TY - CONF A1 - Affagard, Jean-Sebastien A1 - Gleim, Tobias A1 - Louis, Baptiste A1 - Quercetti, Thomas A1 - Ledroit, Frédéric A1 - Létang, Eric T1 - Towards a simplifiedmodel of the delayed impact: numerical and experimental outlook N2 - The IAEA Regulations for the Safe Transport of Radioactive Material No. SSR-6 require drop tests for demonstrating the ability of a package to withstand accident conditions of transport. In case a gap exists between the contents and the lid, it can result in a delayed impact of the content onto the lid system during a vertical drop test. The kinematic energy transmitted to the lid in such cases can be significant due to the stiffness of the impact, leading to high stresses in the bolts compared to a configuration without a gap. In the past, IRSN and BAM have individually investigated the modelling of the delayed impact phenomenon. BAM has examined the effects of the delayed impact phenomenon for spent fuel packages in a comparative analysis between experiments and simulations, which was presented at SMIRT 2013. Meanwhile, IRSN’s has developed a simple numerical tool that can quickly evaluate the influence of various parameters before considering a more complex finite element numerical calculation. This tool was presented at PATRAM 2022. In 2023, BAM and IRSN (now ASNR - French nuclear safety and radiation protection authority) decided to jointly pursue their study on the delayed impact phenomenon, incorporating damping effects, by developing new insights and techniques. To this end, a controlled and simplified set of experiments needs to be developed to calibrate and validate the simplified model. These experiments will record velocity, position of parts, strain and accelerations. The parameters and the configurations necessary for the development of the mock-up, as well as the placement and quantity of sensors, are determined through extensive pre-calculations using both complex finite element models and the simplified approach. A key challenge involves precisely controlling the drop of the mock-up to accurately replicate the theoretical contact and damping conditions. Advanced sensor technologies will be employed to acquire reliable and sufficient data. The tests are designed and conducted at BAM's TTS drop test facility in collaboration with ASNR. T2 - PATRAM 2025 CY - San Antonio, TX, USA DA - 27.07.2025 KW - Experiments KW - Impact KW - Drop Tower PY - 2025 SP - 1 EP - 11 CY - Institute of Nuclear Materials Management (INMM) AN - OPUS4-63870 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Banerjee, Ayantika T1 - Enhancing Requirements for HLW Disposal Containers and Developing Transparent Safety Evaluation Concepts in Collaboration between BAM and BASE (Research Project KAnnE) N2 - The research project KAnnE aims to identify, concretize, and quantify requirements for high-level radioactive waste (HLW) disposal containers in Germany, considering potential types of host rock and repository concepts. The primary goal is to establish a transparent and comprehensible regulatory framework that accounts for different host rocks and operational phases, ensuring the safe long-term containment of radioactive waste. Given the current general legal formulation of these requirements, this study provides a systematic approach to defining them more precisely by integrating scientific and regulatory perspectives. An extensive review is conducted to gather relevant information about safety and design requirements for disposal containers in the respective types of host rock. The analysis considers host rock properties and conditions, operational repository conditions, and their impact on the container design, including key factors such as structural integrity, corrosion resistance, and shielding. Additionally, design specifications, boundary conditions, and expected loads (thermal, mechanical, radiological, chemical, and biological) are systematically reviewed, with quantified values provided where available. The findings support a comprehensive evaluation of both quantitative and qualitative container requirements concerning potential German host rock types, waste forms, and repository concepts. By establishing a clearer basis for defining these requirements, this project serves as a foundation for further refinement, with a more detailed structuring of requirements currently in progress. Such regulatory basis allows a more targeted approach to container development and ensures a transparent evaluation within the repository licensing procedure. Additionally, it supports decision making on the most favorable container design by considering repository conditions that provide the highest level of safety over one million years. This contribution explains the goals and concept of the research project KAnnE, presents its preliminary findings and conclusions, and provides an outlook on the final project phase. T2 - safeND conference 2025 CY - Berlin, Germany DA - 17.09.2025 KW - Radioactive waste KW - Final disposal KW - Waste container KW - Requirements KW - Regulation PY - 2025 AN - OPUS4-64990 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Beckmann, Jörg T1 - Photolytically and Thermally Initiated Destruction on the Stability of (Ultra)High Molecular Polyethylene N2 - THz and mid IR spectroscopy of high-molecular PE (HMW) and ultra high-molecular PE (UHMW) reveals modifications of the molecular structure. Characteristic absorption bands are changed if the two materials are exposed by γ-Co60 radiation up to 600 kGy and subsequently stored at an annealing temperature of 398 K until for 729 days. UHMW-PE and HMW-PE behave differently during the ageing process because of their molecular weight and inherent structure distinctions. The spectroscopic data offer characteristic absorption bands, which have been used to describe the complete ageing process in more detail. For instance, the integral absorption in the B1u THz-region can be used to describe quantitatively the reduction of crystallinity. The formation of trans vinylene unsaturation and the decay of vinyl during ageing can be observed in detail in the mid IR range. T2 - 9th International Workshop on Terahertz Technology and Applications CY - Kaiserslautern, Germany DA - 03.03.2020 KW - Degradation KW - MID IR Spectroscopy PY - 2020 AN - OPUS4-50555 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Brunk, Gerd A1 - Herbrich, Uwe A1 - Weckner, Olaf A1 - Smith, Jay A1 - Rossi, John T1 - Geometrically non-linear large planar deformations of shear deformable beams: Derivation, approximation, and numerical validation using Finite Element analysis N2 - In this paper, we present derivations for the dynamic equations of motion of a shear deformable beam undergoing geometrically non-linear large deformations in a single plane. In the static case, these governing equations are given by six coupled, non-linear ordinary differential equations of first order which we present in the Lagrangian frame of reference and henceforth refer to as the Brunk–Herbrich–Weckner (BHW) beam theory. We confirm analytically that under pure bending an initially straight beam deforms into a circle, a known bench-mark solution. Next, we confirm numerically that our results agree well with the commercial Finite-Element code Abaqus for several example problems with increasing complexity, including the effect of axial preload as well as varying property domains along the beam. Abaqus is regarded as the state-of-the-art for non-linear structural analysis within the aerospace industry. Next, we show that the well-known Timoshenko–Ehrenfest and Euler–Bernoulli beam theory can be obtained from the general Brunk-Herbrich-Weckner beam theory in the limit of small deformations. Finally, we introduce the Iterative Projection Method which improves the results of the linearized Timoshenko–Ehrenfest beam theory without the computational cost associated with the fully non-linear analysis for all numerical test cases considered. KW - Linear and non-linear beam theory KW - Deformation problem KW - Euler–Bernoulli, Timoshenko–Ehrenfest and Brunk–Herbrich–Weckner beam KW - Iterative projection method KW - Finite Element analysis PY - 2025 DO - https://doi.org/https://doi.org/10.1177/10812865251342777 SN - 1081-2865 SP - 1 EP - 26 PB - Sage CY - Thousand Oaks, Calif. AN - OPUS4-64383 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - RPRT A1 - Caruso, Stefano A1 - Vlassopoulos, Efstathios A1 - Dagan, Ron A1 - Fiorito, Luca A1 - Herm, Michel A1 - Jansson, Peter A1 - Kromar, Marjan A1 - Király, Márton A1 - Leppanen, Jaakko A1 - Feria Marquez, Francisco A1 - Metz, Volker A1 - Papaioannou, Dimitrios A1 - Herranz, Luis Enrique A1 - Rochman, Dimitri A1 - Schillebeeckx, Peter A1 - Seidl, Marcus A1 - Hernandez Solis, Augusto A1 - Stankovskiy, Alexey A1 - Alvarez Velarde, Francisco A1 - Verwerft, Marc A1 - Rodriguez Villagra, Nieves A1 - Zencker, Uwe A1 - Žerovnik, Gasper T1 - EURAD - Work Package 8 - Deliverable 8.1 - State-of-the-art report N2 - A state-of-the-art (SOTA) review on characterisation of spent nuclear fuel (SNF) properties in terms of source term and inventory assessment (neutron, gamma-ray emission, decay heat, radionuclide inventory, elemental content) and in terms of out-of-core fuel performance (cladding performance and fuel integrity in view of the safety criteria for SNF interim storage, transport and canister packaging) using several numerical and experimental approaches and methodologies is presented. This SOTA report is a result of the spent fuel characterisation (SFC) work package as part of the European Joint Programme on Radioactive Waste Management (EURAD), which offers an overview of the status of knowledge in the field of SNF characterisation and assessment during the pre-disposal phase. The document aims to focus on the current safety-significant gaps and related challenges, providing a direct link to the goals of the mandated actors of EURAD. The report is expected to be used by all EURAD colleagues in their national programmes as a key resource for knowledge management programmes and to contribute to demonstrating and documenting the state-of-the-art. KW - Radioactive waste management KW - Spent fuel characterisation KW - Extended interim storage KW - Predisposal PY - 2022 UR - https://www.ejp-eurad.eu/publications/eurad-d81-state-art-report SP - 1 EP - 112 PB - Agence Nationale pour la Gestion des Déchets Radioactifs (ANDRA) CY - Châtenay-Malabry AN - OPUS4-59154 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 - CONF A1 - Feldkamp, Martin 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 - Feldkamp, Martin T1 - Outcomes of three large scale fire reference tests conducted in BAM 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 - Pressure Vessels & Piping Virtual Conference 2020 CY - Online meeting DA - 03.08.2020 KW - Fire KW - Testing KW - Convection coefficient KW - Large scale fire testing KW - Propane gas fire test facility PY - 2020 AN - OPUS4-51194 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin 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 - Feldkamp, Martin 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 - Feldkamp, Martin T1 - Fire Testing - Current Activities N2 - The presentation gives an overview over three current fire testing activities for testing transport packages for the transport of high-level radioactive material. 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 smoldering and influence the cask thermal behavior 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. A new test is designed to examine pressure build up and possible mechanical failures in an undamaged impact limiter with a diameter of 1.9 m. The test is designed to measure heat flux from the impact limiter in case of its ignition and burning. To furthermore examine the burning behavior of steel encapsulated wood piles, tests are prepared with wood fire containers and an infrared ignition source. A fire reference package for calorimetric tests was designed. The fire reference package design consists of a closed steel sheet cylinder with a length of 1,500 mm, an exterior diameter of 1,050 mm, and a wall thickness of 10 mm. The cylinder was filled with refractory insulation material and instrumented with thermocouples distributed all over the cylinder. The measured local steel sheet temperatures allow the determination of local as well as overall integral heat fluxes versus time and versus surface temperature. Currently, tests are planned with a new and relatively small fire reference package design. The new design will have a steel sheet cylinder with a length of 182 mm, an exterior diameter of 120 mm. Additionally heat flux sensors will be used in the new fire reference package test design. The propane gas fire will be adjusted with respect to the outcome of the fire tests to meet the requirements of the IAEA-fire. T2 - Technical Exchange IRSN – BAM Transport & Storage of Packages for Radioactive Material CY - Berlin, Germany DA - 08.09.2021 KW - Heat flux sensor KW - Propane gas fire KW - Wood fire PY - 2021 AN - OPUS4-53242 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin 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 - Fire Reference Test PY - 2024 AN - OPUS4-59680 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin T1 - Fire Reference Tests for Qualification of IAEA Fire N2 - A small cylindrical fire reference package was designed. A fire test setup was created using the fire reference package in accordance with IAEA standards. Four propane gas fire tests were conducted to determine heat fluxes into the fire reference package. The initial test setup was modified to create the final design of the fire test facility. This allowed for the heat flux to be adjusted to meet the regulatory IAEA fire qualification criteria. Furthermore, a numerical model of the fire reference package was created using boundary conditions derived from the experimental data. The simulation results demonstrated good agreement with the experimental data and provided additional insights. T2 - Technical Exchange IRSN – BAM Transport & Storage of Packages for Radioactive Material CY - Cadarache, France DA - 13.10.2022 KW - Fire KW - IAEA KW - Propane PY - 2022 AN - OPUS4-57251 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin T1 - Combustion Chamber Design for Encapsulated Wood-Component Testing N2 - eavy-weight packages for the safe transport of radioactive material are equippedwith impact limiters often built ofwood-filled steel sheet structures to fulfil the requirements of the International Atomic Energy Agency (IAEA) regulations. The requirements definemechanical tests followed by a thermal test, including criteria ensuring the package design’s ability to withstand severe accidents and provide a high level of technical safety. Impact limiters are a package component mainly designed for the packages to withstand severe mechanical accident scenarios. In drop tests the impact limiters absorb the kinetic energy during impact of the package. The package must then - with its pre-damaged impact limiters - endure a thermal test defined precisely in the IAEA regulations as a 30-min fully engulfing 800 °C-fire. After the fire, a wood-filled impact limiter may continue to release thermal energy from an ongoing combustion process, defining relevant package temperatures. The energy flow from a possible burning impact limiter to the package is important for the safety evaluation of transport packages. To investigate the combustion behaviour of densely packed layers of spruce wood, encapsulated in pre-damaged cylindrical metal enclosures, a test set-up has been realised. The set-up consists of a combustion chamber to perform these tests under defined boundary conditions. The temperature development of the test specimens will be observed fromoutside with a thermographic imager, with HD-Cameras, and the mass loss will be measured during the entire test. Airflow conditions in the combustion chamber are analysed using Computational Fluid Dynamics (CFD) calculations in OpenFOAM. The planned combustion test setup is described. T2 - Wood & Fire Safety 2024 CY - Štrbské Pleso, Slovakia DA - 12.05.2024 KW - IAEA KW - Combustion Chamber KW - Smouldering KW - Wood PY - 2024 AN - OPUS4-60227 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin T1 - Evaluation of Convective Heat Transfer Coefficients with CFD for Heat Flux Calculation in Combustion Chamber N2 - Packages for the transport of high-level radioactive waste are designed to withstand severe accidents. To obtain approval for transport, these packages must adhere to the specification-based criteria of regulations established by the International Atomic Energy Agency (IAEA). To ensure compliance with the regulations, mechanical and thermal tests need to be conducted regarding the package type. The requirements define mechanical tests followed by a thermal test, including criteria ensuring the package design’s ability to withstand severe accidents. Heavy-weight packages for the safe transport of radioactive materials are equipped with impact limiters, which are often built with porous materials such as densely packed wood reinforced by steel sheet structures. These components absorb the kinetic energy during the impact of the package in drop tests and thus dampen the acceleration of other package components which supports the package to meet the requirements of the IAEA regulations. Following the mechanical tests, the package must, with its predamaged impact limiters, endure a thermal test defined precisely in the IAEA regulations. The thermal test is defined as a 30-min, fully engulfing 800 °C fire and a following time under ambient conditions for a sufficient period to ensure that temperatures in the specimen decrease in all parts of the specimen. During and following the thermal test, the specimen shall not be artificially cooled, and any combustion of materials of the specimen shall be permitted to proceed naturally. A wood-filled impact limiter can continue to release thermal energy during an ongoing combustion process, thus defining relevant package temperatures. Heat flux from a potentially burning impact limiter to the package is important for the safety evaluation of transport packages. A test setup was developed to approach the energy flow investigation and examine the combustion behaviour of porous materials encapsulated in predamaged cylindrical metal enclosures under various conditions. The setup consists of a combustion chamber for thermal tests under adjustable and defined boundary conditions. The temperature development of the test specimens can be observed from outside using a thermographic imager with high-definition cameras, and the mass loss of the test specimen can be measured in the combustion chamber. Convective heat transfer coefficients for various boundary conditions must be defined for use with experimentally gathered test specimen surface temperature data for heat flux evaluations. The airflow conditions in the combustion chamber were analysed using computational fluid dynamics (CFD) calculations in OpenFOAM with respect to the convective heat transfer coefficients at the surface of a hot test specimen. A convergence study was performed, and sensitivity analyses for different test specimen surface temperatures and exhaust gas volume flows were conducted. T2 - ASME PVP2025, Pressure Vessels & Piping Conference CY - Montréal, Quebec, Canada DA - 20.07.2025 KW - Heat Transfer KW - Convective Heat Transfer Coefficient KW - Combustion KW - OpenFOAM KW - Computational Fluid Dynamics (CFD) PY - 2025 AN - OPUS4-63902 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin A1 - Gleim, Tobias A1 - Quercetti, Thomas A1 - Wille, Frank T1 - Combustion Chamber Design for Encapsulated Wood-Component Testing N2 - Heavy-weight packages for the safe transport of radioactive material are equippedwith impact limiters often built ofwood-filled steel sheet structures to fulfil the requirements of the International Atomic Energy Agency (IAEA) regulations. The requirements definemechanical tests followed by a thermal test, including criteria ensuring the package design’s ability to withstand severe accidents and provide a high level of technical safety. Impact limiters are a package component mainly designed for the packages to withstand severe mechanical accident scenarios. In drop tests the impact limiters absorb the kinetic energy during impact of the package. The package must then - with its pre-damaged impact limiters - endure a thermal test defined precisely in the IAEA regulations as a 30-min fully engulfing 800 °C-fire. After the fire, a wood-filled impact limiter may continue to release thermal energy from an ongoing combustion process, defining relevant package temperatures. The energy flow from a possible burning impact limiter to the package is important for the safety evaluation of transport packages. To investigate the combustion behaviour of densely packed layers of spruce wood, encapsulated in pre-damaged cylindrical metal enclosures, a test set-up has been realised. The set-up consists of a combustion chamber to perform these tests under defined boundary conditions. The temperature development of the test specimens will be observed fromoutside with a thermographic imager, with HD-Cameras, and the mass loss will be measured during the entire test. Airflow conditions in the combustion chamber are analysed using Computational Fluid Dynamics (CFD) calculations in OpenFOAM. The planned combustion test setup is described. T2 - Wood & Fire Safety 2024 CY - Štrbské Pleso, Slovakia DA - 12.05.2024 KW - Fire KW - Combustion KW - Smouldering PY - 2024 SN - 978-3-031-59176-1 SN - 978-3-031-59179-2 SN - 978-3-031-59177-8 DO - https://doi.org/10.1007/978-3-031-59177-8 SP - 215 EP - 222 PB - Springer Nature Switzerland AG CY - Gewerbestrasse 11, 6330 Cham, Switzerland AN - OPUS4-60226 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin A1 - Gleim, Tobias A1 - Quercetti, Thomas A1 - Wille, Frank T1 - Evaluation of Convective Heat Transfer Coefficients with CFD for Heat Flux Calculation in Combustion Chamber N2 - Packages for the transport of high-level radioactive waste are designed to withstand severe accidents. To obtain approval for transport, these packages must adhere to the specification-based criteria of regulations established by the International Atomic Energy Agency (IAEA). To ensure compliance with the regulations, mechanical and thermal tests need to be conducted regarding the package type. The requirements define mechanical tests followed by a thermal test, including criteria ensuring the package design’s ability to withstand severe accidents. Heavy-weight packages for the safe transport of radioactive materials are equipped with impact limiters, which are often built with porous materials such as densely packed wood reinforced by steel sheet structures. These components absorb the kinetic energy during the impact of the package in drop tests and thus dampen the acceleration of other package components which supports the package to meet the requirements of the IAEA regulations. Following the mechanical tests, the package must, with its predamaged impact limiters, endure a thermal test defined precisely in the IAEA regulations. The thermal test is defined as a 30-min, fully engulfing 800 °C fire and a following time under ambient conditions for a sufficient period to ensure that temperatures in the specimen decrease in all parts of the specimen. During and following the thermal test, the specimen shall not be artificially cooled, and any combustion of materials of the specimen shall be permitted to proceed naturally. A wood-filled impact limiter can continue to release thermal energy during an ongoing combustion process, thus defining relevant package temperatures. Heat flux from a potentially burning impact limiter to the package is important for the safety evaluation of transport packages. A test setup was developed to approach the energy flow investigation and examine the combustion behaviour of porous materials encapsulated in predamaged cylindrical metal enclosures under various conditions. The setup consists of a combustion chamber for thermal tests under adjustable and defined boundary conditions. The temperature development of the test specimens can be observed from outside using a thermographic imager with high-definition cameras, and the mass loss of the test specimen can be measured in the combustion chamber. Convective heat transfer coefficients for various boundary conditions must be defined for use with experimentally gathered test specimen surface temperature data for heat flux evaluations. The airflow conditions in the combustion chamber were analysed using computational fluid dynamics (CFD) calculations in OpenFOAM with respect to the convective heat transfer coefficients at the surface of a hot test specimen. A convergence study was performed, and sensitivity analyses for different test specimen surface temperatures and exhaust gas volume flows were conducted. T2 - ASME PVP2025, Pressure Vessels & Piping Conference CY - Montreal, Quebec, Canada DA - 20.07.2025 KW - Heat Transfer KW - Convection KW - OpenFOAM KW - Combustion PY - 2025 SP - 1 EP - 7 PB - American society of mechanical engineers (ASME) AN - OPUS4-63901 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin A1 - Quercetti, Thomas A1 - Gleim, Tobias T1 - Temperature and Heat Flux Measurements in Fire Testing N2 - Packages for the transport of Spent Nuclear Fuel and high active radioactive waste are designed to withstand severe accidents. Specific mechanical and thermal tests are defined in the IAEA Regulations for the Safe Transport of Radioactive Materials in order to cover these hypothetical severe accidents. The thermal test mainly consists of a 30 minute fully engulfing fire. Components such as the package impact limiters can lead to supplementary energy release during the thermal test as they might continue burning after the fire phase. Local heat flux into the package can occur. Measurement of heat flux into the package is of importance to evaluate component temperatures and review their acceptance. The usability of heat flux sensors and temperature measurement equipment has to be tested to apply them in impact limiter fire testing. Further questions arise such as :“Is the infrared camera a useful tool to determine heat flux at the boundary surface?” T2 - Technical Exchange IRSN – BAM Transport & Storage of Packages for Radioactive Material CY - Berlin, Germany DA - 08.09.2021 KW - Heat flux sensor PY - 2021 AN - OPUS4-55659 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 - Feldkamp, Martin A1 - Quercetti, Thomas A1 - Wille, Frank T1 - Outcomes of three large scale fire reference tests conducted in BAM 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 - Pressure Vessels & Piping Conference 2020 CY - Online meeting DA - 03.08.2020 KW - Fire KW - Testing KW - Large scale testing KW - Calorimeter KW - Heat flux PY - 2020 SP - 1 EP - 9 PB - ASME CY - New York AN - OPUS4-51192 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gaddampally, Mohan Reddy T1 - Failure Analysis on Irradiated Claddings Subjected to Long-Term Dry Interim Storage N2 - Long-term dry interim storage may adversely affect the mechanical properties of spent fuel rods, possibly resulting in a reduced resilience during handling or transport after storage. Since the cladding is the first barrier for the spent fuel pellets, its integrity must be demonstrated until the end of interim storage and subsequent transportation. An established method for characterizing the cladding material is the ring compression test, in which a small, cylindrical sample of the cladding tube is subjected to a compressive load. This test is a laboratory representation of a load case where the fuel rod is crushed. Pre-storage drying and the early stage of interim storage can subject the cladding to higher temperatures and higher pressure induced tensile hoop stresses than those associated with in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate in zirconium-based alloys (Zircaloy) during slow cooling, which result in embrittlement of the cladding material and eventually a possible sudden failure of cladding integrity under additional mechanical loads. Especially long, continuous radial hydride structures and low temperature can cause severe embrittlement of claddings and finally failure by fracture even at small deformations. Therefore, the study of hydride morphology plays an important role in describing the brittle failure behaviour of the claddings. The focus of the presented research is on the development of appropriate numerical methods for predicting the mechanical behaviour and identification of limiting conditions to prevent brittle fracture of Zircaloy claddings. Typical hydride morphologies are shown. An iterative inverse analysis method is described for deriving the elastic-plastic material properties in the hoop direction of a ring-shaped sample. A modelling approach based on cohesive zones is explained which is able to reproduce the propagation of cracks initiated at radial hydrides in the zirconium matrix. The developed methods are applied to defueled samples of cladding alloy ZIRLO®, which were subjected to a thermo-mechanical treatment to reorient existing circumferential hydrides to radial hydrides. A selected sample showing sudden load drops during a quasi-static ring compression test is analysed by means of fracture mechanics for illustrative purposes. This project as part of the European Joint Programme on Radioactive Waste Management has received funding from the European Union’s Horizon 2020 research and innovation programme under grant agreement no. 847593. T2 - BAM-Kolloquium der Abteilung 3 CY - Berlin, Germany DA - 05.06.2023 KW - Cladding Embrittlement KW - Cohesive Zone Modelling KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2023 AN - OPUS4-57598 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gaddampally, Mohan Reddy T1 - Cohesive Zone Modelling Approach on Irradiated Claddings Subjected to Long-Term Dry Interim Storage N2 - Long-term dry interim storage may adversely affect the mechanical properties of spent fuel rods, possibly resulting in a reduced resilience during handling or transport after storage. Pre-storage drying and the early stage of interim storage can subject the cladding to higher temperatures and higher pressure induced tensile hoop stresses than those associated with in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate in zirconium-based alloys (Zircaloy) during slow cooling, which may result in embrittlement of the cladding material and eventually a sudden failure of cladding under additional mechanical loads. Especially long, continuous radial hydride structures and low temperature can cause severe embrittlement of claddings and finally failure by fracture even at small deformations. The focus of the presented research is on the development of appropriate numerical methods for predicting the mechanical behaviour and identification of limiting conditions to prevent brittle fracture of Zircaloy claddings. An iterative inverse analysis method is used for deriving the elastic-plastic material properties in the hoop direction of a ring-shaped sample. A modelling approach based on cohesive zones is explained which can reproduce the propagation of cracks initiated at radial hydrides in the zirconium matrix. The developed methods are applied to defueled samples of cladding alloy ZIRLO®, which were subjected to a thermo-mechanical treatment to reorient existing circumferential hydrides to radial hydrides. A selected sample showing sudden load drops during a quasi-static ring compression test is analysed by means of fracture mechanics for illustrative purposes. T2 - 7th GRS Workshop on the Safety of Extended Dry Storage of Spent Nuclear Fuel CY - Garching, Germany DA - 24.05.2023 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2023 AN - OPUS4-58553 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias T1 - Polyurethane Foam in Impact Limiters - Experimental and Numerical Analysis N2 - Transport containers with radioactive material are usually shipped in Germany with wooden-filled impact limiters. Alternative energy-absorbing materials for this purpose can be polyurethane foam, for example. In order to perform adequate simulations with PU foam, experimental and numerical investigations must be carried out. With the help of a series of experimental test runs, a material model is developed, tested and compared in LS-Dyna. T2 - Technical Exchange IRSN – BAM: Transport & Storage of Packages for Radioactive Material CY - Berlin, Germany DA - 08.09.2021 KW - Polyurethane Foam, Impact Limiters, PU-Foam PY - 2021 AN - OPUS4-53261 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias T1 - Polyurethane Foam for Impact Limiters - Theory and Experimental Data N2 - Transport containers with radioactive material are usually shipped in Germany with wooden-filled impact limiters. Alternative energy-absorbing materials for this purpose can be polyurethane foam, for example. In order to adequately determine the characteristics for a selected PU foam, various experimental investigations must be carried out. Classically, PU foam specimens are tested under uniaxial compression, confined compression and a tensile test. Using the experimental data and selected material models from the literature, the parameters can be determined to describe and apply material models. T2 - Technical Exchange Sandia – BAM: Transport & Storage of Packages for Radioactive Material CY - Online meeting DA - 31.08.2021 KW - Polyurethane Foam KW - Impact Limiters KW - PU-Foam PY - 2021 AN - OPUS4-53262 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias T1 - Experimental And Numerical Analyses For The Evaluation Of Heat Fluxes Of A Fire Reference Test N2 - Packages for the transport of high-level radioactive material are designed to endure severe accidents. To obtain approval, these transport packages must adhere to the specification-based criteria of the IAEA SSR-6 [1]. To ensure compliance with these requirements, specific mechanical and thermal tests need to be appointed with respect to the package type. Typically, IAEA SSR-6 [1] prescribes mechanical tests followed by a thermal fire test as part of a cumulative test sequence. To approve the fire test, BAM uses a reference package representing the original package in its outer geometry for characterizing the actual fire and its impact on the package. This serves a dual purpose: it allows the precise adjustment of experimental parameters for the package design to be approved, and it provides input parameters for thermomechanical simulations, cf. [2]. With the help of this methodology for characterizing boundary conditions of the package, temperature evolutions within the reference package can be studied in a Finite-Element Analyses (FEA). This helps to compare the experiments with the numerical simulations on the case of the reference package, but also serves at the same time for preliminary simulations with the package design to be approved. The thermal test consists of a 30-minute fully engulfing 800°C pool fire or an equally severe fire, e.g. a propane gas fire. The fire reference test is performed prior to the regulatory fire test with the package design to be approved. The fire reference package in the described case is a closed sheet 316L steel 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 insulating material. T2 - ASME PVP 2024 CY - Bellevue, WA, USA DA - 28.07.2024 KW - Experimental KW - Numerical KW - Heat Flux KW - Reference Test PY - 2024 AN - OPUS4-60847 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias T1 - Safety Aspects for Transport Containers with regard to Ageing and Battery Accident Scenarios N2 - In Germany, the search for a final repository for highly radioactive material is currently based on the assumption that interim storage will take considerably longer than was anticipated. At that time, authorization procedures assumed that interim storage would be brief, until the containment systems could be placed in a final repository. For today's new approvals and extensions, the IAEA (International Atomic Energy Agency) regulations for the transport of radioactive material require an assessment of all components and component groups with regard to ageing. To assess the safety requirements not only over the previously envisaged periods but also over long durations, numerous individual parts and component groups must be examined and evaluated for ageing and interactions between materials. The metal seals of a double lid sealing system, for example, are critical components that must be studied with respect to ageing. In a containment system for radioactive material, metal seals must not only undergo classic ageing, but also withstand mechanical influences and radiation over extended periods. In addition to assessments by the applicants, BAM must also independently analyze and evaluate these ageing mechanisms as part of its sovereign mandate. Another challenge in both new approvals and extensions is the assessment of transport regarding changes in the state of the art, which are not synchronously assessed in the regulations. In transporting radioactive material, a classic oil fire has been assumed in a hypothetical accident scenario from the development of the regulations to the present day. Due to the rapid development of propulsion technologies in recent years, such as battery or hydrogen-powered vehicles, investigations must be conducted in the revision process of the IAEA regulations to assess the impact of these new technologies on safety assessments. It is important for the member states of the IAEA to know whether new technologies are covered by the regulations or if changes are necessary to continue issuing international licenses. By focusing on batteries from the material to the cell to the module and the application, BAM contributes to the evaluation of batteries at every step of the chain. In addition to the safety of current and new battery technologies, BAM also endeavors to address issues of compositional trace and sustainable energy materials. T2 - MSD: Leaf CY - Livermore, CA , USA DA - 07.08.2024 KW - Safety Aspects KW - Ageing KW - Accident Scenarios KW - Battery PY - 2024 AN - OPUS4-60851 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias T1 - Enhancement of Fire Test Stand Performance at Test Site of BAM: Installation and Evaluation of an Augmented System with a Fire Reference Package N2 - Packages for the transport of radioactive material are designed to en-dure severe accidents. Packages for the transportation of radioactive material must demonstrate that the package can withstand certain prescribed tests from the IAEA Regulations [1]. In addition to mechanical tests, a thermal test in form of a fire test must be carried out. As packages to be tested at BAM are signifi-cantly larger than previous package designs, BAM has expanded an existing fire test stand. A modular concept is chosen, which means that the arrangement of the burner nozzles can be adapted to the test specimen. The dimensions of the burner rings, the type, the orientation and the number of burner nozzles can be varied depending on the test specimen. In addition, various pumps can be used to set the corresponding mass flow. With the help of a calorimeter test, the fire test stand can be qualified for a specific size of packages regarding the boundary conditions of the IAEA Regulations [1]. Due to the typically wood filled impact limiters in German package designs, a fire test is necessary, as experiments have shown that possible openings that occurred during a mechanical test contributed to the igni-tion of the wood filled impact limiters within the prescribed 30 minutes of the IAEA Regulations [1]. From a series of experiments, two experiments are pre-sented to show the possibilities to obtain different temperatures and temperature rates in the test specimen. In addition to sensor data, the heat flux into the package is calculated to verify that the IAEA boundary conditions are satisfied. In addition to the temperature data, other data such as wind speed and wind direction are also recorded to explain subsequent effects in the measurement data in a comprehen-sible manner. T2 - Wood & Fire Safety 2024 CY - Štrbské Pleso, Slovakia DA - 13.05.2024 KW - Fire Test Stand KW - Accident Scenario KW - Fire Qualification KW - IAEA Regu-lations PY - 2024 UR - https://link.springer.com/book/9783031591761 AN - OPUS4-60104 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias T1 - Potential Effects of Battery and Hydrogen Fires regarding Regulatory Requirements N2 - Introduction and Necessity of the Investigation The IAEA regulations for the safe transport of radioactive material (IAEA SSR-6) define the safety requirements for different package types and consider different transport conditions. The accident conditions of transport specify different mechanical and thermal tests based on investigations of real accident scenarios. Considering the rapid development of new boundary conditions of transport such as electric mobility and the use of hydrogen as energy source for trucks and other kind of vehicles, potential effects of battery and hydrogen fires in transport accidents should be investigated. The aim is to evaluate the existing test requirements developed and derived decades ago, whether they are covering the current transport situation. This concept paper will briefly present the reasons for detailed investigations as bases for a coordinated research project under the roof of the IAEA. T2 - Technical Exchange IRSN – BAM Transport & Storage of Packages for Radioactive Material CY - Berlin, Germany DA - 04.06.2024 KW - Fire KW - Battery KW - Hydrogen KW - IAEA Regulations PY - 2024 AN - OPUS4-60338 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias T1 - Fracture Mechanical Investigations on a Welding Seam of a Thick-Walled Transport Package N2 - Untersuchung einer Schweißnaht mit verschiedenen Codes. Vergleiche von experimentellen und numerischen Ergebnissen T2 - Technical Exchange IRSN – BAM: Transport & Storage of Packages for Radioactive Material CY - Cadarache, France DA - 13.10.2022 KW - Drop test KW - Fracture initiation KW - Transport package PY - 2022 AN - OPUS4-56055 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias ED - Monavari, Mehran ED - Bertovic, Marija ED - Wille, Frank T1 - AI-Based Documentation Analysis for Safety Assessment of Packages for Radioactive Material N2 - The transportation of radioactive material requires, dependent on type and quantity of the radioactive material, a regulatory approval based on the package type. Safety assessments shall be conducted in compliance with the International Atomic Energy Agency (IAEA) regulations and documented in a comprehensive package design safety report to obtain approval from authority. This comprehensive safety report evaluates a broad range of requirements from the regulations, including mechanical, thermal, shielding, criticality and transport requirements and controls, and testing assessments. Additionally, it encompasses supporting documents such as specifications, inspections, certifications, drawings, and guidelines in a variety of complex documents. Safety and manufacturing reports contain multiple interconnected sub-reports covering various topics. Changes, such as component modifications, material property updates, or regulatory revisions, often impact multiple sections of the safety analysis reports, making even minor adjustments complex and time-consuming. Each transport package has unique requirements to be fulfilled, making every safety report distinct, despite following the same regulatory framework. Most documentation exists in standard digital formats but is often not machine interpretable, preventing automated analysis of the critical dependencies between them. This paper argues that moving beyond simple digitization towards structured knowledge representation is essential for addressing these challenges. We propose a multi-stage approach, beginning with foundational AI technologies such as Large Language Models (LLMs) and Retrieval-Augmented Generation (RAG), and progressing toward the construction of Knowledge Graphs (KGs). KGs convert unstructured and semi-structured information into a connected, queryable network, enabling precise tracing and visualization of complex interdependencies within the documentation landscape. By linking interpretable content directly to datasheets, tables, simulations, experimental results, standards, and regulations, such a system would automatically identify changes and interdependencies. Related conditions could be validated using AI-based tools, reducing the need for manual intervention, improving both efficiency and safety. Human error plays a significant role in drafting, reviewing, and revising safety reports, often requiring iterative review cycles and multiple reviewers before approval. A digital quality infrastructure could reduce iterations and further improve efficiency. Integrating AI into this process could optimize safety assessments and enhance their robustness by leveraging interpretability to enhance safety. This preliminary study explores the readiness and requirements for using intelligent documentation analysis system in the context of regulatory compliance for package safety for the transport of radioactive material. By analysing current documentation workflows, we identify how LLM-based tools can interpret complex safety reports and highlight critical interdependencies and then demonstrate why a KG-based architecture is necessary to robustly manage and query critical interdependencies. This lays the groundwork for future agentic AI systems capable of proactively supporting the safety assessment lifecycle, while stressing the importance of robust data governance and AI reliability in this highly regulated context. T2 - PATRAM 2025 CY - San Antonio, TX, USA DA - 27.07.2025 KW - AI, RAG, LLM, Knowledge Graph PY - 2025 SP - 1 EP - 10 PB - Institute of Nuclear Materials Management (INMM) AN - OPUS4-63863 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias T1 - AI-Driven Documentation Analysis Supporting Safety Assessment of Transport Packages N2 - The transportation of radioactive material requires, dependent on type and quantity of the radioactive material, a regulatory approval based on the package type. Safety assessments shall be conducted in compliance with the International Atomic Energy Agency (IAEA) regulations and documented in a comprehensive package design safety report to obtain approval from authority. This comprehensive safety report evaluates a broad range of requirements from the regulations, including mechanical, thermal, shielding, criticality and transport requirements and controls, and testing assessments. Additionally, it encompasses supporting documents such as specifications, inspections, certifications, drawings, and guidelines in a variety of complex documents. Safety and manufacturing reports contain multiple interconnected sub-reports covering various topics. Changes, such as component modifications, material property updates, or regulatory revisions, often impact multiple sections of the safety analysis reports, making even minor adjustments complex and time-consuming. Each transport package has unique requirements to be fulfilled, making every safety report distinct, despite following the same regulatory framework. Most documentation exists in standard digital formats but is often not machine interpretable, preventing automated analysis of the critical dependencies between them. This paper argues that moving beyond simple digitization towards structured knowledge representation is essential for addressing these challenges. We propose a multi-stage approach, beginning with foundational AI technologies such as Large Language Models (LLMs) and Retrieval-Augmented Generation (RAG), and progressing toward the construction of Knowledge Graphs (KGs). KGs convert unstructured and semi-structured information into a connected, queryable network, enabling precise tracing and visualization of complex interdependencies within the documentation landscape. By linking interpretable content directly to datasheets, tables, simulations, experimental results, standards, and regulations, such a system would automatically identify changes and interdependencies. Related conditions could be validated using AI-based tools, reducing the need for manual intervention, improving both efficiency and safety. Human error plays a significant role in drafting, reviewing, and revising safety reports, often requiring iterative review cycles and multiple reviewers before approval. A digital quality infrastructure could reduce iterations and further improve efficiency. Integrating AI into this process could optimize safety assessments and enhance their robustness by leveraging interpretability to enhance safety. This preliminary study explores the readiness and requirements for using intelligent documentation analysis system in the context of regulatory compliance for package safety for the transport of radioactive material. By analysing current documentation workflows, we identify how LLM-based tools can interpret complex safety reports and highlight critical interdependencies and then demonstrate why a KG-based architecture is necessary to robustly manage and query critical interdependencies. This lays the groundwork for future agentic AI systems capable of proactively supporting the safety assessment lifecycle, while stressing the importance of robust data governance and AI reliability in this highly regulated context. T2 - Sandia-BAM Technical Exchange CY - Albuquerque, NM, USA DA - 23.07.2025 KW - AI, RAG, LLM, Knowledge Graph PY - 2025 AN - OPUS4-63860 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias T1 - Enhancement of Fire Test Stand Performance at Test Site of BAM: Installation and Evaluation of an Augmented System with a Fire Reference Package N2 - Packages for the transport of radioactive material are designed to endure severe accidents. Packages for the transportation of radioactive material must demon-strate that the package can withstand certain prescribed tests from the IAEA Reg-ulations [1]. In addition to mechanical tests, a thermal test in form of a fire test must be carried out. As packages to be tested at BAM are significantly larger than previous package designs, BAM has expanded an existing fire test stand. A modular concept is chosen, which means that the arrangement of the burner noz-zles can be adapted to the test specimen. The dimensions of the burner rings, the type, the orientation and the number of burner nozzles can be varied depending on the test specimen. In addition, various pumps can be used to set the corre-sponding mass flow. With the help of a calorimeter test, the fire test stand can be qualified for a specific size of packages regarding the boundary conditions of the IAEA Regulations [1]. Due to the typically wood filled impact limiters in German package designs, a fire test is necessary, as experiments have shown that possible openings that occurred during a mechanical test contributed to the ignition of the wood filled impact limiters within the prescribed 30 minutes of the IAEA Regula-tions [1]. From a series of experiments, two experiments are presented to show the possibilities to obtain different temperatures and temperature rates in the test specimen. In addition to sensor data, the heat flux into the package is calculated to verify that the IAEA boundary conditions are satisfied. In addition to the tem-perature data, other data such as wind speed and wind direction are also recorded to explain subsequent effects in the measurement data in a comprehensible man-ner. T2 - Sandia-BAM Technical Exchange CY - Albuquerque, NM, USA DA - 23.07.2025 KW - IAEA Regulations KW - Fire Test Stand KW - Accident Scenario KW - Fire Qualification PY - 2025 AN - OPUS4-63861 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias T1 - AI-Based Documentation Analysis for Safety Assessment of Packages for Radioactive Material N2 - The transportation of radioactive material requires, dependent on type and quantity of the radioactive material, a regulatory approval based on the package type. Safety assessments shall be conducted in compliance with the International Atomic Energy Agency (IAEA) regulations and documented in a comprehensive package design safety report to obtain approval from authority. This comprehensive safety report evaluates a broad range of requirements from the regulations, including mechanical, thermal, shielding, criticality and transport requirements and controls, and testing assessments. Additionally, it encompasses supporting documents such as specifications, inspections, certifications, drawings, and guidelines in a variety of complex documents. Safety and manufacturing reports contain multiple interconnected sub-reports covering various topics. Changes, such as component modifications, material property updates, or regulatory revisions, often impact multiple sections of the safety analysis reports, making even minor adjustments complex and time-consuming. Each transport package has unique requirements to be fulfilled, making every safety report distinct, despite following the same regulatory framework. Most documentation exists in standard digital formats but is often not machine interpretable, preventing automated analysis of the critical dependencies between them. This paper argues that moving beyond simple digitization towards structured knowledge representation is essential for addressing these challenges. We propose a multi-stage approach, beginning with foundational AI technologies such as Large Language Models (LLMs) and Retrieval-Augmented Generation (RAG), and progressing toward the construction of Knowledge Graphs (KGs). KGs convert unstructured and semi-structured information into a connected, queryable network, enabling precise tracing and visualization of complex interdependencies within the documentation landscape. By linking interpretable content directly to datasheets, tables, simulations, experimental results, standards, and regulations, such a system would automatically identify changes and interdependencies. Related conditions could be validated using AI-based tools, reducing the need for manual intervention, improving both efficiency and safety. Human error plays a significant role in drafting, reviewing, and revising safety reports, often requiring iterative review cycles and multiple reviewers before approval. A digital quality infrastructure could reduce iterations and further improve efficiency. Integrating AI into this process could optimize safety assessments and enhance their robustness by leveraging interpretability to enhance safety. This preliminary study explores the readiness and requirements for using intelligent documentation analysis system in the context of regulatory compliance for package safety for the transport of radioactive material. By analysing current documentation workflows, we identify how LLM-based tools can interpret complex safety reports and highlight critical interdependencies and then demonstrate why a KG-based architecture is necessary to robustly manage and query critical interdependencies. This lays the groundwork for future agentic AI systems capable of proactively supporting the safety assessment lifecycle, while stressing the importance of robust data governance and AI reliability in this highly regulated context. T2 - PATRAM 2025 CY - San Antonio, TX, USA DA - 27.07.2025 KW - Knowledge Graph KW - AI KW - RAG KW - LLM PY - 2025 AN - OPUS4-63862 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias T1 - Potential Effects of Battery and Hydrogen Fires regarding Regulatory Requirements N2 - Die internationalen Sicherheitsanforderungen für den Transport radioaktiver Stoffe sind in den IAEO-Vorschriften SSR-6 festgelegt. Diese wurden in den 1960er Jahren entwickelt und seitdem nur in begrenztem Umfang angepasst. Sie bilden die Basis für die Regelwerke der verschiedenen Verkehrsträger – darunter auch das ADR (Europäisches Übereinkommen über die internationale Beförderung gefährlicher Güter auf der Straße). Die SSR-6 wird regelmäßig vom TRANSSC-Gremium der IAEO überprüft und aktualisiert. In diesen Prozess fließen aktuelle wissenschaftliche Erkenntnisse sowie praktische Erfahrungen ein, um die Sicherheit beim Transport radioaktiver Materialien kontinuierlich zu verbessern. Im Rahmen der 47. TRANSSC-Sitzung stellte die BAM (Bundesanstalt für Materialforschung und -prüfung) aktuelle Forschungsergebnisse zu Unfallszenarien mit Batterien und Wasserstoff vor und adressierte zentrale sicherheitsrelevante Fragestellungen. Aufgrund des großen internationalen Interesses, insbesondere da vergleichbare Erkenntnisse aus anderen UN-Mitgliedstaaten bislang nicht vorlagen, wurden diese Fragestellungen ab der 48. Sitzung offiziell in das Arbeitsprogramm aufgenommen. Vor der 48. TRANSSC-Sitzung führten BAM und BASE eine gemeinsame Befragung aller UN-Mitgliedstaaten mittels eines Fragebogens durch. Die Ergebnisse wurden im Rahmen der 48. TRANSSC-Sitzung vorgestellt. Dabei zeigte sich, dass in den befragten Ländern bislang keine spezifischen nationalen Regelungen oder Einschränkungen im Hinblick auf Fahrzeuge mit alternativen Antrieben bestehen. Gleichzeitig wurde jedoch auf ein potenzielles Risiko mit derzeit nicht abschätzbarem Ausmaß hingewiesen. N2 - Hier: The IAEA regulations for the safe transport of radioactive material (IAEA SSR-6) define the safety requirements for different package types and consider different transport conditions. The accident conditions of transport specify different mechanical and thermal tests based on investigations of real accident scenarios. Considering the rapid development of new boundary conditions of transport such as electric mobility and the use of hydrogen as energy source for trucks and other kind of vehicles, potential effects of battery and hydrogen fires in transport accidents should be investigated. The aim is to evaluate the existing test requirements developed and derived decades ago, whether they are covering the current transport situation. This presentation will briefly present the reasons for detailed investigations as bases for a coordinated research project under the roof of the IAEA. T2 - WP.15 IWG-EV 22nd session CY - Online-Meeting DA - 23.07.2025 KW - Requirements KW - Battery Fire KW - Experiment KW - IAEA PY - 2025 AN - OPUS4-63900 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias T1 - Welding Seam Safety Evaluation in a Thick-Walled Type B Transport Package N2 - The safety demonstration of heavy weight type B transport packages used for storing spent nuclear fuel (SNF) or radioactive waste is ensured by a combination of physical testing and numerical calculations. While experiments are performed in accordance with the IAEA regulations for selected drop scenarios, Finite-Element-Method (FEM) simulations are used to predict the most damaging case and to investigate additional drop positions. BAM as competent authority in Germany has performed different investigations of a welding seam for a typical large transport package made of A508 forged steel, where the bottom plate is welded to the cylindrical shell. The package has a mass of approx. 120 t. Results of physical drop tests with a full-scale model and accompanying preliminary FEM simulations are presented to determine the decisive stresses in the welding seam. A drop test only represents one set of a package and test parameters. A further parameter analysis is considered to account for allowable variations of packaging properties (e.g. resulting from the manufacturing process) and, based on IAEA requirements, the temperature dependence of the material behaviour. The results of the stress analyses from the drop test and the simulation form the basis and provide the input parameters for a fracture mechanics analysis. In addition to the IAEA specifications, further standards are taken into account for an in-depth investigation, see R6 [1], BS 7910 [2] and API 579-1/ASME FFS1 [3]. All the above-mentioned standards require a manufacturer-specific defect analysis with respect to size and position. Both result from the welding process and the following heat treatment regime. The maximum defect sizes are ensured with non-destructive test methods (such as ultrasonic or particle methods) as integral part of the manufacturing process of the welding seam. Another important parameter in the welding process is the residual stress (secondary stress). The combination of the primary and secondary stress determines the total stress in the welding seam. The most damaging case of the welding seam is determined and evaluated with help of the above-mentioned standards and taking into account the IAEA requirements with respect to defect sizes, material properties, primary and residual stress, yield strength etc. T2 - PATRAM 22 - The International Symposium on the Packaging and Transportation of Radioactive Materials CY - Juan-Les-Pins, France DA - 11.06.2023 KW - Welding KW - Fracture Mechanics KW - Transport Package PY - 2023 AN - OPUS4-57695 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias T1 - AI-Based Documentation Analysis for Safety Assessment of Packages for Radioactive Material N2 - The transportation of radioactive material requires, dependent on type und quantity of the radioactive material, a regulatory approval based on the type of the package. Safety assessments shall be conducted in compliance with the IAEA regulations and documented in a comprehensive package design safety report to obtain approval from authority. This comprehensive safety report evaluates a broad range of requirements from the regulations, including mechanical, thermal, shielding, criticality and transport requirements and controls, and testing assessments. Additionally, it encompasses supporting documents such as specifications, inspections, certifications, drawings, and guidelines in a variety of complex documents. Safety and manufacturing reports contain multiple interconnected sub-reports covering various topics. Changes, such as component modifications, material property updates, or regulatory revisions, often impact multiple sections of the safety analysis reports, making even minor adjustments complex and time-consuming. Each transport package has unique requirements, making every safety report distinct, despite following the same regulatory framework. Most documentation exists in standard digital formats but is often not machine interpretable. This lack of interpretability means individuals must manually identify and address dependencies within the reports. An AI-based documentation system that visualizes these dependencies would significantly enhance the efficiency of safety assessments. By linking interpretable content directly to datasheets, tables, simulations, experimental results, standards, and regulations, such a system would automatically identify changes and interdependencies. Related conditions could be validated using AI-based tools, reducing the need for manual intervention, improving both efficiency and safety. Human error plays a significant role in drafting, reviewing, and revising safety reports, often requiring iterative review cycles and multiple reviewers before approval. A digital quality infrastructure could reduce iterations and further improve efficiency. Integrating AI into this process could optimize safety assessments and enhance their robustness by leveraging interpretability to enhance safety. This preliminary study explores the readiness and requirements for using LLMs and RAG in the context of regulatory compliance for transport of radioactive material. By analyzing current documentation workflows, we identify how LLM-based tools can interpret complex safety reports and highlight critical interdependencies. We stress the importance of robust data governance, confidentiality measures, and AI reliability in this highly regulated context. T2 - BAM – Fraunhofer Materials LLM Days CY - Würzburg, Germany DA - 03.07.2025 KW - KI KW - RAG KW - LLM PY - 2025 AN - OPUS4-63727 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias A1 - Affagard, Jean-Sebastien T1 - Towards a simplifiedmodel of the delayed impact: numerical and experimental outlook N2 - The IAEA Regulations for the Safe Transport of Radioactive Material No. SSR-6 require drop tests for demonstrating the ability of a package to withstand accident conditions of transport. In case a gap exists between the contents and the lid, it can result in a delayed impact of the content onto the lid system during a vertical drop test. The kinematic energy transmitted to the lid in such cases can be significant due to the stiffness of the impact, leading to high stresses in the bolts compared to a configuration without a gap. In the past, IRSN and BAM have individually investigated the modelling of the delayed impact phenomenon. BAM has examined the effects of the delayed impact phenomenon for spent fuel packages in a comparative analysis between experiments and simulations, which was presented at SMIRT 2013. Meanwhile, IRSN’s has developed a simple numerical tool that can quickly evaluate the influence of various parameters before considering a more complex finite element numerical calculation. This tool was presented at PATRAM 2022. In 2023, BAM and IRSN (now ASNR – French nuclear safety and radiation protection authority) decided to jointly pursue their study on the delayed impact phenomenon, incorporating damping effects, by developing new insights and techniques. To this end, a controlled and simplified set of experiments needs to be developed to calibrate and validate the simplified model. These experiments will record velocity, position of parts, strain and accelerations. The parameters and configurations necessary for the development of the mock-up, as well as the placement and quantity of sensors, are determined through extensive pre-calculations using both complex finite element models and the simplified approach. A key challenge involves precisely controlling the drop of the mock-up to accurately replicate the theoretical contact and damping conditions. Advanced sensor technologies will be employed to acquire reliable and sufficient data. The tests are designed and conducted at BAM's TTS drop test facility in collaboration with ASNR. T2 - PATRAM 2025 CY - San Antonio, TX, USA DA - 27.07.2025 KW - Experiments KW - Impact KW - Drop Tower PY - 2025 AN - OPUS4-63869 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Gleim, Tobias A1 - Chen, Hefeng ED - Kleiber, M. ED - Onate, E. T1 - High-Order Accurate Methods for the Numerical Analysis of a Levitation Device N2 - This paper establishes different axisymmetric and two-dimensional models 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−alpha 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. KW - Maxwell equations KW - Levitation KW - High-Order KW - Accurate-Numerical-Methods KW - Generalized Newmark-alpha PY - 2020 DO - https://doi.org/10.1007/s11831-020-09427-z SN - 1134-3060 VL - 28 IS - 3 SP - 1517 EP - 1543 PB - Springer Nature AN - OPUS4-50734 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 - Gleim, Tobias A1 - Feldkamp, Martin A1 - Quercetti, Thomas A1 - Musolff, André A1 - Werner, Jan A1 - Wille, Frank T1 - Enhancement of Fire Test Stand Performance at Test Site of BAM: Installation and Evaluation of an Augmented System with a Fire Reference Package N2 - Packages for the transport of radioactive material are designed to en-dure severe accidents. Packages for the transportation of radioactive material must demonstrate that the package can withstand certain prescribed tests from the IAEA Regulations [1]. In addition to mechanical tests, a thermal test in form of a fire test must be carried out. As packages to be tested at BAM are signifi-cantly larger than previous package designs, BAM has expanded an existing fire test stand. A modular concept is chosen, which means that the arrangement of the burner nozzles can be adapted to the test specimen. The dimensions of the burner rings, the type, the orientation and the number of burner nozzles can be varied depending on the test specimen. In addition, various pumps can be used to set the corresponding mass flow. With the help of a calorimeter test, the fire test stand can be qualified for a specific size of packages regarding the boundary conditions of the IAEA Regulations [1]. Due to the typically wood filled impact limiters in German package designs, a fire test is necessary, as experiments have shown that possible openings that occurred during a mechanical test contributed to the igni-tion of the wood filled impact limiters within the prescribed 30 minutes of the IAEA Regulations [1]. From a series of experiments, two experiments are pre-sented to show the possibilities to obtain different temperatures and temperature rates in the test specimen. In addition to sensor data, the heat flux into the package is calculated to verify that the IAEA boundary conditions are satisfied. In addition to the temperature data, other data such as wind speed and wind direction are also recorded to explain subsequent effects in the measurement data in a comprehen-sible manner. T2 - Wood & Fire Safety 2024 CY - Strbske Pleso, Slowakei DA - 13.05.2024 KW - IAEA Regu-lations KW - Fire Test Stand KW - Accident Scenario KW - Fire Qualification PY - 2024 UR - https://link.springer.com/book/9783031591761 SN - 978-3-031-59176-1 SN - 978-3-031-59179-2 SN - 978-3-031-59177-8 DO - https://doi.org/10.1007/978-3-031-59177-8_22 VL - 1 SP - 189 EP - 197 PB - Springer Cham AN - OPUS4-60101 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias A1 - Feldkamp, Martin A1 - Quercetti, Thomas A1 - Wille, Frank T1 - Experimental and Numerical Analyses for the Evaluation of Heat Fluxes of a Fire Reference Test N2 - Packages for the transport of high-level radioactive material are designed to endure severe accidents. To obtain approval, these transport packages must adhere to the specification-based criteria of the IAEA SSR-6. To ensure compliance with these requirements, specific mechanical and thermal tests need to be appointed with respect to the package type. Typically, IAEA SSR-6 prescribes mechanical tests followed by a thermal fire test as part of a cumulative test sequence. To approve the fire test, BAM uses a reference package representing the original package in its outer geometry for characterizing the actual fire and its impact on the package. This serves a dual purpose: it allows the precise adjustment of experimental parameters for the package design to be approved, and it provides input parameters for thermomechanical simulations, cf. With the help of this methodology for characterizing boundary conditions of the package, temperature evolutions within the reference package can be studied in a Finite-Element Analyses (FEA). This helps to compare the experiments with the numerical simulations on the case of the reference package, but also serves at the same time for preliminary simulations with the package design to be approved. The thermal test consists of a 30-minute fully engulfing 800°C pool fire or an equally severe fire, e.g. a propane gas fire. The fire reference test is performed prior to the regulatory fire test with the package design to be approved. The fire reference package in the described case is a closed sheet 316L steel 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 insulating material. T2 - ASME PVP 2024 CY - Bellevue, WA, USA DA - 28.07.2024 KW - Reference Test KW - Experimental KW - Numerical KW - Heat Flux PY - 2024 SN - 978-0-7918-8851-3 SP - 1 EP - 8 AN - OPUS4-60848 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias A1 - Komann, Steffen T1 - Fracture Mechanical Analyses of a Welding Seam of a Thick-Walled Transport Package N2 - Transport packages shall satisfy various safety criteria regarding mechanical, thermal and radiation phenomena. Typical requirements focusing mechanical aspects are usually drop tests in accordance with IAEA regulations [1]. The drop tests are usually carried out experimentally and, as an additional measure, finite element analyses (FEA) are performed. A specific part of the investigations presented is the evaluation of the welding seam connecting cask shell and cask bottom. Experimental results and FEA are presented and compared. The evaluation of the welding seam performed includes a variety of aspects. In addition to the experimental and analytical stresses determined, different standards are used to investigate a possible crack initiation. Several destructive and non-destructive tests are performed for quality assurance in the manufacturing process as well as for different input parameters. The necessary monitoring and non-destructive measurement methods to define the boundary conditions of the standards are introduced. Taking into account all required parameters, the welding seam is examined and evaluated using the failure assessment diagrams (FAD) of the respective standards. It can be shown under the given boundary conditions that considering the experimental data, the welding seam is in the context of crack initiation below the enveloping curve in the acceptable region. More critical drop tests to be conducted are proposed and need to be investigated in future work. T2 - Pressure Vessels & Piping Conference® 2022 CY - Las Vegas, NV, USA DA - 17.07.2022 KW - Drop test KW - Fracture initiation KW - Transport package PY - 2022 AN - OPUS4-55374 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias A1 - Monavari, Mehran A1 - Bertovic, Marija A1 - Wille, Frank T1 - AI-driven Documentation Analysis for Safety Assessment of Packages with Hazardous Goods N2 - Transport packages for radioactive material require, dependent on type und quantity of the radioactive material, a regulatory approval. Safety assessments shall be conducted in compliance with the IAEA regulations and documented in a comprehensive package design safety report to obtain approval from authority. This comprehensive safety report evaluates a broad range of requirements from the regulations, including physical specifications, control measures, and testing assessments. Additionally, it encompasses supporting documents such as specifications, plans in a variety of complex documents. Safety and manufacturing reports contain multiple interconnected sub-reports covering various topics. Changes, such as component modifications, material property updates, standards, or regulatory revisions, often impact multiple sections of the safety analysis reports, making even minor adjustments complex and time-consuming. Each transport package has unique constraints, making every safety report distinct, despite following the same regulatory framework. This project aims to enhance data analysis and processing through AI-based approaches [4]. Automated methods for analysing and interlinking documentation will improve efficiency, accuracy, and consistency while reducing human error in safety assessments. A key challenge is that pre-trained Large Language Models (LLMs) lack domain-specific data on packaging safety, potentially leading to inaccurate results. To mitigate this, the project applies Retrieval-Augmented Generation (RAG) in conjunction with LLMs. This approach integrates the strengths of pre-trained models with expert knowledge from databases and document repositories, ensuring accurate, well-founded, and transparent assessments. Beyond technical challenges, human factors must be considered early. New technologies often trigger resistance, if not introduced properly. Furthermore, long-term AI-reliance may lead to loss of expertise needed to solve complex problems. User-centred approach ensures effective implementation and lasting viability. This study evaluates the feasibility of using LLMs [2] and RAG [1, 3] for regulatory compliance in radioactive material transport. By analysing current documentation workflows, we assess how AI-driven tools can interpret complex safety reports and identify critical dependencies. Furthermore, we highlight the necessity of robust data governance, confidentiality measures, and AI reliability in this highly regulated domain. T2 - KONTEC 2025 CY - Dresden, Germany DA - 17.09.2025 KW - AI-driven KW - QI-Digital KW - Retrieval-Augmented Generation KW - Large Language Model PY - 2025 SP - 1 EP - 9 PB - Kontec Gesellschaft für technische Kommunikation GmbH CY - Dresden AN - OPUS4-64154 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias A1 - Neumann, Martin A1 - Linnemann, Konrad A1 - Komann, Steffen A1 - Wille, Frank T1 - Fracture Mechanical Analyses of a Welding Seam of a Thick-Walled Transport Package N2 - Transport packages shall satisfy various safety criteria regarding mechanical, thermal and radiation phenomena. Typical requirements focusing mechanical aspects are usually drop tests in accordance with IAEA regulations. The drop tests are usually carried out experimentally and, as an additional measure, finite element analyses (FEA) are performed. A specific part of the investigations presented is the evaluation of the welding seam connecting cask shell and cask bottom. Experimental results and FEA are presented and compared. The evaluation of the welding seam performed includes a variety of aspects. In addition to the experimental and analytical stresses determined, different standards are used to investigate a possible crack initiation. Several destructive and non-destructive tests are performed for quality assurance in the manufacturing process as well as for different input parameters. The necessary monitoring and non-destructive measurement methods to define the boundary conditions of the standards are introduced. Taking into account all required parameters, the welding seam is examined and evaluated using the failure assessment diagrams (FAD) of the respective standards. It can be shown under the given boundary conditions that considering the experimental data, the welding seam is in the context of crack initiation below the enveloping curve in the acceptable region. More critical drop tests to be conducted are proposed and need to be investigated in future work. T2 - Pressure Vessels & Piping Conference® 2022 CY - Las Vegas, NV, USA DA - 17.07.2022 KW - Transport package KW - Drop test KW - Fracture initiation PY - 2022 SP - 1 EP - 9 AN - OPUS4-55375 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias A1 - Neumann, Martin A1 - Linnemann, Konrad A1 - Wille, Frank T1 - Welding Seam Safety Evaluation in a Thick-Walled Type B Transport Package N2 - The safety demonstration of heavy weight type B transport packages used for storing spent nuclear fuel (SNF) or radioactive waste is ensured by a combination of physical testing and numerical calculations. While experiments are performed in accordance with the IAEA regulations for selected drop scenarios, Finite-Element-Method (FEM) simulations are used to predict the most damaging case and to investigate additional drop positions. BAM as competent authority in Germany has performed different investigations of a welding seam for a typical large transport package made of A508 forged steel, where the bottom plate is welded to the cylindrical shell. The package has a mass of approx. 120 t. Results of physical drop tests with a full-scale model and accompanying preliminary FEM simulations are presented to determine the decisive stresses in the welding seam. A drop test only represents one set of a package and test parameters. A further parameter analysis is considered to account for allowable variations of packaging properties (e.g. resulting from the manufacturing process) and, based on IAEA requirements, the temperature dependence of the material behaviour. The results of the stress analyses from the drop test and the simulation form the basis and provide the input parameters for a fracture mechanics analysis. In addition to the IAEA specifications, further standards are taken into account for an in-depth investigation, see R6 [1], BS 7910 [2] and API 579-1/ASME FFS1 [3]. All the above-mentioned standards require a manufacturer-specific defect analysis with respect to size and position. Both result from the welding process and the following heat treatment regime. The maximum defect sizes are ensured with non-destructive test methods (such as ultrasonic or particle methods) as integral part of the manufacturing process of the welding seam. Another important parameter in the welding process is the residual stress (secondary stress). The combination of the primary and secondary stress determines the total stress in the welding seam. The most damaging case of the welding seam is determined and evaluated with help of the abovementioned standards and taking into account the IAEA requirements with respect to defect sizes, material properties, primary and residual stress, yield strength etc. T2 - PATRAM 22 - The International Symposium on the Packaging and Transportation of Radioactive Materials CY - Juan-Les-Pins, Antibes DA - 11.06.2023 KW - Welding KW - Transport Package KW - Fracture Mechanics PY - 2023 SP - 1 EP - 11 AN - OPUS4-59421 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias A1 - Neumann, Martin A1 - Linnemann, Konrad A1 - Wille, Frank T1 - Fracture Mechanical Analyses of a Welding Seam of a Thick-Walled Transport Package N2 - The safety demonstration of heavy weight type B transport packages used for storing spent nuclear fuel (SNF) or radioactive waste is ensured by a combination of physical testing and numerical calculations. While experiments are performed in accordance with the IAEA regulations for selected drop scenarios, Finite-Element-Method (FEM) simulations are used to predict the most damaging case and to investigate additional drop positions. BAM as competent authority in Germany has performed different investigations of a welding seam for a typical large transport package made of A508 forged steel, where the bottom plate is welded to the cylindrical shell. The package has a mass of approx. 120 t. Results of physical drop tests with a full-scale model and accompanying preliminary FEM simulations are presented to determine the decisive stresses in the welding seam. A drop test only represents one set of a package and test parameters. A further parameter analysis is considered to account for allowable variations of packaging properties (e.g. resulting from the manufacturing process) and, based on IAEA requirements, the temperature dependence of the material behaviour. The results of the stress analyses from the drop test and the simulation form the basis and provide the input parameters for a fracture mechanics analysis. In addition to the IAEA specifications, further standards are taken into account for an in-depth investigation, see R6, BS 7910 and API 579-1/ASME FFS1. All the above-mentioned standards require a manufacturer-specific defect analysis with respect to size and position. Both result from the welding process and the following heat treatment regime. The maximum defect sizes are ensured with non-destructive test methods (such as ultrasonic or particle methods) as integral part of the manufacturing process of the welding seam. Another important parameter in the welding process is the residual stress (secondary stress). The combination of the primary and secondary stress determines the total stress in the welding seam. The most damaging case of the welding seam is determined and evaluated with help of the above-mentioned standards and taking into account the IAEA requirements with respect to defect sizes, material properties, primary and residual stress, yield strength etc. T2 - PATRAM 22 - The International Symposium on the Packaging and Transportation of Radioactive Materials CY - Juan-Les-Pins, Antibes, France DA - 11.06.2023 KW - Transport Package KW - Welding KW - Fracture Mechanics PY - 2023 SP - 1 EP - 10 AN - OPUS4-57696 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Guitton, Matthieu A1 - Chrupek, Thierry A1 - Leblanc, Vincent A1 - Van derTuijn, Jeroen A1 - Kaufhold, Dominik A1 - Kesting, Frederik A1 - Müller, Lars T1 - Multilateral approval of a SCO-III (three spent steam generator lower parts from EDF Fessenheim nuclear power plant) multimodal shipment N2 - A Surface Contaminated Object (SCO) is defined as a solid object which is not itself radioactive, but which has radioactive material distributed on its surfaces. The 2018 Edition of the IAEA Regulations for the Safe Transport of Radioactive Material (SSR-6 Rev. 1) introduced the group SCO-III to transport unpackaged large objects. The two pressurized water reactors of the Fessenheim Nuclear Power Plant, owned by EDF, were shut down in 2020. Previously, for maintenance operations in the year 2002 and 2010, the three steam generators – over 20 metres long, 300 tons – of each reactor were taken out from the reactors, divided in two and stored on site: - EDF shipped the six upper parts, as SCO-I, to Cyclife, in Studsvik (Sweden), in 2021, to be recycled; - EDF plans two transports of the lower parts (three parts for each transport), as SCO-III (considering the higher activity due to the NPP primary coolant) to the same facility. Modes of transport are, from consignor to consignee, road, inland waterways and maritime. The countries crossed by inland waterways, from Neuf-Brisach to Dunkerque (France), are France, Germany, Belgium and the Netherlands. According to the para. 825 of SSR-6 Rev. 1, “multilateral approval shall be required for the shipment of SCO-III”. EDF has submitted to all Competent Authorities an application (para. 827A of SSR-6 Rev. 1), including a transport plan and demonstrations to withstand normal conditions of transport. Despite SSR-6 Rev. 1, national regulations have their own specificities and each Competent Authority has its own assessment procedures. Nevertheless, involved Authorities agreed for a joint review, which included: - the understanding of national regulatory requirements, - the expected level of detail and quality of the safety case, - the feedback from previous shipments of similar type, - the contents, including the conditions, of the approvals to be issued. This joint review demonstrated good coordination between authorities, enabling the application to be processed rapidly, and the various authorities to benefit from each other's expertise and feedback. This was of particular importance, as it was the first application for the transport of an SCO-III assessed by the Competent Authorities involved. Finally, the authorities were able to issue approvals, whose format is certainly in line with their national provisions, but whose content and level of requirements in terms of transport safety are consistent. T2 - PATRAM 2025 CY - San Antonio, Texas, USA DA - 27.07.2025 KW - Shipment KW - Multilateral approval KW - Steam generator KW - SCO-III PY - 2025 VL - 2025 SP - 1 EP - 8 PB - Institute of Nuclear Materials Management (INMM) CY - Indianapolis AN - OPUS4-64026 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias T1 - An update on metal seal tests performed at BAM and implications for interim storage N2 - The investigation of the long-term performance of sealing systems employed in containers for radioactive waste and spent nuclear fuel is one research focus area for division 3.4 “Safety of Storage Containers” at the Bundesanstalt für Materialforschung und -prüfung. Metallic seals are in use in interim storage casks for used fuel and high active waste. Their change in performance over time is of high importance for maintaining safe enclosure. Therefore, investigations on these systems were started at BAM to get a general understanding of the relevant processes. Our investigations comprise investigations on different parameters which influence the seal performance and the main part is focussed on the time and temperature dependent behaviour. In this contribution an update on the performed tests and their respective results will be given in respect to the overview presented in 2020. The understanding of the long-term behaviour of the sealing systems and the performance evaluation during interim storage and subsequent transportation are our goals. T2 - SEDS-Workshop: Sicherheit der längerfristigen trockenen Zwischenlagerung CY - Garching, Germany DA - 15.05.2024 KW - Metal seals KW - Ageing KW - Leakage rate KW - Creep PY - 2024 AN - OPUS4-60640 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias T1 - Tests of metallic seals at Bam 3.4 performed over more than 15 years N2 - The investigation of the long-term performance of sealing systems employed in containers for radioactive waste and spent nuclear fuel is one research focus area for division 3.4 “Safety of Storage Containers” at the Bundesanstalt für Materialforschung und -prüfung (BAM) in Germany. The time dependant behaviour of metallic seals, which are employed in interim storage casks for used fuel and high active waste, is of high importance for ensuring safe enclosure. Therefore, investigations on these systems were started at BAM to get a general understanding of the relevant processes. Our investigations comprise investigations on different parameters which influence the seal performance and the main part is focussed on the time and temperature dependent behaviour. In this contribution an overview of the performed tests and their respective results will be given. T2 - 5th sealing workshop by Maestral CY - Avignon, France DA - 05.11.2024 KW - Metal seal KW - Aging KW - Interim storage PY - 2024 AN - OPUS4-62048 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias T1 - Investigations at BAM on Fuel Cladding Integrity and DPC Seal Performance 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 2022 CY - Charlotte, NC, USA DA - 07.11.2022 KW - Metal seal KW - Fuel cladding KW - Ring compression test KW - Brittle failure PY - 2022 AN - OPUS4-56400 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias T1 - Overview of metal seal tests performed at BAM and implications for extended interim storage N2 - The investigation of the long-term performance of sealing systems employed in containers for radioactive waste and spent nuclear fuel is one research focus area for division 3.4 “Safety of Storage Containers” at the Bundesanstalt für Materialforschung und -prüfung. The behaviour of metallic seals, which are employed in interim storage casks for used fuel and high active waste, over time is of high importance for ensuring safe enclosure. Therefore, investigations on these systems were started at BAM to get a general understanding of the relevant processes. Our investigations comprise investigations on different parameters which influence the seal performance and the main part is focussed on the time and temperature dependent behaviour. In this contribution an overview of the performed tests and their respective results will be given. A focus will lie on a comparison between seals with large torus diameter in comparison to small diameters. Our aim is to understand the long-term behaviour of the sealing systems for evaluation of their performance during possible extended interim storage and subsequent transportation. T2 - 4th Workshop on Safety of Extended Dry Storage of Spent Nuclear Fuel CY - Online meeting DA - 03.06.2020 KW - Seal performance KW - Metal seal KW - Leakage rate PY - 2020 AN - OPUS4-50843 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias T1 - Elastomer seals - overview and development N2 - The talk summarizes the current and future investigations of elastomer seal materials in division 3.4. T2 - IRSN-BAM Symposium Safety of Transport and Storage Packages CY - Online meeting DA - 19.11.2020 KW - Seal KW - Degradation KW - Ageing PY - 2020 AN - OPUS4-51751 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias T1 - Aging investigations of metal seals used in containers for interim storage of spent fuel N2 - In Germany dual purpose containers with double lid systems are used for the transport and storage of spent fuel and high level radioactive waste. The bolted lids are equipped with metal seals (e.g. Helicoflex®) to ensure the safe enclosure of the inventory for the required time period. The used metal seals have a layered structure consisting of three components. In the center a helical spring is positioned that is surrounded by two C-shaped jackets. The helical spring is mainly responsible for generation of the required restoring force. The inner jacket is made of stainless steel and homogenizes the restoring force generated by the helical spring. The outer jacket is made of silver or aluminum. Both metals are soft in comparison to the contact partners (lid and container body). During bolting of the lid the seal is compressed between lid and container body. The generated restoring force of the helical spring causes a plastic deformation of the outer jacket which adapts to the surfaces of the lid and the container body. Hence, leakage paths are closed and the sealing function is caused. Typical durations for existing interim storage licenses in Germany are 40 years, but due to the challenges in selection and construction of a final repository it can be expected that they have to be extended to longer periods. An extension of license periods requires knowledge of the long-term behavior of the seals. To meet this challenge long-term investigations have been started at Bundesanstalt für Materialforschung und –prüfung (BAM) in 2009. These tests focus on seals assembled in test flanges which are stored at temperatures ranging from room temperature to 150 °C for accelerated aging. The aged seals are tested repeatedly after certain aging steps and the performance is evaluated in respect to the leakage rate, the remaining seal force, and the useable resilience upon decompression. In this contribution an update on the running investigations in respect to earlier publications (Grelle et al. 2019, Goral et al. 2023) will be given and the implications of the results for resilient long-term safety will be presented. New activities will be addressed and their expected outcome be discussed. T2 - PVP2024, Pressure Vessels & Piping Conference CY - Bellevue, WA, USA DA - 29.07.2024 KW - Metal seal KW - Aging KW - Interim storage PY - 2024 AN - OPUS4-61355 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias T1 - Low temperature properties of rubber seals N2 - Understanding the failure mechanisms that lead to leakage of rubber seals at low temperatures is important. Rubber properties are highly temperature dependent. Changes in behaviour at low temperatures can result from the rubber-glass transition and for some rubbers by partial crystallisation. Experimental results on the low temperature behaviour of different rubber seals were discussed to help understand seal failure processes. T2 - Rubber in Engineering Group (RIEG) webinar an Elastomers at Low Temperatures CY - Online meeting DA - 11.09.2020 KW - Low temperature KW - Rubber seal KW - Compression set KW - Glass transition PY - 2020 UR - https://www.iom3.org/rubber-engineering-group/event/elastomers-low-temperatures AN - OPUS4-51306 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias T1 - Investigations of Aged Metal Seals for Interim Storage N2 - The storage of spent fuel and high level radioactive waste in Germany is performed in interim storage containers with double lid systems. The lids are bolted and equipped with metal seals (e.g. Helicoflex®) to ensure the safe enclosure of the inventory. The used metal seals have a layered structure consisting of three components as can be seen schematically in the cross-sectional view in Fig. 1. In the center a helical spring is positioned that is surrounded by two C-shaped jackets and is mainly responsible for generation of the required restoring force. The inner jacket is made of stainless steel and homogenizes the restoring force of the helical spring. The outer jacket is made of silver or aluminium which both are soft metals in comparison to the contact partners (lid and container body). During bolting of the lid to the container body the seal is compressed. The generated restoring force of the helical spring causes a plastic deformation of the outer jacket and adapts to the surfaces of the lid and the container body. Hence, leakage paths are closed and the sealing function is generated. Typical durations for existing interim storage licenses in Germany are 40 years, but it can be expected that they have to be extended to longer periods as a final repository will not be available before the end of the running licence periods. This extension of license periods requires a solid understanding of the long-term behaviour of the seals under storage conditions. To meet this challenge long-term investigations have been started at Bundesanstalt für Materialforschung und –prüfung (BAM) in 2009. These tests focus on seals assembled in test flanges which are stored at temperatures ranging from room temperature to 150 °C for accelerated ageing. The aged seals are tested repeatedly after certain ageing steps and the leakage rate as indicator for sealing performance, the remaining seal force, and the useable resilience upon decompression are determined. In the poster an update on the performed investigations in respect to earlier publications (Grelle et al. 2019, Goral et al. 2023) will be given and the implications of the results for resilient long term safety will be presented. Additionally, a focus will be laid on the currently planned further investigations and the question “What is additionally needed for evaluation of an interim storage period extension in regard to the used metal seals?” will be addressed. T2 - safeND2023: Forschungssymposium des BASE CY - Berlin, Germany DA - 13.09.2023 KW - Metal seal KW - Interim storage KW - Ageing PY - 2023 AN - OPUS4-58568 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias A1 - Goral, Milan A1 - Sagradov, Ilja A1 - Wolff, Dietmar A1 - Völzke, Holger T1 - Aging investigations of metal seals used in containers for interim storage of spent fuel N2 - For bolted casks for used fuel and high level waste the sealing system has an important impact of ensuring safe enclosure of the inventory The used metal seals show some changes over time which have been addressed for nearly 15 years in running experiments at BAM The observed changes are temperature de pendent and therefore it is expected that test s at higher temperature allow for an acceleration of the effects and ideally the proposal of an ageing model covering the expected time of use . Nevertheless, questions concerning the influence of the experimental approach on the results and effects taking place at the initial days of ageing remain. To address these a new project was launched to give a better understanding of the seal behavior. T2 - PVP2024, Pressure Vessels & Piping Conference CY - Bellevue, WA, USA DA - 29.07.2024 KW - Metal seal KW - Aging KW - Interim storage PY - 2024 SN - 978-0-7918-8851-3 SP - 1 EP - 4 PB - ASME AN - OPUS4-61357 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Jaunich, Matthias A1 - Goral, Milan A1 - Wolff, Dietmar A1 - Völzke, Holger ED - Krüssenberg, A. T1 - Overview of metal seal tests performed at BAM and implications for extended interim storage N2 - The investigation of the long-term performance of sealing systems employed in containers of radioactive waste and spent nuclear fuel is one research focus area for division 3.4 \Safety of Storage Containers" at the Bundesanstalt für Materialforschung und -prüfung. The behaviour of metallic seals, which are employed in interim storage casks for spent nuclear fuel and high active waste, over time is of high importance for ensuring safe enclosure. Therefore, investigations on these systems were started at BAM to get a general understanding of the relevant processes. Our investigations comprise investigations on different parameters which influence the seal Performance and the main part is focussed on the time and temperature dependent behaviour. Our aim is to understand the longterm behaviour of the sealing systems for evaluation of their performance during possible extended interim storage and subsequent transportation. In this contribution an overview of the performed tests and their respective results is given. A focus lies on a comparison between seals with large torus diameter in comparison to small diameters. T2 - Safety of Extended Dry Storage of Spent Nuclear Fuel SEDS2020: International online-workshop CY - Online meeting DA - 03.06.2020 KW - Metal seal KW - Seal performance KW - Long-term investigation KW - Ageing PY - 2020 DO - https://doi.org/10.3139/124.200071 SN - 0932-3902 VL - 85 IS - 6 SP - 440 EP - 443 PB - Carl Hanser Verlag CY - München AN - OPUS4-51833 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen T1 - Shipment of SCO-II – Authority assessment of mechanical aspects and quality management N2 - The decommissioning of nuclear facilities necessitates either the storage or disposal of large radioactive components such as steam generators, pressurizers, reactor pressure vessels and heads, or coolant pumps. These components or objects are large in size and mass with up to 6 meters in diameter and 20 meters in length and a weight of up to 400 tons. They are often transported to a storage, disposal, or recycling facility. Large components from nuclear facilities may often not be packed and need to be transported unpackaged due to size and weight. T2 - PATRAM 2025 CY - San Antonio, TX, USA DA - 27.07.2025 KW - Transport KW - Radioaktive Stoffe KW - Rückbau kerntechnischer Anlagen KW - Gefahrgut PY - 2025 AN - OPUS4-63877 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen 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 - CONF A1 - Komann, Steffen A1 - Ballheimer, Viktor A1 - Quercetti, Thomas A1 - Scheidemann, Robert A1 - Wille, Frank T1 - Mechanical and thermal assessment by BAM of a new package design for the transport of SNF from a german research reactor N2 - For disposal of the research reactor of the Technical University Munich FRM II a new transport and storage cask design was under approval assessment by the German authorities on the basis of International Atomic Energy Agency (IAEA) requirements. The cask body is made of ductile cast iron and closed by two bolted lid systems with metal seals. The material of the lids is stainless steel. On each end of the cask the wood-filled impact limiters are installed to reduce impact loads to the cask under drop test conditions. In the cavity of the cask a basket for five spent fuel elements is arranged. This design has been assessed by the Bundesanstalt für Materialforschung und -prüfung (BAM) in view to the mechanical and thermal safety analyses, the activity release approaches, and subjects of quality assurance and surveillance for manufacturing and operation of the package. For the mechanical safety analyses of the package a combination of experimental testing and analytical/numerical calculations were applied. In total, four drop tests were carried out at the BAM large drop test facility. Two tests were carried out as a full IAEA drop test sequence consisting of a 9m drop test onto an unyielding target and a 1m puncture bar drop test. The other two drop tests were performed as single 9m drop tests and completed by additional analyses for considering the effects of an IAEA drop test sequence. The main objectives of the drop tests were the investigation of the integrity of the package and its safety against release of radioactive material as well as the test of the fastening system of the impact limiters. Furthermore, the acceleration and strain signals measured during the tests were used for the verification of finite-element (FE) models applied in the safety analysis of the package design. T2 - Pressure Vessels & Piping Conference 2020 CY - Online meeting DA - 03.08.2020 KW - Mechanik KW - radioaktives Material KW - Transportbehälter KW - Antragsverfahren KW - Zulassungen KW - Typ-B Versandstück KW - Thermik PY - 2020 VL - 2020 SP - 1 EP - 7 PB - ASME CY - New York AN - OPUS4-51103 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Linnemann, Konrad A1 - Wille, Frank A1 - Reiche, Ingo A1 - Ramsay, J. T1 - New SCO-iii regulations to ship large comonents as surface contaminated objects N2 - The decommissioning or refurbishment of nuclear facilities necessitates either the storage or disposal of large radioactive components such as steam generators, pressurizers, reactor pressure vessels and heads, and coolant pumps, to list the major contributors. These components or objects are large in size and mass, measuring up to approximately 6 meters in diameter, up to 20 meters in length, and weighing over 400 000 kg. In many situations, the components are transported off-site to a storage, disposal or recycling/treatment facility. Previously, many large objects had to be transported under special arrangement. T2 - PATRAM22 CY - Juan les Pins, France DA - 11.06.2023 KW - Gefahrgut KW - Radioaktive Stoffe KW - Rückbau kerntechnischer Anlagen KW - Transport PY - 2023 SP - 1 EP - 8 AN - OPUS4-57750 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Müller, Lars A1 - Neumann, Martin A1 - Wille, Frank T1 - Shipment of SCO-II – Authority assessment of mechanical aspects and quality management N2 - The decommissioning of nuclear facilities necessitates either the storage or disposal of large radioactive components such as steam generators, pressurizers, reactor pressure vessels and heads, or coolant pumps. These components or objects are large in size and mass with up to 6 meters in diameter and 20 meters in length and a weight of up to 400 tons. They are often transported to a storage, disposal, or recycling facility. Large components from nuclear facilities may often not be packed and need to be transported unpackaged due to size and weight. T2 - PATRAM 2025 CY - San Antonio, TX, USA DA - 27.07.2025 KW - radioaktive Stoffe, Rückbau kerntechnischer Anlagen, Transport, Gefahrgut PY - 2025 SP - 1 EP - 10 AN - OPUS4-63878 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Wille, Frank A1 - Linnemann, Konrad A1 - Ramsay, Jeff A1 - Reiche, Ingo T1 - Transport of large nuclear power plant components – new SCO-III regulations and mechanical design assessment N2 - The decommissioning or refurbishment of nuclear facilities necessitates either the storage or disposal of large radioactive components such as steam generators, pressurizers, reactor pressure vessels and heads, and coolant pumps, to list the major contributors. These components or objects are large in size and mass, measuring up to approximately 6 meters in diameter, up to 20 meters in length, and weighing over 400 000 kg. In many situations, the components are transported off-site to a storage, disposal or recycling/treatment facility. Previously, many large objects had to be transported under special arrangement. The latest 2018 edition of the International Atomic Energy Agency (IAEA) Regulations for the Safe Transport of Radioactive Material, No. SSR-6, incorporates regulations for the shipment of contaminated, but not activated large objects as a new category of surface contaminated object, SCO-III, based on the IAEA “performance package” concept. This paper provides background and practical guidance on these regulations. Additionally, first experiences with SCO-III in the approval process are presented. For SCO-III components a mechanical assessment regarding the tests specified in the transport regulations SSR-6 for demonstrating ability to withstand normal conditions of transport is required - a free drop of the package onto an unyielding target and a penetration test have to be investigated. Large components are unique objects; therefore, it is not possible to choose experimental testing as assessment method. The application of a complex numerical analysis for mechanical proof is necessary. The assessment of the loads takes place based on local stress distributions. The large nuclear components have typically been transported in an unpackaged manner, so that the external shell of the component provides the containment wall. The paper will represent the methods, which are used for the evaluation of the mechanical integrity of the object for transport approval. T2 - PVP2024, Pressure Vessels & Piping Conference CY - Bellevue, Washington USA DA - 29.07.2024 KW - Transport KW - Radioaktive Stoffe KW - Rückbau kerntechnischer Anlagen KW - Gefahrgut PY - 2024 SN - 978-0-7918-8851-3 SP - 1 EP - 4 AN - OPUS4-60829 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kömmling, Anja T1 - Experiences and perspectives for the design evaluation of Konrad containers N2 - The former iron ore mine Konrad near Salzgitter is the approved final repository for radioactive waste with negligible heat production in Germany and should go into operation in the 2030s. The waste comes from operation and decommissioning of nuclear power plants, as well as from medicine, research and industry using radiation sources or radioactive materials. These types of waste are comparable to international low- and certain types of intermediate-level waste. The waste includes a large variety of products, materials and properties. The containers for such waste have the primary purpose of enabling handling and storing of the waste by safely enclosing the radioactive inventory and shielding of radiation. Due to the large variety of waste forms to be disposed of in the Konrad repository, different container types are specified to account for the different requirements related to each waste type. Each container design for waste disposal in Konrad has to be approved by BGE on the basis of the “final disposal conditions” (Endlagerungsbedingungen) and the respective “product control” (Produktkontrolle). For demonstrating the container’s compliance with the regulatory framework, the waste producers or container manufacturers have to apply for design approval on the basis of a comprehensive safety assessment including all relevant reports about analytical, numerical and experimental safety demonstrations and quality assurance measures for container manufacturing and operation. Usually, BGE commissions independent experts like BAM to evaluate these documents and the design testing (e.g. drop or fire tests). While the basic assessment principle is to verify the container design and quality to be in line with all regulatory requirements, there are often challenges associated with the container’s safety evaluation. As containers of different types have already been produced in large numbers and partially loaded with waste before the regulatory framework became effective (so-called “old” containers), they are considered in the regulations as well and safety assessments have to be provided in an equivalent way. As the requirements for the documentation of manufacturing quality were not yet known at the time these containers were manufactured, this is often challenging in practice. On the other hand, containers requiring the highest safety level (ABK II, sf) due to their nuclear content sometimes cause significant challenges and very long approval processes concerning safety demonstrations for severe accidental conditions like a 5 m drop without impact limiter onto a nearly unyielding target or a one hour 800 °C fire scenario. Furthermore, as the regulatory framework has not been updated for several decades, some requirements and the respective safety assessment methods have to be interpreted under consideration of the current state of knowledge in science and technology. T2 - safeND conference 2025 CY - Berlin, Germany DA - 17.09.2025 KW - Radioactive waste KW - Final disposal KW - Waste container KW - Design evaluation PY - 2025 AN - OPUS4-64989 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kömmling, Anja T1 - Analysis of O-ring seal failure in the context of radioactive waste containers 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. Besides examining the low-temperature behaviour and irradiation effects of elastomeric seals, it is our goal to estimate 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) 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. Further methods characterising seal performance are compression stress relaxation (CSR) reflecting the loss of sealing force of a compressed seal over time, and compression set (CS) which represents the recovery behaviour of a seal after release from compression. Additionally, material properties such as hardness, elastic modulus, glass transition temperature and viscoelastic loss factor as well as relaxation and recovery behaviour are examined in order to understand the underlying ageing mechanisms in each material. 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 aging caused by diffusion-limited oxidation (DLO) effects, resulting in distorted bulk properties such as compression stress relaxation and compression set. However, if DLO-affected data is excluded, extrapolations of CS data are possible using time-temperature shifts and Arrhenius graphs. 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 and even exhibited an improved, i.e. decreased leakage rate while other properties already indicated strong deterioration. These tests were extended by seal tests with fast partial decompression. Overall an update will be given on the current investigations and results and the planned activities. T2 - Kautschuk Herbst Kolloquium 2022 CY - Online meeting DA - 08.11.2022 KW - Seal KW - Ageing KW - Compression set PY - 2022 AN - OPUS4-56397 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kömmling, Anja T1 - Aging of elastomer O-rings and PE neutron shielding materials for radioactive waste containers N2 - Our institution BAM in Berlin, Germany is concerned with research and testing of materials in the context of safety in chemistry and technology. Our working group is involved in the licensing procedures of casks for radioactive waste. Besides, we’re doing research on aging and lifetime prediction of elastomer O-rings and investigate degradation and thermal expansion of PE neutron shielding materials. T2 - Polymers in nuclear applications CY - Online meeting DA - 01.12.2021 KW - Rubber KW - Polyethylene KW - Thermal expansion KW - Lifetime KW - Degradation PY - 2021 UR - https://energiforsk.se/media/30631/bam_material_aging_analysis.pdf AN - OPUS4-54191 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Kömmling, Anja A1 - Grelle, Tobias A1 - Jaunich, Matthias A1 - Goral, Milan A1 - Wolff, Dietmar T1 - Three-dimensional thermal expansion of neat and irradiated (U)HMWPE materials at elevated temperatures N2 - The thermal expansion of polymeric parts can be an issue in many applications where the available space is limited, or exact dimensions of the part are required. For this study, a device was designed and built that allowed measuring the thermal expansion simultaneously in all three spatial directions on cubic samples with real-scale dimensions (78 mm edge length). The results are shown between 25 °C and 125 °C for two PE materials, one HMWPE and one tempered UHMWPE, for non-irradiated samples as well as cubes that have been irradiated with 100 and 400 kGy. The results measured with the new device were very similar to those measured with conventional thermo-mechanical analysis equipment and to literature data of UHMWPE. The HMWPE material shows a much larger thermal expansion coefficient in one direction compared to the other two directions during the first heating due to frozen stresses from the pressing step during material manufacturing. These stresses are mostly released by the expansion during the first heating, so that the expansion during the second heating is more uniform. The overall volumetric expansion is the same for both heating runs. By contrast, the tempered UHMWPE material shows no significant difference between first and second heating run, as the stresses from processing could already relax in the tempering step. The irradiation treatment does not affect the values significantly for the given test set-up. KW - Lupolen KW - Ultra high molecular weight polyethylene KW - GUR KW - Coefficient of thermal expansion KW - High temperature PY - 2022 UR - https://nbn-resolving.org/urn:nbn:de:kobv:b43-563987 DO - https://doi.org/10.1016/j.polymertesting.2022.107841 SN - 0142-9418 VL - 117 SP - 1 EP - 8 PB - Elsevier Ltd. AN - OPUS4-56398 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Kömmling, Anja A1 - Jaunich, Matthias A1 - Goral, Milan A1 - Wolff, Dietmar T1 - Insights for lifetime predictions of O-ring seals from five-year long-term aging tests N2 - O-rings made of HNBR, EPDM and FKM were aged in the compressed and uncompressed state at 150 °C, 125 °C, 100 °C, 75 °C, 60 °C and 23 °C for aging times of up to five years. Hardness was measured and increased with aging time and temperature for HNBR and EPDM, but it remained practically constant for FKM. Indenter modulus measurements were performed on the lateral O-ring surface (that was free of DLO effects) to assess an influence of the compression during aging, but none was detected. The equilibrium compression set (CS) exhibited faster and stronger degradation than hardness and was used for lifetime predictions using the time-temperature superposition (TTS) principle. With an end-of-lifetime criterion of 70 % CS, lifetimes of 4.5 years, 50 years and 526 years at 75 °C were estimated for HNBR, EPDM and FKM, respectively. The activation energies derived from an Arrhenius plot of the shift factors from the TTS were 85 kJ/mol, 99 kJ/mol and 78 kJ/mol for HNBR, EPDM and FKM, respectively, revealing that a higher activation energy does not necessarily mean that the material has a higher lifetime at lower temperatures. Furthermore, the measured lifetime of EPDM O-rings at 100 °C (5 years) was compared to that predicted on the basis of the lifetime at 150 °C as well as 125 °C using the corresponding shift factors. The error of the prediction was only ± 4 %. However, this precise prediction could only be achieved using the five-year long-term aging data. When using only data from aging times up to 0.5 years and 2 years, the lifetime of EPDM O-rings at 100 °C was underestimated by 31 % and 22 %, respectively. KW - HNBR KW - EPDM KW - FKM KW - DLO KW - Hardness KW - Compression set KW - Rubber KW - Elastomer KW - Degradation PY - 2020 DO - https://doi.org/10.1016/j.polymdegradstab.2020.109278 VL - 179 SP - 109278 PB - Elsevier Ltd. AN - OPUS4-51060 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Linnemann, Konrad 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 - Linnemann, Konrad 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 - BAM-IRSN-Workshop 2021 CY - Berlin, Germany DA - 08.09.2021 KW - Transport packages KW - Bolt connections KW - Finite element analysis KW - Experimental investigation PY - 2021 AN - OPUS4-53251 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Linnemann, Konrad T1 - Experimental and numerical investigation of 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 packages are assessed to specific transport conditions which are specified in the IAEA safety standards SSR-6 (IAEA 2018). These transport conditions, especially the so-called accident conditions of transport, imply high dynamic loading on the lids and the bolt connections of the package. The response of the lid systems on the mechanical accident conditions is generally investigated by experimental drop tests or numerically, e.g., by finite element analyses. The interpretation of the drop test results for the verification of the numerical models is often not obvious due to the complex superposition of different effects in the real tests. BAM has started a research project to get a better understanding about the behavior of prestressed bolt connections under loadings typical for these drop tests. In this context an experimental test set-up was developed to investigate the response of a single bolt connection under a prescribed lateral displacement of clamped parts. The bolt is instrumented by strain gauges to get the pretensional, the torsional and the bending stress in the bolt shank. Furthermore, the lateral movement and the tilt of the bolt head is measured during the test. A finite element model of the test set-up has been created in Abaqus FEA (Simulia 2021). The very detailed instrumentation of the test set-up shall give the opportunity to investigate and validate the numerical model. The aim of this paper is to give an overview about the proposed research project and to present first results. T2 - SMIRT26 - 26th International Conference on Structural Mechanics in Reactor Technology CY - Potsdam, Germany DA - 10.07.2022 KW - Transport packages KW - Bolt connections KW - Finite element analysis KW - Experimental investigation PY - 2022 AN - OPUS4-55426 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Linnemann, Konrad A1 - Komann, Steffen A1 - Wille, Frank A1 - Reiche, I. A1 - Ramsay, J. T1 - New Sco-III regulations to ship large objects as surface contaminated objects N2 - The decommissioning or refurbishment of nuclear facilities necessitates either the storage or disposal of large radioactive components such as steam generators, pressurizers, reactor pressure vessels and heads, and coolant pumps, to list the major contributors. These components or objects are large in size and mass, measuring up to approximately 6 meters in diameter, up to 20 meters in length, and weighing over 400 000 kg. In many situations, the components are transported off-site to a storage, disposal or recycling/treatment facility. Previously, many large objects had to be transported under special arrangement. The latest 2018 edition of the International Atomic Energy Agency (IAEA) Regulations for the Safe Transport of Radioactive Material, No. SSR-6 [1], incorporates regulations for the shipment of large objects as a new category of surface contaminated object, SCO-III, based on the IAEA “performance package” concept. This paper provides background and practical guidance on these regulations. Additionally, the experiences of BAM with the appoval of two steam converters of the NPP Lingen are presented as the first approval process for SCO-III objects in Germany. The primary additions to SSR-6 include SCO-III classification and requirements, approval and administrative requirements for the new classification, and the addition of SCO-III to the proper shipping name for UN 2913. Advisory material drafted for the new requirements will be included in the next revision of SSG-26, Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material, expected to be published soon. Note that at this time the proposed provisions for large objects do not include components such as reactor vessels, due to the more limited experience and greater radioactivity levels. The SCO-III concept lays the groundwork and may be extended to cover other large objects that are classified as low specific activity (LSA) material in the future. T2 - Kerntechnik 2022 CY - Leipzig, Germany DA - 21.06.2022 KW - Transport packages KW - Radioactive materials KW - SCO-III KW - Large objects PY - 2022 SP - 1 EP - 7 AN - OPUS4-55423 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 - Experimental and numerical investigation of 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 packages are assessed to specific transport conditions which are specified in the IAEA safety standards SSR-6 (IAEA 2018). These transport conditions, especially the so-called accident conditions of transport, imply high dynamic loading on the lids and the bolt connections of the package. The response of the lid systems on the mechanical accident conditions is generally investigated by experimental drop tests or numerically, e.g., by finite element analyses. The interpretation of the drop test results for the verification of the numerical models is often not obvious due to the complex superposition of different effects in the real tests. BAM has started a research project to get a better understanding about the behavior of prestressed bolt connections under loadings typical for these drop tests. In this context an experimental test set-up was developed to investigate the response of a single bolt connection under a prescribed lateral displacement of clamped parts. The bolt is instrumented by strain gauges to get the pretensional, the torsional and the bending stress in the bolt shank. Furthermore, the lateral movement and the tilt of the bolt head is measured during the test. A finite element model of the test set-up has been created in Abaqus FEA (Simulia 2021). The very detailed instrumentation of the test set-up shall give the opportunity to investigate and validate the numerical model. The aim of this paper is to give an overview about the proposed research project and to present first results. T2 - SMIRT 26 (26th conference on Structural Mechanics in Reactor Technology) CY - Potsdam, Germany DA - 10.07.2022 KW - Bolt connections KW - Finite element anaylsis KW - Experimental testing KW - Transport package for radioactive materials PY - 2022 UR - https://www.lib.ncsu.edu/resolver/1840.20/40614 SP - 1 EP - 7 PB - IASMiRT AN - OPUS4-59378 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Liu, Y. T1 - Cooperation in Advanced Technologies Associated with the Back End of Nuclear Fuel Cycle N2 - A Memorandum of Understanding (MOU) between Argonne National Laboratory (Argonne) and the German Bundesanstalt für Materialforschung und -prüfung (BAM) was signed in October 2014. Its objectives are to promote cooperation among scientists and specialists at Argonne and BAM and establish a framework for collaboration in advanced technologies associated with the back end of the nuclear fuel cycle. Collaborative activities involving Argonne and BAM may be implemented through the promotion of joint research activities and scientific workshops and conferences; exchange of technical information; and visits by scientists, specialists, and graduate, postgraduate, and Ph.D. students. The MOU was renewed in January 2020 for another five years. Highlights of the progress and achievements are provided in identified topical areas for collaboration on ageing management guidance, storage, transportation and disposal R&D, joint conference activities, and conclude with the benefits of the MOU in fostering Argonne and BAM collaboration. T2 - 26th International Conference on Structural Mechanics in Reactor Technology - SMiRT 26 CY - Potsdam, Germany DA - 10.07.2022 KW - Nuclear fuel cycle KW - Back end KW - Spent fuel performance KW - Storage KW - Transportation PY - 2022 AN - OPUS4-55862 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Liu, Yung A1 - Völzke, Holger A1 - Zencker, Uwe A1 - Jaunich, Matthias T1 - Cooperation in Advanced Technologies Associated with the Back End of Nuclear Fuel Cycle N2 - A Memorandum of Understanding (MOU) between Argonne National Laboratory (Argonne) and the German Bundesanstalt für Materialforschung und -prüfung (BAM) was signed in October 2014. Its objectives are to promote cooperation among scientists and specialists at Argonne and BAM and establish a framework for collaboration in advanced technologies associated with the back end of the nuclear fuel cycle. Collaborative activities involving Argonne and BAM may be implemented through the promotion of joint research activities and scientific workshops and conferences; exchange of technical information; and visits by scientists, specialists, and graduate, postgraduate, and Ph.D. students. The MOU was renewed in January 2020 for another five years. Highlights of the progress and achievements are provided in identified topical areas for collaboration on ageing management guidance, storage, transportation and disposal R&D, joint conference activities, and conclude with the benefits of the MOU in fostering Argonne and BAM collaboration. T2 - 26th International Conference on Structural Mechanics in Reactor Technology - SMiRT 26 CY - Potsdam, Germany DA - 10.07.2022 KW - Nuclear fuel cycle KW - Back end KW - Spent fuel performance KW - Storage KW - Transportation PY - 2022 SP - 1 EP - 10 AN - OPUS4-55861 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Musolff, Andre T1 - Fire Influence to Wood Filled Impact Limiters - Implications for the Package Design Safety Case N2 - Impact limiters with wooden components are widely used in the design of packages for transportation of radioactive material. In most designs, the wood is encapsulated with steel sheets. Impact limiters mainly determine the mechanical and thermal behaviour of the package in accident conditions of transport in accordance with the IAEA Regulations. In context with research and development for package design approval competence, the thermal behaviour of heavy-weight packages was investigated at BAM with an artificially pre-damaged generic impact limiter design. Within this first investigation, the pre-damaged impact limiter with a diameter of 2.3 meters was mounted on a water tank simulating the thermal capacity of a cask during the fire test. The water tank is part of a water circulating system built of several components such as pump, heater, cooler, sliding valve, flow meter, thermocouples and control unit in order to measure the heat flux. Furthermore, the investigations focus on the effects this additional heat generation would have on the cask and especially on the lid-closure system with the gasket. The results of these experiments could find consideration in the safety case of the transport packages of radioactive material using wood filled impact limiters. T2 - 20th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Juan-les-Pins, France DA - 11.06.2023 KW - Fire test KW - Impact limiter KW - Mechanical and thermal behaviour PY - 2023 AN - OPUS4-57719 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Müller, Lars 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 - Müller, Lars T1 - Thightness Assessment of welded Lids for Encapsulations of damaged Spent Nuclear Fuel in the Design Approval Process of Dual-Purpose Casks (DPC) N2 - The disposal of spent nuclear fuel in Germany is ensured using dual-purpose casks (DPC) for transport and interim storage. The leak tightness of the DPC and resulting containment is one of the most important aspects. Additional encapsulations are required for damaged spent nuclear fuel (DSNF) to guarantee safe handling and a separate tight closure. Due to the general design of DPCs for standard fuel assemblies should special requirements be considered for the design of the encapsulations for DSNF to ensure the loading in existing package designs. The absence of a replaceable sealing in the tightness barrier is the main difference for the encapsulations for damaged spent nuclear fuel. Instead, they are welded shut with a lid. The leak tightness of the encapsulation shall be proven in the design approval process for all transport conditions. This is especially valid for accident conditions of transport, where high internal impact forces may occur. BAM as German competent authority is responsible for the safety assessment of mechanical and thermal design, retention of radioactive material and quality assurance aspects of manufacturing and operation. BAM carried out a comprehensive safety assessment concerning the mechanical package design. As there are no representative standards for verifying the leak tightness of a welded lid, two approaches were being pursued. Established German standards may be used for verifying the leak tightness of a weld - but limited to low stresses. Therefore, physical tests were required for higher impact loads. Representative drop tests and highly sensitive leakage tests were performed. The paper presents an overview of the containment assessment by BAM and points out the main findings for the design of welds regarding leak tightness. Both verification procedures are described, on one hand with German standards for lower loads and on the other hand with physical tests for higher impact loads. The leak tightness of the encapsulation weld could be approved based on leakage tests and a corresponding evaluation for all transport conditions. T2 - PATRAM 2025 CY - San Antonio, TX, USA DA - 27.07.2025 KW - Tightness assessment KW - Welded lids KW - Transport packages KW - Encapsulations PY - 2025 AN - OPUS4-64017 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Müller, Lars A1 - Scheidemann, Robert A1 - Neumeyer, Tino A1 - Komann, Steffen A1 - Wille, Frank T1 - Tightness Verification Procedures of Welded Lids for Encapsulations for Damaged Spent Nuclear Fuel in the Design Approval Process of Dual-Purpose Cask (DPC) N2 - Damaged spent nuclear fuel (DSNF) can be loaded in Germany in dual-purpose casks (DPC) for transport and interim storage. Therefore, special encapsulations are needed to guarantee a safe handling and a reliable tight closure. These encapsulations must be durable and leak-tight for a possible long interim storage term and the transport thereafter. Because components inside the package containment like these special encapsulations are not accessible within periodical inspections of the DPC. Due to the general design of DPCs for standard fuel assemblies, specific requirements needed to be considered for the design of the special encapsulations for DSNF to ensure the loading in existing package designs. Especially the content of the encapsulations with damaged spent nuclear fuel must be avoided to enter in the inner space of the DPC, due to the possible higher amount of corrosive and radioactive material as well as greater quantities of water, which can be trapped in defect fuel rods. The main difference of the encapsulations for damaged spent nuclear fuel is that they have no replaceable sealing as tightness barrier. They have a specific welded lid. Therefor the tightness needs to be shown in the design approval process for all transport conditions especially for accident conditions, where high internal impact loads can occur. As German competent authority, BAM is responsible for the safety assessment of the mechanical and thermal design, the release of radioactive material and the quality assurance aspects of manufacturing and operation. Concerning the mechanical package design BAM carried out a comprehensive safety assessment for the specific encapsulations during the package design approval process of the commonly used DPC. For the proof of tightness of the welded lid no representative standards are available. For minor loads the German KTA standards can verify the tightness of a weld seam but only for low stresses. For higher impact loads physical tests are required. These tests were performed with representative drop tests and highly sensitive leak tests. The present paper gives an overview of the applied assessments by BAM and points out the main findings of the mechanical design for the welded sealing. Both verification procedures are described, on one hand with German standards for lower loads and on the other hand with physical tests for higher impact loads. The drop tests based on complex numerical pre-calculations by the applicant which result in optimized dimensions and properties of the test specimens and representative drop heights for the drop tests. The horizontal drop orientation with a plane-parallel impact on the unyielding basement was most challenging and led to specific adaptations during preparation and conduction of the drop tests performed by BAM. With leak tests and the corresponding test evaluation, it could be finally approved that the specific welded sealing of an encapsulations for DSNF is tight for all transport conditions. T2 - ASME PVP 2024 CY - Bellevue, WA, USA DA - 28.07.2024 KW - Tightness verification KW - Weld seams KW - Drop tests KW - Encapsulations for damaged spent nuclear fuel KW - Helium leak tests PY - 2024 SN - 978-0-7918-8851-3 SP - 1 EP - 10 CY - New York AN - OPUS4-61351 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Müller, Lars A1 - Scheidemann, Robert A1 - Neumeyer, Tino A1 - Neumann, Martin A1 - Komann, Steffen A1 - Wille, Frank T1 - Thightness Assessment of welded Lids for Encapsulations of damaged Spent Nuclear Fuel in the Design Approval Process of Dual-Purpose Casks (DPC) N2 - The disposal of spent nuclear fuel in Germany is ensured using dual-purpose casks (DPC) for transport and interim storage. The leak tightness of the DPC and resulting containment is one of the most important aspects. Additional encapsulations are required for damaged spent nuclear fuel (DSNF) to guarantee safe handling and a separate tight closure. Due to the general design of DPCs for standard fuel assemblies should special requirements be considered for the design of the encapsulations for DSNF to ensure the loading in existing package designs. The absence of a replaceable sealing in the tightness barrier is the main difference for the encapsulations for damaged spent nuclear fuel. Instead, they are welded shut with a lid. The leak tightness of the encapsulation shall be proven in the design approval process for all transport conditions. This is especially valid for accident conditions of transport, where high internal impact forces may occur. BAM as German competent authority is responsible for the safety assessment of mechanical and thermal design, retention of radioactive material and quality assurance aspects of manufacturing and operation. BAM carried out a comprehensive safety assessment concerning the mechanical package design. As there are no representative standards for verifying the leak tightness of a welded lid, two approaches were being pursued. Established German standards may be used for verifying the leak tightness of a weld - but limited to low stresses. Therefore, physical tests were required for higher impact loads. Representative drop tests and highly sensitive leakage tests were performed. The paper presents an overview of the containment assessment by BAM and points out the main findings for the design of welds regarding leak tightness. Both verification procedures are described, on one hand with German standards for lower loads and on the other hand with physical tests for higher impact loads. The leak tightness of the encapsulation weld could be approved based on leakage tests and a corresponding evaluation for all transport conditions. T2 - PATRAM 2025 CY - San Antonio, Texas, USA DA - 27.07.2025 KW - Transport packages KW - Tightness assessment KW - Welded lids KW - Encapsulations PY - 2025 VL - 2025 SP - 1 EP - 11 PB - Institute of Nuclear Materials Management (INMM) CY - Indianapolis AN - OPUS4-64023 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 - JOUR A1 - Nagel, Thomas A1 - Rühaak, Wolfram A1 - Amann, Florian A1 - Bracke, Guido A1 - Buske, Stefan A1 - Kowalski, Julia A1 - Reiche, Sönke A1 - Schäfer, Thorsten A1 - Scheytt, Traugott A1 - Stumpf, Thorsten A1 - Völzke, Holger A1 - Wellmann, Florian A1 - Kolditz, Olaf T1 - Deep geological disposal - Editorial to the topical collection N2 - The subsurface geological environment, historically a key source of fossil and mineral resources, has seen a significant broadening of its potential applications in recent years, with increasing emphasis on its use for energy production and storage. A critical factor in utilizing subsurface space is ensuring the sustainability of management practices, with particular attention to the potential environmental impacts that may result. The current Topical Collection on Deep Geological Disposal places greater emphasis on the geotechnical aspects of geological disposal, with a particular focus on the safe isolation of hazardous materials, including radioactive waste. The Topical Collection on Deep Geological Disposal covers both experimental (laboratory and in-situ) and theoretical studies of various geotechnical applications, such as geological disposal of radioactive or other materials, e.g. coal-based solid waste, carbon dioxide sequestration in fractured reservoirs, and brine solutions, but also aspects of uranium mining. The articles in the Topical Collections are arranged under the following headings: – Fundamental studies – Experimental studies – Numerical studies KW - Disposal KW - Radioactive Waste KW - Geology PY - 2025 UR - https://nbn-resolving.org/urn:nbn:de:kobv:b43-624955 DO - https://doi.org/10.1007/s12665-024-12075-2 SN - 1866-6299 VL - 84 IS - 3 SP - 1 EP - 6 PB - Springer Nature AN - OPUS4-62495 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Nagelschmidt, Sven T1 - Application study of time-temperature correlation methods on test data of metal seals N2 - Reliable experimental data and prediction models are required to evaluate the behavior of components and materials over long periods of time. One method is based on the time-temperature equivalence principle, according to which a state or material property occurs or can be achieved equivalently at different time-temperature combinations. Higher temperatures generally shorten the duration and vice versa. Based on this approach, the long-term behavior of metal seals, as used in storage containers for spent nuclear fuels and high-level radioactive waste, has been studied at BAM and other institutions around the world for several years. However, such tests are often not planned as a preliminary point regarding the basic conditions of time-temperature related models. Gaps are then sometimes identified during the evaluation, for example because certain temperature values are missing, or the measurement intervals were not selected appropriately. This study evaluates exemplarily the status of the test data of metal seals at the BAM regarding the applicability of existing time-temperature correlation methods and corresponding models. Based on the existing database, the aim is to show which method and model are more likely to be applicable and which additional tests might be necessary to obtain a more suitable database. The tests so far have provided important findings for the planning and adaptation of possible further tests and for suitable model developments. Identified gaps in the data basis led to further development of the methods. In the present work, a time-temperature correlation was determined based on a time-temperature parameter from GRAHAM and WALLES for a selected data range. T2 - Pressure Vessels & Piping Conference PVP2025 CY - Montreal, Canada DA - 20.07.2025 KW - Time-temperature equivalence KW - Metal seals, KW - Time-temperature superposition KW - Time-temperature parameters PY - 2025 VL - 2025 SP - 1 EP - 4 PB - ASME CY - USA AN - OPUS4-63789 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Nagelschmidt, Sven A1 - Herbrich, Uwe A1 - Keller, Christian A1 - Qiao, Linan T1 - Re-evaluation of tensile and creep rupture data of metals using a modified Larson–Miller approach N2 - More than 70 years ago, in 1952 Larson and Miller adapted an existing relationship whichevaluates the relative effects of time and temperature on creep rupture behavior of various alloys by using existing data and by constructing so-called master curves. Since that time, the formulated Larson–Miller relation, commonly known as Larson–Miller parameter, is a widely used time–temperature parameter for various applications, e.g. for rupture life, creep and relaxation analyses of metals and some other materials such as polymers, concrete and ceramics.Nevertheless, the physical meaning of this parameter as well as the determination of master curves are still the subject of major criticism. In this work, both aspects are reviewed and analyzed with data originally considered in thepaper of Larson and Miller based on the following approach: (a) the parameter was modifiedregarding a reference time and a reference temperature according to the time–temperature superposition principle; (b) master curves were generated for five materials investigated originally by Larson and Miller, based on a stretched exponential function type. It has been shown, that: (1) With normalized time and temperature, the Larson–Miller parameter corresponds to real logarithmic rupture time. (2) For the relationship between Creep rupture stress and the modified Larson–Miller parameter, the stretched exponential function is a good choice as a master function. (3) Corresponding model parameters have physical meaning and can be easily determined based on data from creep test data. The investigations provide a better understanding and applicability of Larson–Miller approach. KW - Larson–Miller parameter KW - Creep rupture time KW - Time–temperature equivalence KW - Stretched exponential function KW - Extrapolation of short-term data PY - 2025 UR - https://nbn-resolving.org/urn:nbn:de:kobv:b43-644486 DO - https://doi.org/10.1016/j.engfracmech.2025.111582 SN - 0013-7944 VL - 329 SP - 1 EP - 16 PB - Elsevier Ltd. AN - OPUS4-64448 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Naster, Maximilian T1 - Development of a Hydrogen Fire Test Rig for Thermal Testing N2 - A new hydrogen-based test rig is being developed for a feasibility study on the use of hydrogen as an energy source for thermal testing in accordance with the IAEA SSR-6 guidelines for transport packages containing radioactive materials. The test rig will be capable of combusting hydrogen for a wide range of different burner geometries, mass flows and if necessary hydrogen blends. As this type of fire test according to the IAEA boundary conditions (SSR-6 & SSG-26) does not yet exist, a large number of preliminary investigations, safety assessments and calculations must be carried out in order to develop a viable concept for hydrogen fires. T2 - 1st JF BMUV Förderprogramm CY - Berlin, Germany DA - 22.05.2024 KW - IAEA Fire KW - H2 Fire KW - IAEA Regulations PY - 2024 AN - OPUS4-60555 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Naster, Maximilian T1 - Development of a Hydrogen Fire Test Rig for Thermal Testing N2 - A new hydrogen-based test rig is being developed for a feasibility study on the use of hydrogen as an energy source for thermal testing in accordance with the IAEA SSR-6 guidelines for transport packages containing radioactive materials. The test rig will be capable of combusting hydrogen for a wide range of different burner geometries, mass flows and if necessary hydrogen blends. As this type of fire test according to the IAEA boundary conditions (SSR-6 & SSG-26) does not yet exist, a large number of preliminary investigations, safety assessments and calculations must be carried out in order to develop a viable concept for hydrogen fires. T2 - Technical Exchange IRSN – BAM Transport & Storage of Packages for Radioactive Material CY - Berlin, Germany DA - 04.06.2024 KW - IAEA Fire KW - H2 Fire KW - IAEA Regulations PY - 2024 AN - OPUS4-60554 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Naster, Maximilian T1 - Experimental and Numerical Analyses for the Evaluation of Hydrogen as an Energy Source for Thermal Testing of Transport Packages of Radioactive Material N2 - In this paper we present a new hydrogen-based test rig for an ongoing feasibility study of using hydrogen as an energy source for the thermal testing of transport packages containing radioactive materials. The test rig will be capable of combusting hydrogen for a wide range of different burner geometries, mass flows and if necessary hydrogen blends. As this type of fire test according to the IAEA boundary conditions does not yet exist, a large number of preliminary investigations, safety assessments and calculations must be carried out in order to develop a viable concept for hydrogen fires. In the first step of the feasibility study, the temperature, structure, and radiation of various hydrogen flames are surveyed. In future works, the results will make it possible to design burner frames that are suitable for fire reference tests in order to make comparisons with pool and propane fires used in assessment procedures today. In parallel comparative numerical simulations are conducted to model the thermal behaviour of hydrogen flames using the software package Ansys®. On the one hand, the numerical simulations support the experiments by providing an overview of numerous parameters and the measuring range; on the other hand, they will help with the design of the burner frame in future work. This paper gives an overview in the design and capabilities of the test rig. Furthermore, the results of the parameter studies show that burner geometry and mass flow provide a significant design margin for the thermal shape of the hydrogen flames. In addition, the results of the initial numerical calculations will be used to determine the necessary sensors, the positions, and their operating range. Only the optimal interaction allows a controlled system that permits user-defined hydrogen fires. T2 - PVP2024, Pressure Vessels & Piping Conference CY - Bellevue, Washington, USA DA - 29.07.2024 KW - Hydrogen KW - Fire KW - IAEA Regulations KW - CFD KW - Burner PY - 2024 AN - OPUS4-60855 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Naster, Maximilian T1 - Hydrogen Fire Testing N2 - In this presentation we present a new hydrogen-based test rig for an ongoing feasibility study of using hydrogen as an energy source for the thermal testing of transport packages containing radioactive materials. The test rig will be capable of combusting hydrogen for a wide range of different burner geometries, mass flows and if necessary hydrogen blends. As this type of fire test according to the IAEA boundary conditions does not yet exist, a large number of preliminary investigations, safety assessments and calculations must be carried out in order to develop a viable concept for hydrogen fires. In the first step of the feasibility study, the temperature, structure, and radiation of various hydrogen flames are surveyed. In future works, the results will make it possible to design burner frames that are suitable for fire reference tests in order to make comparisons with pool and propane fires used in assessment procedures today. In parallel comparative numerical simulations are conducted to model the thermal behaviour of hydrogen flames using the software package Ansys®. On the one hand, the numerical simulations support the experiments by providing an overview of numerous parameters and the measuring range; on the other hand, they will help with the design of the burner frame in future work. This paper gives an overview in the design and capabilities of the test rig. Furthermore, the results of the parameter studies show that burner geometry and mass flow provide a significant design margin for the thermal shape of the hydrogen flames. In addition, the results of the initial numerical calculations will be used to determine the necessary sensors, the positions, and their operating range. Only the optimal interaction allows a controlled system that permits user-defined hydrogen fires. T2 - Technical Exchange IRSN – BAM Transport & Storage of Packages for Radioactive Material CY - Berlin, Germany DA - 04.06.2024 KW - Hydrogen PY - 2024 AN - OPUS4-60341 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Naster, Maximilian T1 - Experimental and numerical analyses of hydrogen flames for the thermal testing of transport packages for radioactive material N2 - In this paper we present an update of the hydrogen-based test rig for an ongoing feasibility study of using hydrogen as an energy source for the thermal testing of transport packages containing radioactive materials. The test rig is capable of combusting hydrogen for a wide range of different burner geometries, mass flows, hydrogen blends and single jet flame operations as well as a full array of burners for thermal testing can be set up. As this type of fire test according to the IAEA boundary conditions does not yet exist, a large number of preliminary investigations, safety assessments and simulations must be carried out in order to develop a viable concept for hydrogen fires. In a first step of the feasibility study, the temperature, structure, and radiative behavior of hydrogen jet flames must be surveyed. The simulation with a single hydrogen flame was investigated in a previous work. In the next step the results are used to study the interaction and structural behavior of multiple jet flames in proximity with varying nozzle distances. With the test rig completed, it will be possible in future works to design burner frames suitable for fire reference tests to make comparisons with pool and propane fires used in assessment procedure today. Thus, preliminary comparative numerical simulations are conducted to model the behavior of overlapping hydrogen jet flames using the software package Ansys®. This paper gives an overview on the current state and design of the test rig. Furthermore, the results of the simulations show that nozzle geometry, mass flow and nozzle distance provide significant design margin for designing a test fire capable of fully engulfing a specimen. T2 - PVP2025, Pressure Vessels & Piping Conference CY - Montreal, Quebec, Canada DA - 20.07.2025 KW - Computational Fluid Dynamics (CFD) KW - Hydrogen KW - IAEA fire test PY - 2025 AN - OPUS4-63865 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Naster, Maximilian A1 - Gleim, Tobias A1 - Wille, Frank T1 - Experimental and Numerical Analyses for the Evaluation of Hydrogen as an Energy Source for Thermal Testing of Transport Packages of Radioactive Material N2 - In this paper we present a new hydrogen-based test rig for an ongoing feasibility study of using hydrogen as an energy source for the thermal testing of transport packages containing radioactive materials. The test rig will be capable of combusting hydrogen for a wide range of different burner geometries, mass flows and if necessary hydrogen blends. As this type of fire test according to the IAEA boundary conditions does not yet exist, a large number of preliminary investigations, safety assessments and calculations must be carried out in order to develop a viable concept for hydrogen fires. In the first step of the feasibility study, the temperature, structure, and radiation of various hydrogen flames are surveyed. In future works, the results will make it possible to design burner frames that are suitable for fire reference tests in order to make comparisons with pool and propane fires used in assessment procedures today. In parallel comparative numerical simulations are conducted to model the thermal behaviour of hydrogen flames using the software package Ansys®. On the one hand, the numerical simulations support the experiments by providing an overview of numerous parameters and the measuring range; on the other hand, they will help with the design of the burner frame in future work. This paper gives an overview in the design and capabilities of the test rig. Furthermore, the results of the parameter studies show that burner geometry and mass flow provide a significant design margin for the thermal shape of the hydrogen flames. In addition, the results of the initial numerical calculations will be used to determine the necessary sensors, the positions, and their operating range. Only the optimal interaction allows a controlled system that permits user-defined hydrogen fires. T2 - PVP2024, Pressure Vessels & Piping Conference CY - Bellevue, Washington USA DA - 29.07.2024 KW - Hydrogen KW - Fire KW - Computational Fluid Dynamics (CFD) KW - IAEA fire test KW - Burner PY - 2024 SN - 978-0-7918-8851-3 SP - 1 EP - 9 AN - OPUS4-60852 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Naster, Maximilian A1 - Gleim, Tobias A1 - Wille, Frank T1 - Experimental and numerical analyses of hydrogen flames for the thermal testing of transport packages for radioactive material N2 - In this paper we present an update of the hydrogen based test rig for an ongoing feasibility study of using hydrogen as an energy source for the thermal testing of transport packages containing radioactive materials [ The test rig is capable of combusting hydrogen for a wide range of different burner geometries, mass flows , hydrogen blends and single jet flame operation s as well as a full array of burners for thermal testing can be set up. As this type of fire test according to the IAEA boundary conditions does not yet exist, a large number of preliminary investigations, safety assessments and simulations must be carried out in order to develop a viable concept for hydrogen fires. In a first step of the feasibility study, the temperature , structure, and radiative behavior of hydrogen jet flames must be surveyed. The simulation with a single hydrogen flame was investigated in a previous work. In the next step the results are used to study the interaction and structural behavior of multiple jet flames in proximity with varying nozzle distances. With the test rig completed, it will be possible in future works to design burner frames suitable for fire reference tests to make comparisons with pool and propane fires used in assessment procedure today. Thus, preliminary comparative numerical simulations are conducted to model the behavior of overlapping hydrogen jet flames using the software package Ansys®. This paper gives an overview on the current state and design of the test rig. Furthermore, the results of the simulations show that nozzle geometry, mas s flow and nozzle distance provide significant design margin for designing a test fire capable of fully engulfing a specimen. T2 - PVP2025, Pressure Vessels & Piping Conference CY - Montreal, Quebec, Canada DA - 20.07.2025 KW - Computational Fluid Dynamics (CFD) KW - Fire testing KW - Hydrogen KW - IAEA PY - 2025 SP - 1 EP - 10 AN - OPUS4-63864 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Nehrig, Marko T1 - Investigation of the internal impact during a 9 m drop test of an accident-safe waste package N2 - The safety assessment of packages for the transport of radioactive material follows the IAEA regulations and guidance. The specified regulatory tests cover severe accidents and demonstrate the package containment system integrity. Special attention must be drawn to the behaviour of the content which could move inside the package due to unpreventable gaps caused by the loading procedure and the structure of the content. A possible internal impact of the content which occurs during the drop tests onto the lid system is investigated. The IAEA regulations SSR-6 and the Guidance SSG-26, revised recently, consider input from Germany and France related to the significance of internal gaps. In the context of a waste package design assessment, a model was equipped with a representative content to conduct a drop test with an internal impact. The weight and kinetic impact of this content covered all possible real contents. The objective of the test was to maximize the load onto the lid system and to prove the mechanical integrity by complying with the required leak tightness. The test was conducted conservatively at a package temperature lower than -40 °C at the BAM Test Site Technical Safety. This paper gives an overview of efforts to address internal gaps and their consequences, and the BAM efforts with the implementation of this topic into IAEA regulations and guidance material. The paper then focuses on the conduction of a drop test and investigation of internal component impact. T2 - ASME 2021 Pressure Vessels & Piping Conference (PVP2021) CY - Online meeting DA - 13.07.2021 KW - Drop test KW - Internal gaps KW - IAEA PY - 2021 AN - OPUS4-54744 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Nehrig, Marko A1 - Scheidemann, Robert A1 - Wille, Frank A1 - Ballheimer, Viktor T1 - Investigation of the internal impact during a 9 m drop test of an accident-safe waste package N2 - The safety assessment of packages for the transport of radioactive material follows the IAEA regulations and guidance. The specified regulatory tests cover severe accidents and demonstrate the package containment system integrity. Special attention must be drawn to the behaviour of the content which could move inside the package due to unpreventable gaps caused by the loading procedure and the structure of the content. A possible internal impact of the content which occurs during the drop tests onto the lid system is investigated. The IAEA regulations SSR-6 and the Guidance SSG-26, revised recently, consider input from Germany and France related to the significance of internal gaps. In the context of a waste package design assessment, a model was equipped with a representative content to conduct a drop test with an internal impact. The weight and kinetic impact of this content covered all possible real contents. The objective of the test was to maximize the load onto the lid system and to prove the mechanical integrity by complying with the required leak tightness. The test was conducted conservatively at a package temperature lower than -40 °C at the BAM Test Site Technical Safety. This paper gives an overview of efforts to address internal gaps and their consequences, and the BAM efforts with the implementation of this topic into IAEA regulations and guidance material. The paper then focuses on the conduction of a drop test and investigation of internal component impact. T2 - ASME 2021 Pressure Vessels & Piping Conference (PVP2021) CY - Online meeting DA - 13.07.2021 KW - Internal gaps KW - Drop test KW - IAEA PY - 2021 SN - 978-0-7918-8535-2 DO - https://doi.org/10.1115/PVP2021-60996 SP - 1 EP - 6 PB - American Society of Mechanical Engineers (ASME) CY - New York, NY, USA AN - OPUS4-54742 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Neumann, Martin T1 - Nuclear disposal - transport and storage of radioactive waste N2 - This presentation deals with transport, interim storage, and final disposal of radioactive material in Germany in the context of the University of Magdeburg course "Introduction to nuclear plant design and safety". The IAEA regulatory framework, relevant tests for ensuring safety, and transport packages are a major part of the transport aspect. Different interim storage designs are discussed as well as the current state of final repositories in Germany analyzed. Finally, a European perspective is given. T2 - Vorlesungsreihe "Introduction to nuclear plant design and safety" CY - Magdeburg, Germany DA - 13.06.2025 KW - Nuclear KW - Transport KW - Radioactive KW - Material PY - 2025 AN - OPUS4-63433 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Neumann, Martin T1 - Experiences with the implementation of ageing management for packages for transport of radioactive materials in Germany N2 - The consideration of ageing mechanisms is now obligatory for the design of transport packages with integration of the para 613A into IAEA SSR-6 (Rev. 1). In addition, para 809(f) of SSR-6 (Rev. 1) requires 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. Para 503(e) requires that these packages have been maintained during storage in a manner that all requirements specified in SSR-6 (Rev.1) and in the applicable certificates of approval have been fulfilled. The evaluation of ageing mechanisms and their effects including monitoring are part of BAM’s authority assessment tasks related to the mechanical and thermal package design and quality assurance aspects. BAM has compiled the guideline BAM-GGR 023 for the implementation of ageing assessment and related measures into the approval procedure. The implementation of ageing management measures is obligatory in case of extension/renewal of package design approval certificates. BAM has evaluated package designs which are used only for transport as well as package designs for long term interim storage. The assessment of ageing mechanisms associated with the identification of ageing effects on components is the main part of the ageing management plan (AMP). Different approaches regarding AMP structure are introduced. Experiences and approaches about the evaluation of components for the expected package operating time are shown. We are focusing the evaluations of proofs for not accessible and not replaceable components. Operational experiences for these package designs are available and should be considered in the ageing evaluation. Corresponding measures for package monitoring are to be derived based on these results. The measures for monitoring shall be fixed in the Ageing Surveillance Program (ASP) to maintain a specification conform package for the transport on public routes. We show exemplary how results from ageing evaluation during the approval procedure are transferred into the ASP. T2 - PATRAM 2025 CY - San Antonio, TX USA DA - 27.07.2025 KW - RAM KW - Ageing management KW - Transport KW - Radioactive material PY - 2025 AN - OPUS4-63985 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Neumann, Martin T1 - Conclusions from the first return campaign of vitrified High-Level Waste on sea going vessels from the competent authority N2 - Germany has to take back vitrified waste from the reprocessing plants in France and Great Britain. The waste resulted from decades of transporting spent fuel for reprocessing to La Hague and Sellafield. The return campaign from France was concluded in 2024 with the transport of five CASTOR® HAW28M casks to the interim storage facility in Philippsburg in southern Germany. The return of waste from Sellafield comprises three campaigns and a total of 20 CASTOR® HAW28M casks. The first campaign was performed in 2020, consisting of six CASTOR® HAW28M casks, while the second campaign was performed in March 2025, consisting of seven CASTOR® HAW28M casks. The casks were transported by rail from Sellafield to the port in Barrow-in-Furness, where they were loaded into a dedicated seagoing vessel, certified as INF Class 3 according to the INF Code. This was the first time that vitrified high level waste with considerable heat load was transported under a German design approval certificate. The third campaign is expected to be performed during 2026. BAM as part of the German competent authority system was among others involved in the assessment of the sea transport. BAM required for the first transport, among others, an assessment of temperature distribution during transport, logging of temperatures of cargo bays and graphical imaging of temperatures of the bay with the cask to ensure compliance with temperature specifications, e.g. maximal neutron absorber and gasket temperatures. Special interest was taken in the identification of possible events exceeding the specified temperatures considering the different philosophies of the IMDG code and its supplement the INF code regarding temperature control of hatches. Results show compliance with assumed conditions. T2 - PATRAM 2025 CY - San Antonio, TX, USA DA - 27.07.2025 KW - RAM KW - Transport KW - Radioactive material KW - Sea PY - 2025 AN - OPUS4-63983 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Neumann, Martin T1 - Design assessment, approval of management systems and ageing aspects of transport packages for radioactive material not requiring competent authority approval of design N2 - Most transports of radioactive materials are carried out with packages not requiring competent authority approval of design. These encompass – in accordance with the IAEA SSR-6 [1] regulations – packages of the classification excepted, Industrial packages Type 1, 2 and 3 and Type A packages. Currently an upsurge in number and variation of these package designs can be seen in Germany, resulting from the phase out of nuclear energy in Germany as well as e. g. increased use of radioactive material for medical purposes. A design assessment regarding the package safety is required in the international IAEA SSR-6 [1] regulations. BAM operates facilities for the performance of all regulatory tests required such as drop towers for a wide range of package masses and dimensions, fire test, leak tightness measurements and pressure test facilities. Experiences with several package types are shown. Additionally, IAEA SSR-6 [1] requires the establishment of a management system for design, manufacture, maintenance, and repair of the packaging as well as for the preparation, consigning, loading, carriage, unloading and receipt of the package. Relevant for Germany, BAM has published guidance material on the process of management system acceptance in the technical guide BAM-GGR 016 [2]. The requirements encompass quality management plans for the manufacturing of packages including independent manufacturing surveillance and specific instructions for operation, maintenance, and repair of packagings. Examples for management system specifics and requirements are given. Since the latest edition of the IAEA SSR-6 [1] regulations an ageing evaluation including systematic ageing management measures are required for all kind of package types. BAM is going to update the guidance material BAM-GGR 016 [2] to support the stakeholders with relevant information to fulfil the ageing aspect for packages not requiring competent authority approval. The paper explains how the ageing aspect may be included in the safety evaluation process and the management system measures and will give an outlook for the future guidance material. T2 - 20th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 22) CY - Juan-les-Pins, France DA - 11.06.2023 KW - Radioactive material KW - Type A package KW - Management system KW - Tests KW - Non-approved PY - 2023 AN - OPUS4-57704 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - GOVDOC A1 - Neumann, Martin T1 - Measures for quality and ageing management for non-competent authority approved package designs for the transport of radioactive material N2 - The BAM-GGR 016 explains the requirements of ADR, RID, ADN and IMDG for management systems for design, manufacturing, testing, documentation, use, maintenance, and inspection of packagings for package designs for the transport of radioactive material not requiring competent authority approval for all designs of exempted packages, packages type IP-1, IP-2, IP-3 and type A. KW - Not requiring competent authority approval KW - Radioactive material KW - Transport PY - 2024 UR - https://nbn-resolving.org/urn:nbn:de:kobv:b43-636506 VL - BAM-GGR 016 SP - 1 EP - 25 PB - Bundesanstalt für Materialforschung und -prüfung (BAM) CY - Berlin ET - Revision 1 AN - OPUS4-63650 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Neumann, Martin T1 - Friction coefficients for wood-wood and wood-steel contacts N2 - The friction at the interfaces between pieces of wood and steel is of crucial importance for the impact and energy absorption in impact limiters of transport casks for radioactive material. Here, the friction coefficient for the combinations wood-wood and wood-steel was measured in the temperature range between -40°C and 90°C. Results show decreasing friction with increasing temperature, ranging from 0.43 at -40°C to 0.22 for 90°C for wood-steel combinations and from 0.3 at -40°C to 0.24 at 90°C for a wood-wood combination. T2 - Sandia-BAM Workshop CY - Berlin, Germany DA - 21.11.2022 KW - Wood KW - Friction KW - Transport cask PY - 2022 AN - OPUS4-56342 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Neumann, Martin T1 - Friction coefficients for wood-wood and wood-steel interfaces in impact limiters for transport casks N2 - Wood is widely used in impact limiters of transport casks for radioactive material. Encapsulated by an outer and inner steel structure, spruce wood is often applied in layers of alternating direction. The friction at the interfaces between these layers is of crucial importance for the impact and energy absorption e.g., at an accidental impact of a cask against a hard target. In order to get detailed information for corresponding numerical calculations, in this study the friction coefficient for the combinations wood-wood and wood-steel was measured in the temperature range between -40°C and 90°C according to the relevant stress conditions for such casks. Results show decreasing friction with increasing temperature, ranging from 0.43 at -40°C to 0.22 for 90°C for wood-steel combinations and from 0.3 at -40°C to 0.24 at 90°C to for a wood-wood combination. T2 - 20th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 22) CY - Juan-les-Pins, France DA - 11.06.2023 KW - Wood KW - Friction KW - Transport cask PY - 2023 AN - OPUS4-57702 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -