport conditions which are defined in the safety regulations of the International Atomic Energy Agency. In general, gastight fuel rods constitute the first barrier of the containment system. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important inputs for the evaluation of the package safety under transport conditions. The objective of this paper is to discuss the methodologies accepted by German authority BAM for the evaluation of spent fuel behavior within the package design approval procedure. Specific test conditions will be analyzed with regard to assumptions to be used in the activity release and criticality safety analysis. In particular the different failure modes of the fuel rods, which can cause release of gas, volatiles, fuel particles or fragments, have to be properly considered in these assumptions. The package as a mechanical system is characterized by a complex set of interactions, e.g. between the fuel rods within the assembly as well as between the fuel assemblies, the basket, and the cask containment. This complexity together with the limited knowledge about the material properties and the variation of the fuel assemblies regarding cladding material, burn-up and the operation history makes an exact mechanical analysis of the fuel rods nearly impossible. The application of sophisticated numerical models requires extensive experimental data for model verification, which are in general not available. The gaps in information concerning the material properties of cladding and pellets, especially for the high burn-up fuel, make the analysis more complicated additionally, and require a conservative approach. In this context some practical approaches based on experiences by BAM within safety assessment of packages for transport of spent fuel will be discussed. Ongoing research activities to investigate SNF mechanical behavior in view of gas and fissile material release under transport loads are presented. T2 - IRSN Conference on Safe Transport of Radioactive Material CY - Fontenay aux Roses, France DA - 13.11.2018 KW - ENSA KW - SNL KW - Real transport KW - IAEA KW - Drop test KW - Multi-Modal Transportation Test KW - Radioactive KW - ACT KW - NCT KW - Hot cell KW - Activity release KW - Fuel rod KW - Fuel rods KW - Burn-up KW - Cladding alloy KW - Criticality safety analysis KW - Containment analysis KW - Fuel assemblies KW - Spent fuel PY - 2018 AN - OPUS4-46880 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Erenberg, Marina A1 - Bletzer, Claus A1 - Feldkamp, Martin A1 - Musolff, André A1 - Nehrig, Marko A1 - Wille, Frank T1 - Experimental investigations of the burning behaviour of transport package impact limiters and of fire spread impact onto the cask N2 - Accident safe packages for the transport of spent nuclear fuel and high-level waste shall fulfil international IAEA safety requirements. Compliance is shown by consecutive mechanical and thermal testing. Additional numerical analysis are usually part of the safety evaluation. For damage protection some package designs are equipped with wood filled impact limiters encapsulated by steel sheets. The safety of these packages is established in compliance with IAEA regulations. Cumulative mechanical and fire tests are conducted to achieve safety standards and to prevent loss of containment. Mechanical reliability is proven by drop tests. Drop testing might cause significant damage of the impact limiter steel sheets and might enable sufficient oxygen supply to the impact limiter during the fire test to ignite the wood filling. The boundary conditions of the fire test are precisely described in the IAEA regulatory. During the test the impact limiter will be subjected to a 30 minute enduring fire phase. Subsequent to the fire phase any burning of the specimen has to extinguish naturally and no artificial cooling is allowed. At BAM a large-scale fire test with a real size impact limiter and a wood volume of about 3m3 was conducted to investigate the burning behaviour of wood filled impact limiters in steel sheet encapsulation. The impact limiter was equipped with extensive temperature monitoring equipment. Until today burning of such impact limiters is not sufficiently considered in transport package design and more investigation is necessary to explore the consequences of the impacting fire. The objective of the large scale test was to find out whether a self-sustaining smouldering or even a flaming fire inside the impact limiter was initiated and what impact on the cask is resulting. The amount of energy, transferred from the impact limiter into the cask is of particular importance for the safety of heavy weight packages. With the intention of heat flux quantification a new approach was made and a test bench was designed. T2 - ASME 2018 Pressure Vessels and Piping Conference CY - Prague, Czech Republic DA - 15.07.2018 KW - Shock absorber KW - Impact limiter KW - Wood KW - Thermal testing KW - Fire KW - Smoldering KW - IAEA KW - Fire test PY - 2018 SN - 978-0-7918-5170-8 VL - PVP2018 SP - 84714-1 EP - 84714-10 AN - OPUS4-46984 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Zencker, Uwe T1 - Numerical Approach to Determine the Correct Length for the IAEA Puncture Bar Drop Test N2 - Among other mechanical tests the 1 meter drop onto a steel puncture bar shall be considered for accident safe packages for the transport of radioactive material. According with the IAEA regulations “the bar shall be of solid mild steel of circular section, 15.0 ± 0.5 cm in diameter and 20 cm long, unless a longer bar would cause greater damage…”. The most damaging puncture bar length can be estimated by iterative processes in numerical simulations. On the one hand, a sufficient puncture bar length has to guarantee that shock absorbers or other attachments do not prevent or reduce the local load application to the package, on the other hand, a longer and thus less stiff bar causes a smaller maximum contact force. The contrary influence of increasing puncture bar length and increasing effective drop height shall be taken into account if a shock absorber is directly placed in the target area. The paper presents a numerical approach to identify the bar length that causes maximum damage to the package. Using the example of two typical package masses the sensitivity of contact forces and puncture bar deformations to the initial length are calculated and assessed with regard to the international IAEA package safety requirements. T2 - ASME Pressure Vessels and Piping Conference 2018 CY - Prague, Czech Republic DA - 15.07.2018 KW - Transport of radioactive materials KW - Mechanical assessment KW - Puncture bar test KW - Length of puncture bar KW - Numerical simulation PY - 2018 AN - OPUS4-45785 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Rolle, Annette A1 - Ballheimer, Viktor A1 - Neumeyer, Tino A1 - Wille, Frank T1 - Application of leakage rates measured on scaled cask or component models to the package containment safety assessment N2 - The mechanical and thermal loadings associated with the routine, normal and accident conditions of transport can have a significant effect on the leak tightness of the containment system of transport casks for spent fuel and high radioactive waste. The containment systems of such transport casks usually include bolted lids with metallic or elastomeric seals. Scaled cask models are often used for providing the required mechanical and thermal tests series. Leak tests have been conducted on those models. It is common practice to use scaled component tests to investigate the influence of deformations or displacements of the lids and the seals on the standard leakage rate as well as to study the temperature and time depending alteration of the seals. In this paper questions and open points of the transferability of scaled test results to the fullscale design of the containment system will be discussed. T2 - 11th International Conference on the Transport, Storage and Disposal of Radioactive Materials CY - London, UK DA - 15. Mai 2018 KW - Transport packages KW - Radioactve KW - Seals KW - Leakage rate PY - 2018 AN - OPUS4-46734 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CHAP A1 - Wolff, Dietmar A1 - Bevilacqua, A. A1 - Carlsen, B. A1 - Chiguer, M. A1 - González-Espartero, A. A1 - Grahn, P. A1 - Saegusa, T. A1 - Sampson, M. A1 - Seelev, I. A1 - Wasinger, K. A1 - Waters, M. ED - González-Espartero (IAEA), A. T1 - Storing spent fuel until transport to reprocessing or disposal N2 - Spent fuel storage periods well beyond those originally foreseen are a reality. This publication offers several ideas and approaches that may be considered to address the increasingly longer storage times. The aim is to raise awareness, encourage dialogue and provide ideas on how to manage spent fuel. Key messages include the following: - Delays in reprocessing or disposal could result in spent fuel being stored for 100 years or longer. Safe, secure and effective storage of spent fuel manages fuel degradation while preserving future fuel cycle options. - Ageing management programmes apply engineering, operations and maintenance actions to ensure safety is maintained during storage, future handling and transport. - Site selection and facility and equipment design can significantly reduce the risks and costs of spent fuel storage over longer periods. - Spent fuel storage configurations can be selected to accommodate uncertain storage periods, to facilitate ageing management and to provide flexibility needed to accommodate the uncertainty of future end points, such as reprocessing or disposal. - By considering multiple licence renewals, regulatory frameworks can be designed to ensure safe storage until an acceptable end point is achieved. - Safety can be assured by maintaining shielding, containment, decay heat removal and criticality control. Navigating the complexity of societal beliefs and values, as well as political systems, has proven to be a greater challenge for the management of spent fuel than maintaining its safety and security or addressing the technical and economic aspects. - Sustainable spent fuel management requires policies and strategies to provide a clear, consistent and stable direction because they drive the need for spent fuel storage as well as the available options and timing for achieving an acceptable end point. Unless States address spent fuel reprocessing and disposal on a sufficient scale to accommodate their spent fuel discharges, then storage for longer and longer periods becomes the de facto end point — which is not considered to be consistent with the responsibility to protect human health and the environment. An effective, periodic licence renewal process can ensure effective ageing management and strong institutional control. Hence, spent fuel can be safely and securely stored for as long as it may be necessary until transport for reprocessing or disposal. However, the risks and costs of storing the growing inventory of spent fuel will continue to increase; and in the absence of an end point, it will eventually become a significant societal burden. KW - Storing spent fuel KW - Extended storing KW - Unknown storage duration PY - 2019 UR - https://www-pub.iaea.org/MTCD/publications/PDF/P1846_web.pdf SN - 978-92-0-100719-3 SN - 1995-7807 SP - 1 EP - 54 PB - IAEA Publishing Section CY - Wien, Austria AN - OPUS4-47693 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin A1 - Quercetti, Thomas A1 - Gleim, Tobias A1 - Wille, Frank T1 - Evaluation of Heat Fluxes in Fire Reference Test Conducted in BAM Propane Gas Fire Test Facility N2 - Packages for the transport of intermediate- and high-level radioactive waste are designed to withstand severe accidents. The International Atomic Energy Agency (IAEA) has established specific mechanical and thermal tests. Packages for the transport of radioactive material must withstand these tests to comply with the Regulations for the Safe Transport of Radioactive Materials IAEA [IAEA (2018)]. A fire reference package was developed with the primary objective to demonstrate that the fire meets the regulatory requirements. Another aim is to characterise the boundary conditions of the actual fire as input parameters for thermo-mechanical simulations. A simple method to characterise the boundary conditions of a real steady state fire with a fire reference package is presented. The thermal test mainly consists of a 30 minute fully engulfing 800°C pool fire or an equally severe fire, such as a propane gas fire. The fire reference tests are performed prior to the actual fire test with the real package. The heat fluxes into the package depend substantially on the fire characteristics and the surface temperature of the package. To investigate local and overall heat fluxes over a wide range of surface temperatures in this test facility a fire reference package was designed for repeated use. The fire reference package presented in this paper represents the outer geometry of a small transport container for radioactive material and is used as a device in civil engineering. It is designed as a closed steel sheet cylinder with a wall thickness of 10 mm, a length of 182 mm and a diameter of 102 mm. The package was instrumented with thermocouples and filled with heat resistant insulation material. Open-air fire tests were performed in a BAM propane gas fire test facility with the fire reference package. The measured temperatures are used to determine the changes of heat fluxes into the fire reference package in relation to the package surface temperature. The calculated heat fluxes allow its fitting to express the thermal exposure as simple mathematical boundary condition. Therefore, in a first approach, fire properties such as adiabatic surface temperature (AST) as proposed by Wickström et al. (2007), convection coefficient and emissivity are determined mathematically fitting the heat flux development presented in this paper. The evaluated results provide an initial picture of local fire characteristics of the conducted propane gas fire and are a further development of previous works from Feldkamp et al. (2020). The results can be used in thermal and thermo-mechanical models to simulate the load on the real transport package in fire. The test shows that the examined propane gas fire covers the IAEA-fire over a wide range of surface temperatures with the chosen test stand parameters. T2 - SMiRT 27 (27th conference on Structural Mechanics in Reactor Technology) CY - Yokohama, Japan DA - 03.03.2024 KW - Fire KW - Propane KW - Heat Flux KW - Fire Reference PY - 2024 SP - 1 EP - 10 PB - IASMiRT AN - OPUS4-59679 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Quercetti, Thomas A1 - Scheidemann, Robert T1 - Test Facilities for Radioactive Materials Transport and Storage Packagings at BAM N2 - BAM acts as authority and for service in safety assessment of packages for transport and storage of radioactive materials. We offer extensive test capabilities and application of analytical methods for design verification and simulation for all types of packages for the transport and storage of radioactive materials according with the international IAEA Regulations for the safe transport and for national storage acceptance criteria. BAM operates several test facilities for drop and stacking testing, leak testing and thermal testing. The large drop test tower allows dropping full-scale specimens up to 200,000 kg in any drop orientation as requested. The comprehensive test facilities combined with long-term experience, newest equipment and measurement devices according to the latest state-of-the-art technology ensures realisation of complex test campaigns for package safety evaluation. Beyond that, non-destructive and destructive material test devices and experts are available. Equipment and application of all kinds of typical measurement categories can be offered for testing campaigns. In recent years we performed testing of full-scale type B package models with complex handling and preparation procedures. The results were contributed for different package design approval procedures. Type A packages mainly designed for medical related transport purposes, were continuously tested according to the transport regulations over recent years as well. Moreover, we work on research topics with relevance to package safety. The mechanical behaviour of lid closure systems under transport and storage conditions and the thermal behaviour of impact limiters were recently of special importance for the assessment competencies of BAM and were investigated under use of our test facilities. The paper describes the test facilities and capabilities for package design safety evaluation at BAM and shows examples from our recent work. T2 - PATRAM 2022, 20th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Juan-les-Pins, France DA - 11.06.2023 KW - IAEA KW - Fire test KW - Drop testing KW - Transport KW - Package PY - 2023 SP - 1 EP - 12 AN - OPUS4-57967 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Schubert, Sven A1 - Reichardt, Adrian A1 - Müller, Lars A1 - Neumann, Martin A1 - Wille, Frank T1 - Ageing management measures for transport packages for radioactive materials N2 - The consideration of ageing mechanisms is with integration into the IAEA-requirements now obligatory for the design of transport packages. For packages intended to be used for shipment after storage the consideration of the effects of ageing mechanisms during storage in safety analyses and the implementation of corresponding instructions for operation and maintenance are required. Therefore, these packages have been maintained during storage in a manner that all requirements specified in the IAEA-requirements and in the applicable certificates of approval have been fulfilled. BAM is the competent authority in Germany for evaluating the thermal and mechanical design of packages for the transport of radioactive materials. In this context, BAM also evaluates the aging mechanisms and aging effects occurring during the long-term storage period of such transport packages. BAM has developed a guideline for the implementation of ageing assessment and of the measures for ageing management into the approval procedure based on IAEA-requirements. The goal of the presented paper is to introduce this guideline and the general approach for ageing management requirements and to describe the application to a specific package design. The type and amount of the ageing management not only depend on the use of the packaging, but also on the accessibility of the components that are considered in the ageing assessment. The different operating phases of a packaging are considered for the ageing assessment, such as loading, interim storage or transport. The implementation of measures for ageing management is divided into three levels – systemic measures, package design related measures and documentation. The systemic measures are attributed to the general management system and define the whole activities for organization of ageing management like structure, responsibilities, documentation, reports and evaluation. The package design related measures are defined in an ageing management plan (AMP). Therefore, an ageing surveillance program (ASP) and, if necessary, a gap analysis program has to be developed. The ageing management documentation (AMD) ensures the continuous documentation of the compliance of a specific package to the approved package design, comprising mainly records resulting from operation and surveillance. T2 - PVP2023-Pressure, Vessels & Piping Conference CY - Atlanta, GA, USA DA - 16.07.2023 KW - Radioaktive Stoffe KW - Transport KW - Gefahrgut PY - 2023 AN - OPUS4-58003 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Zaghdoudi, Maha A1 - Kömmling, Anja A1 - Jaunich, Matthias A1 - Wolff, Dietmar T1 - Understanding the recovery behaviour and the degradative processes of EPDM during ageing N2 - Recovery is an important measure for seal applications representing to which extent the elastomer regains its initial shape after deformation and release of an applied force. Compression set (CS) indicates the degree of recovery. Ethylene propylene diene rubber (EPDM) was aged at 75 ◦C, 100 ◦C, 125 ◦C and 150 ◦C for different ageing times up to five years and compression set measurements were performed at different times after disassembly and after additional tempering. Short- and long-term recovery up to one year after release for samples aged at 125 ◦C and 150 ◦C was also studied. To assess the curvature in the Arrhenius diagram that may occur due to non-sufficiently aged specimens, a degradation-rate based model was fitted to the CS data after tempering. For each ageing temperature, two decay fit functions were proposed, each with an activation energy and a corresponding degradative process. The influence of ageing on the leak-tightness after fast small partial release is investigated and estimated through the analysis of the shift factors from time temperature superposition (TTS) of CS measurements at different times after disassembly. Shift factors of CS measurement after 1 s and after additional tempering are in good agreement. KW - Compression set KW - Ageing KW - Recovery KW - Degradative processes KW - Leakage rate PY - 2023 UR - https://nbn-resolving.org/urn:nbn:de:kobv:b43-573699 DO - https://doi.org/10.1016/j.polymertesting.2023.107987 SN - 0142-9418 VL - 121 SP - 107987 PB - Elsevier Ltd. AN - OPUS4-57369 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Grelle, Tobias A1 - Wolff, Dietmar A1 - Probst, Ulrich A1 - Jaunich, Matthias A1 - Völzke, Holger T1 - Investigation of the time and temperature dependent behavior of metal seals in radioactive waste containers N2 - The Bundesanstalt für Materialforschung und –Prüfung (BAM) runs an investigation program on the long-term behavior of multi-component metal seals. Such seals are used in a wide area of applications including transport and storage casks for spent nuclear fuel and high level radioactive waste. The seal function is mainly based on the compression of the inner helical spring, which generates the necessary seal force to keep the sealing surfaces in close contact. This in turn leads to a plastic deformation of the outer jacket of the seal, comprised of highly ductile aluminum or silver that adapts to the sealing surfaces of cask body and bolted lid, thus providing high Level leak tightness. In Germany, those casks are licensed for Interim storage periods of up to 40 years or more if extended Interim storage would become necessary before a final repository is available. Thus, the sealing performance has to be evaluated, including factors like elevated temperature due to decay heat or mechanical loads due to transport under normal as well as accident conditions. Long-term investigations at BAM have been running over the last nine years to identify and evaluate the seal performance by measuring the remaining seal force, the useable resilience and the leakage rate after various time intervals at temperatures ranging from room temperature up to 150 °C. It was found that the seal force and useable resilience decrease with time and temperature, caused by creep deformation of the outer jacket. In order to obtain an analytical description for the seal behavior and to achieve more information on the material behavior under application conditions a comprehensive investigation program with Focus on aluminum as outer jacket material was launched. The program includes material investigations such as compression and tension creep tests with representative basic materials. An additional test setup allows for the continuous measurement of the remaining seal force at temperatures of up to 150 °C. Furthermore, seal segments are compressed and stored in heating chambers, thus producing segments at different stages of the aging process. The segments are investigated regarding the development of the contact area width, jacket thickness and microstructural changes. This data will be used to develop material models and an analytical description of the time and temperature dependent long-term sealing behavior. This paper explains the current status of gained test results and modelling approaches and closes with an outlook to the future Project plans. T2 - ASME 2018 Pressure Vessels and Piping Conference CY - Prague, Czech Republic DA - 15.07.2018 KW - Metal seal KW - Creep KW - Long-term behavior PY - 2018 SP - PVP2018-84584, 1 EP - 6 AN - OPUS4-46110 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger A1 - Jaunich, Matthias A1 - Zencker, Uwe A1 - Simbruner, Kai A1 - Nagelschmidt, Sven A1 - Herbrich, Uwe A1 - Keller, Christian A1 - Qiao, Linan T1 - Research at BAM on Metal Seal Performance and Fuel Rod Integrity N2 - The presentation provides an update on preliminary results from research projects in the area of long term performance of metal seals and fuel rod integrity as safety relevant components of spent fuel transport and storage casks for spent nuclear fuel. T2 - Extended Storage Collaboration Program (ESCP) Winter Meeting 2020 CY - Online meeting DA - 09.11.2020 KW - Metal seal KW - Safety KW - Interim storage KW - Spent nuclear fuel KW - Fuel rod PY - 2020 AN - OPUS4-51615 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - 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 AN - OPUS4-59147 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Simbruner, Kai A1 - Zencker, Uwe A1 - Völzke, Holger T1 - Embrittlement of Spent Fuel Claddings ‒ Results of Ring Compression Tests N2 - An established method for characterizing cladding material is the Ring Compression Test (RCT), where a small, cylindrical sample of the cladding tube is subjected to a compressive load. This test is a laboratory representation of a fuel rod load case and has shown a high susceptibility to failure under the occurrence of radial hydrides. Certain hydride morphologies and low temperature might even lead to brittle fracture at very small loads. The BAM research project BRUZL (Fracture mechanical analysis of spent fuel claddings during long-term dry interim storage) aims to study the results of quasi-static RCTs and to establish numerical models to gain an in-depth understanding of the stress state during such experiments. Using a fracture mechanical approach, cases of sudden failure during the RCT procedure are used to characterize the material behaviour and establish a failure criterion. As an integral part of the project BRUZL, quasi-static Ring Compression Tests have been performed to identify all experimental details, which might be helpful for numerical modelling. Unirradiated samples of the cladding material ZIRLO® have been subjected to hydrogen charging and a thermo-mechanical treatment for radial hydride reorientation. Sample preparation, testing procedures, and analysis results are presented. A numerical model has been established and an elastic-plastic material model was derived from as-received RCT samples by inverse finite element analyses. Cohesive zone modelling has been implemented to reproduce sudden load drops during RCT. The project is funded by the Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561. T2 - 4th GRS Workshop on Safety of Extended Dry Storage of Spent Nuclear Fuel CY - Online meeting DA - 03.06.2020 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2020 AN - OPUS4-51988 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias A1 - Kömmling, Anja A1 - Horn, Jutta A1 - Völzke, Holger A1 - Wolff, Dietmar T1 - Long-Term Performance of Elastomer seals - From Aging Tests to Lifetime Estimations N2 - Elastomers show a high versatility which makes them ideal materials for sealing applications in various fields. Especially under changing application conditions the high recovery potential of this class of material is beneficial to compensate temperature or pressure fluctuation, and geometrical changes resulting from mechanical loads in e.g. accident conditions. Out of these reasons elastomers are also used in containers for low and intermediate level radioactive waste and for spent fuel transportation casks. In cask designed for low and intermediate level waste elastomer seals can act as primary seal responsible for the containment function whereas in spent fuel storage and transportation casks (dual purpose casks (DPC)) elastomer seals are used as auxiliary seals to allow leakage rate measurements of metal barrier seals. An inherent prerequisite for this kind of application is the long time-scale of operation without or with limited possibility of seal replacement. In Germany an interim storage license for DPC`s is typically issued for 40 years, a timeframe which might increase in future due to challenges of the final repository siting procedure. For low and intermediate level waste, also long time periods are required before final disposal can be achieved. Therefore, the performance of elastomer seals over extended time periods is, as for other applications, of high importance. A typical approach to ensure long-term functionality is to perform accelerated aging tests to calculate an estimated lifetime by assuming e.g. Arrhenius like equations for the time-temperature relationship. This approach requires a suitable end of life criterion considering the application of interest. This often can represent a challenge on its own. As BAM is involved in most of the cask licensing procedures and especially responsible for the evaluation of cask-related long-term safety issues we initiated several test programs for investigating the behavior of elastomer seals. Experiments concerning the low temperature performance down to -40°C and the influence of gamma irradiation have been started first. Currently the thermal aging behavior of elastomer seals, which is the topic of this contribution, is examined. For our aging investigations we use a broad approach to first determine the property changes in different elastomer materials due to thermo-oxidative aging at elevated temperatures and secondly we test how the typical methods of lifetime extrapolation can be applied to these results. This approach enables us to detect and exclude undesired side effects which very often influence lifetime estimations. In this contribution, our recent results are discussed. T2 - ASME Pressure Vessels and Piping Conference 2018 CY - Prague, Czech Republic DA - 15.07.2018 KW - Seal KW - Rubber KW - Degradation PY - 2018 AN - OPUS4-45923 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Simbruner, Kai A1 - Billone, M. C. A1 - Zencker, Uwe A1 - Liu, Y. Y. A1 - Völzke, Holger T1 - Brittle Failure Analysis of High-Burnup PWR Fuel Cladding Alloys N2 - The general 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. T2 - 26th International Conference on Structural Mechanics in Reactor Technology - SMiRT 26 CY - Potsdam, Germany DA - 10.07.2022 KW - Cladding Embrittlement KW - Ring Compression Test KW - Cohesive Zone Modelling KW - Spent Nuclear Fuel PY - 2022 AN - OPUS4-55435 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Grelle, Tobias A1 - Wolff, Dietmar A1 - Probst, Ulrich A1 - Jaunich, Matthias A1 - Völzke, Holger T1 - Component and material investigations on metal seals for high level radioactive waste containers N2 - Long-term investigations performed at BAM look to extend the state of knowledge on safety-related components of interim storage containers. Metal seals act as the primary sealing barrier in the bolted double lid closure system of the containers. The behaviour of metal seals has been investigated for ageing times up to 8.5 years and for various temperatures. The main cause for reduction in useable resilience overtime was due to creep deformation of the outer jacket of the seal. KW - Metal seal KW - Radioactive waste containers KW - Creep KW - Long-term behaviour PY - 2018 SN - 1745-2058 SP - 32 EP - 34 PB - Nuclear Institute CY - London AN - OPUS4-48206 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -