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 - 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 - Völzke, Holger T1 - Update on the German Spent Fuel Management Strategy N2 - The presentation provides an update on the German spent fuel Management strategy in the areas of transportation, dry interim storage and deep geological disposal. T2 - IAEA 18th TWG-NFCO CY - Online meeting DA - 07.09.2020 KW - Spent nuclear fuel KW - Final disposal KW - Safety KW - Transport and storage cask PY - 2020 AN - OPUS4-51616 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - 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 - 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 - 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 - 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 - CHAP A1 - Droste, Bernhard A1 - Wolff, Dietmar A1 - Bevilacqua, A. A1 - Reiche, I. A1 - Harvey, J. A1 - Hirose, M. A1 - Kumano, Y. A1 - McConnell, P. A1 - Saegusa, T. A1 - Einziger, R. ED - González-Espartero, A. T1 - Methodology for a Safety Case of a Dual Purpose Cask for Storage and Transport of Spent Fuel N2 - Spent nuclear fuel which is generated in the operation of nuclear reactors needs to be safely managed following its removal from the reactor core. On-site power reactor storage pools were designed on the assumption that after a short period of time spent nuclear fuel would be removed for reprocessing and disposal or further storage elsewhere. The amount of highly radioactive spent fuel that needs to be stored over longer periods of time is growing and additional storage capacity is required. One of the widely used options for additional storage capacity is the use of casks for dry storage of spent fuel. Among various existing dry storage concepts, several Member States are utilizing a concept of dual purpose casks (DPCs). This publication provides practical advice on the structure and contents of a DPC integrated safety case with reference to existing IAEA requirements relevant to the licensing and use of transport and storage casks for spent fuel. KW - Dual Purpose Cask KW - Spent Fuel KW - Storage KW - Transport PY - 2020 UR - https://www.iaea.org/publications/10966/methodology-for-a-safety-case-of-a-dual-purpose-cask-for-storage-and-transport-of-spent-fuel SN - 978-92-0-131620-2 SN - 1011–4289 SP - 1 EP - 109 PB - IAEA Publishing Section CY - Vienna, Austria AN - OPUS4-51962 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Ruiz-Hervias, J. A1 - Simbruner, Kai A1 - Cristobal-Beneyto, M. A1 - Perez-Gallego, D. A1 - Zencker, Uwe T1 - Failure mechanisms in unirradiated ZIRLO® cladding with radial hydrides N2 - The purpose of this paper is to investigate the relationship between hydride morphology, in particular the presence of radial hydrides (RHs), stress state and failure mechanisms associated with the ring compression test (RCT). Samples of ZIRLO®cladding were pre-hydrided and subjected to thermo-mechanical treatments to precipitate long radial hydrides. The results show that the reorientation treatment was very successful. A considerable fraction of RHs was generated, the radial hydride continuity factor being around 80 to 90% of the wall thickness. The samples with reoriented hydrides were tested using the RCT at room temperature. Macroscopic brittle failure was observed with sudden load drops for displacements around 0.5 mm, with a calculated “offset strain”between 0.5 and 1%. Crack nucleation occurs in RHs located in regions with the highest values of hoop stress. These locations are the inner diameter of cladding at the vertical plane of the sample (12 and 6 o’clock positions) and the outer diameter at the horizontal plane (3 and 9 o’clock positions). Noticeable load drops in the RCT are associated with unstable crack propagation events through the radial hydride network, the crack front reaching up to 90% of the wall thickness in some cases. The failure micro-mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth and coalescence in the Zr matrix, with ductile tearing patches connecting neighboring hydrides. The main conclusion is that radial hydride metrics is not the only parameter that determines cladding failure in the presence of RHs, but the interaction between the location and continuity of RHs and the stress normal to the hydride (the hoop stress in this case). Consequently, if a radial hydride is located at a position within the cladding where the hoop stress is small, a crack will not be initiated easily in the RCT. KW - ZIRLO® KW - Cladding Embrittlement KW - Radial Hydrides KW - Failure Mechanism KW - Ring compression test PY - 2021 DO - https://doi.org/10.1016/j.jnucmat.2020.152668 SN - 0022-3115 VL - 544 SP - 152668 PB - Elsevier B.V. AN - OPUS4-52000 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Simbruner, Kai T1 - Brittle failure analysis of spent fuel zirconium alloy claddings during long-term dry interim storage - Current approach at BAM N2 - Understanding the mechanical behaviour of fuel cladding is decisive for an extended dry interim storage period beyond the currently licensed period of 40 years in Germany and subsequent handling, transport, and disposal considering normal and accidental conditions. Since fuel cladding acts as the first barrier for spent fuel, its integrity should be demonstrated to enable safe and secure discharge and handling options after interim storage and subsequent transportation in preparation of final disposal. An established method for characterizing cladding material is the Ring Compression Test (RCT), where a small, cylindrical sample of the cladding tube is subjected to a compressive load. This test is a laboratory representation of a fuel rod load case and has shown a high susceptibility to failure under the occurrence of radial hydrides. Certain hydride morphologies and low temperature might even lead to brittle fracture at very small loads. The BAM research project BRUZL (Fracture mechanical analysis of spent fuel claddings during long-term dry interim storage) aims to study the results of quasi-static RCTs and to establish numerical models to gain an in-depth understanding of the stress state during the experiments. Using a fracture mechanical approach, cases of sudden failure during the RCT procedure are used to characterize the material behaviour and establish a failure criterion. As an integral part of the project BRUZL, quasi-static Ring Compression Tests have been performed to identify all experimental details, which might be helpful for numerical modelling. Unirradiated samples of the cladding material ZIRLO® have been subjected to hydrogen charging and a thermo-mechanical treatment for radial hydride reorientation. Sample preparation, testing procedures, and analysis results are presented. A numerical model has been established and an elastic-plastic material model was derived from as-received RCT samples by inverse finite element analyses. Cohesive zone modelling has been implemented to reproduce sudden load drops during RCT. The project is funded by the Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561. T2 - Kolloquium der Abteilung 3 CY - Online meeting DA - 05.11.2020 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2020 AN - OPUS4-51987 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Simbruner, Kai T1 - Embrittlement of Spent Fuel Claddings ‒ Results of Ring Compression Tests N2 - An established method for characterizing cladding material is the Ring Compression Test (RCT), where a small, cylindrical sample of the cladding tube is subjected to a compressive load. This test is a laboratory representation of a fuel rod load case and has shown a high susceptibility to failure under the occurrence of radial hydrides. Certain hydride morphologies and low temperature might even lead to brittle fracture at very small loads. The BAM research project BRUZL (Fracture mechanical analysis of spent fuel claddings during long-term dry interim storage) aims to study the results of quasi-static RCTs and to establish numerical models to gain an in-depth understanding of the stress state during such experiments. Using a fracture mechanical approach, cases of sudden failure during the RCT procedure are used to characterize the material behaviour and establish a failure criterion. As an integral part of the project BRUZL, quasi-static Ring Compression Tests have been performed to identify all experimental details, which might be helpful for numerical modelling. Unirradiated samples of the cladding material ZIRLO® have been subjected to hydrogen charging and a thermo-mechanical treatment for radial hydride reorientation. Sample preparation, testing procedures, and analysis results are presented. A numerical model has been established and an elastic-plastic material model was derived from as-received RCT samples by inverse finite element analyses. Cohesive zone modelling has been implemented to reproduce sudden load drops during RCT. The project is funded by the Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561. T2 - 4th GRS Workshop on Safety of Extended Dry Storage of Spent Nuclear Fuel CY - Online meeting DA - 03.06.2020 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2020 AN - OPUS4-51988 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - EURAD WP 8 (SFC) - Task 3: Behaviour of nuclear fuel and cladding after discharge N2 - The European Joint Programme on Radioactive Waste Management (EURAD) deals in work package (WP) 8 with Spent Fuel Characterization (SFC). Inspired by the EURAD activities, an international Coordinated Research Project (CRP) on SFC was established by the International Atomic Energy Agency (IAEA). The EURAD WP SFC participants are collaborating as a team on the IAEA CRP on SFC. The EURAD WP SFC project is divided into four tasks. Task 3 investigates the behaviour of nuclear fuel and cladding after discharge. The aim of these activities is to understand and describe the evolution of the cladding-pellet system and its ageing under conditions of extended interim storage, transportation and emplacement in a final disposal system. During an IAEA consultancy meeting, BAM as leader of Task 3 reported on the current status of the research work. T2 - IAEA Consultancy Meeting on the Coordinated Research Project on Spent Fuel Characterization CY - Online meeting DA - 28.06.2021 KW - Nuclear Fuel KW - Cladding KW - Spent Fuel Characterization KW - Extended Interim Storage KW - Final Disposal PY - 2021 AN - OPUS4-52909 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Simbruner, Kai T1 - Embrittlement of Spent Fuel Claddings – Fracture Mechanical Evaluation and Failure Criteria N2 - The BAM research project BRUZL (Fracture mechanical analysis of spent fuel claddings during long-term dry interim storage) evaluates results of quasi-static Ring Compression Tests (RCT). Due to its simple realisation, the RCT is an established method for characterising the mechanical behaviour of fuel cladding, where a small tube sample is subjected to a compressive load and the relation between load and displacement is obtained. During tests at room temperature, unirradiated ZIRLO® samples with a distinct extend of radial hydrides have shown high susceptibility to sudden, brittle failure at relatively small loads. Hydride morphology and radial hydride metrics have been an integral part of the characterisation of degradation mechanisms of cladding. A software tool for the analysation of metallographic micrographs was developed. The program allows the automated calculation of hydride lengths, detection of continuous hydride structures, their statistical evaluation and visualisation. Numerical models have been established to reproduce experimental RCT results featuring an elastic-plastic material model, which was derived from inverse finite element analyses of as-received RCT samples. A cohesive zone model (CZM) is introduced to the model to reproduce load drops in the linear regime of the load vs. displacement curve. Using a traction-separation approach for the CZM, the cohesive parameters are fitted to match numerical and experimental results yielding the fracture mechanical material properties. Combining the three key aspects hydride morphology, fracture mechanics and RCT stress-state a failure criterion can be established as a result of the BRUZL project. The project is funded by the Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561. T2 - 5th GRS Workshop on Safety of Extended Dry Storage of Spent Nuclear Fuel CY - Online meeting DA - 09.06.2021 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2021 AN - OPUS4-53779 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - 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 - 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 - Simbruner, Kai T1 - Embrittlement of Spent Fuel Claddings - Micromechanically Motivated Cohesive Zone Modelling N2 - Since cladding acts as the first barrier for nuclear fuel, it is key to understand the degrading mechanisms leading to cladding embrittlement. Zirconium alloy based cladding samples subjected to simulated pre-storage drying conditions have shown high susceptibility to brittle failure under the occurrence of hoop stress and low temperature as expected after long-term interim storage. The stress state associated with a compressive pinch-load scenario is reproduced by the Ring Compression Test (RCT), which results in sudden failure at relatively small mechanical loads with high sensitivity to the presence of radial hydrides. As part of the BAM research project BRUZL (Fracture mechanics analysis of spent fuel claddings during long-term dry interim storage), static RCTs were conducted on non-irradiated, pre-hydrided ZIRLO® ring-shaped samples that were subjected to a radial hydride treatment. All samples failed in macroscopically brittle fashion with sudden load drops and severe crack propagation through almost the entire wall thickness. In a post-RCT examination of the fracture surfaces using scanning electron microscopy, radial hydrides could be identified as the weakest link of the structure. However, the microscopic failure mechanism is much more complex due to cracks initiating at different axial positions in depth direction of the ring, overlapping cracks, and arresting cracks. A considerable portion of the fracture surfaces appeared to have undergone plastic deformation of the zirconium matrix resulting in void growth and coalescence associated with ductile failure. The geometry of the ductile patches was measured, and the ductile surface fraction was determined. Based on these observations, a statistical model was developed that can reproduce the composition of the fracture surface as a one-dimensional projection. A micromechanically motivated cohesive zone model (CZM) was implemented in existing RCT finite element analysis (FEA) models with different constitutive laws for brittle hydrides and ductile matrix respectively. In this work, FEA results are presented for non-irradiated ZIRLO® and high-burnup M5® claddings. The impact of varying hydride-matrix combinations is discussed. Acknowledgement: The project was funded by the former Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561. T2 - 6th GRS Workshop on the Safety of Extended Dry Storage of Used Nuclear Fuel CY - Garching, Germany DA - 01.06.2022 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2022 AN - OPUS4-55023 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - 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 - 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 - Simbruner, Kai T1 - Brittle Failure Analysis of High-Burnup PWR Fuel Cladding Alloys N2 - The general aim of this research is the development of methods for predicting mechanical behavior and identification of limiting conditions to prevent brittle failure of high-burnup (HBU) pressure water reactor (PWR) fuel cladding alloys. A finite element (FE) model of the ring compression test (RCT) was created to analyze the failure behavior of zirconium-based alloys with radial hydrides during the RCT. An elastic-plastic material model describes the zirconium alloy. The stress-strain curve needed for the elastic-plastic material model was derived by inverse finite element analyses. Cohesive zone modeling is used to reproduce sudden load drops during RCT loading. Based on the failure mechanism in non-irradiated ZIRLO® claddings, a micro-mechanical model was developed that distinguishes between brittle failure along hydrides and ductile failure of the zirconium matrix. Two different cohesive laws representing these types of failure are present in the same cohesive interface. The key differences between these constitutive laws are the cohesive strength, the stress at which damage initiates, and the cohesive energy, which is the damage energy dissipated by the cohesive zone. Statistically generated matrix-hydride distributions were mapped onto the cohesive elements and simulations with focus on the first load drop were performed. Computational results are in good agreement with the RCT results conducted on high-burnup M5® samples. It could be shown that crack initiation and propagation strongly depend on the specific configuration of hydrides and matrix material in the fracture area. T2 - 26th International Conference on Structural Mechanics in Reactor Technology - SMiRT 26 CY - Potsdam, Germany DA - 10.07.2022 KW - Cladding Embrittlement KW - Ring Compression Test KW - Cohesive Zone Modelling KW - Spent Nuclear Fuel PY - 2022 AN - OPUS4-55435 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - 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 - 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 - Wille, Frank T1 - Test facilities for transport and storage packages at BAM N2 - BAM acts as authority and for service in safety assessment of packages for transport and storage of radioactive materials. We offer extensive test capabilities and application of analytical methods for design verification and simulation for all types of packages for the transport and storage of radioactive materials according with the international IAEA Regulations for the safe transport and for national storage acceptance criteria. BAM operates several test facilities for drop and stacking testing, leak testing and thermal testing. The large drop test tower allows dropping full-scale specimens up to 200,000 kg in any drop orientation as requested. The comprehensive test facilities combined with long-term experience, newest equipment and measurement devices according to the latest state-of-the-art technology ensures realisation of complex test campaigns for package safety evaluation. Beyond that, non-destructive and destructive material test devices and experts are available. Equipment and application of all kinds of typical measurement categories can be offered for testing campaigns. In recent years we performed testing of full-scale type B package models with complex handling and preparation procedures. The results were contributed for different package design approval procedures. Type A packages mainly designed for medical related transport purposes, were continuously tested according to the transport regulations over recent years as well. Moreover, we work on research topics with relevance to package safety. The mechanical behaviour of lid closure systems under transport and storage conditions and the thermal behaviour of impact limiters were recently of special importance for the assessment competencies of BAM and were investigated under use of our test facilities. The paper describes the test facilities and capabilities for package design safety evaluation at BAM and shows examples from our recent work. T2 - PATRAM 2022, 20th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Juan-les-Pins, France DA - 11.06.2023 KW - Transport KW - Drop testing KW - Fire test KW - IAEA KW - Package PY - 2023 AN - OPUS4-58088 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 - Rolle, Annette A1 - Komann, Steffen A1 - Wille, Frank T1 - Ageing aspect in design approval of special form radioactive material N2 - In accordance with the IAEA transport regulations, the design of special form radioactive material (SFRM) shall resist a severe transport accident without undue loss or dispersal of radioactive material. The safety assessment for design approval includes besides the program for physical tests (impact, percussion, bending and heat test) also the evaluation of the management system for design, manufacture, testing, documentation, use, maintenance, and inspection. SFRM source design plus management system shall ensure, that every specimen of the approved design is able to survive the severe mechanical and thermal tests at any time of its SFRM-working life. Due to the long-term use of SFRM designs in most cases, the assessment of the source ageing behavior is an important aspect in the approval procedure. Different fields of application imply a wide range of environmental conditions, from clean room atmosphere to highly aggressive industrial conditions. Besides of radioactive content, corrosion is a main factor for possible SFRM design degradation. Although the IAEA Advisory Material SSG-26 already implies an indication of the need for considering ageing mechanisms, suitable amendments in the regulatory requirements of SSR-6 should be introduced to make the approval procedure more transparent and help to reduce rounds of questions by the authority. A supplementary requirement for considering of ageing mechanisms could be a helpful contribution to an international harmonization of the approval procedure. This paper will describe major influencing factors to be considered to assess the ageing behavior of a SFRM design and will identify the need for a regulatory specification of a SFRM-working life as basis for the assessment of the SFRM design regarding time-dependent weakening. A proposal for an explicit requirement for consideration of ageing mechanisms in safety assessment of SFRM, which should be considered in the ongoing SSR-6 revision cycle, will be explained. T2 - PATRAM22 Conference CY - Juan-Les-Pins, France DA - 11.06.2023 KW - Radioaktive Stoffe KW - Beförderung KW - Zulassung KW - Strahlenschutz PY - 2023 AN - OPUS4-57781 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Rolle, Annette A1 - Komann, Steffen A1 - Wille, Frank T1 - Ageing aspect in design approval of special form radioactive material N2 - In accordance with the IAEA transport regulations, the design of special form radioactive material (SFRM) shall resist a severe transport accident without undue loss or dispersal of radioactive material. The safety assessment for design approval includes besides the program for physical tests (impact, percussion, bending and heat test) also the evaluation of the management system for design, manufacture, testing, documentation, use, maintenance, and inspection. SFRM source design plus management system shall ensure, that every specimen of the approved design is able to survive the severe mechanical and thermal tests at any time of its SFRM-working life. Due to the long-term use of SFRM designs in most cases, the assessment of the source ageing behavior is an important aspect in the approval procedure. Different fields of application imply a wide range of environmental conditions, from clean room atmosphere to highly aggressive industrial conditions. Besides of radioactive content, corrosion is a main factor for possible SFRM design degradation. Although the IAEA Advisory Material SSG-26 already implies an indication of the need for considering ageing mechanisms, suitable amendments in the regulatory requirements of SSR-6 should be introduced to make the approval procedure more transparent and help to reduce rounds of questions by the authority. A supplementary requirement for considering of ageing mechanisms could be a helpful contribution to an international harmonization of the approval procedure. This paper will describe major influencing factors to be considered to assess the ageing behavior of a SFRM design and will identify the need for a regulatory specification of a SFRM-working life as basis for the assessment of the SFRM design regarding time-dependent weakening. A proposal for an explicit requirement for consideration of ageing mechanisms in safety assessment of SFRM, which should be considered in the ongoing SSR-6 revision cycle, will be explained. T2 - PATRAM22 Conference CY - Juan-Les-Pins, Antibes, France DA - 11.06.2023 KW - Ttransport KW - Radioactive material KW - Sealed sources KW - Ageing PY - 2023 SP - 1 EP - 6 AN - OPUS4-57786 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Quercetti, Thomas A1 - Scheidemann, Robert T1 - Test Facilities for Radioactive Materials Transport and Storage Packagings at BAM N2 - BAM acts as authority and for service in safety assessment of packages for transport and storage of radioactive materials. We offer extensive test capabilities and application of analytical methods for design verification and simulation for all types of packages for the transport and storage of radioactive materials according with the international IAEA Regulations for the safe transport and for national storage acceptance criteria. BAM operates several test facilities for drop and stacking testing, leak testing and thermal testing. The large drop test tower allows dropping full-scale specimens up to 200,000 kg in any drop orientation as requested. The comprehensive test facilities combined with long-term experience, newest equipment and measurement devices according to the latest state-of-the-art technology ensures realisation of complex test campaigns for package safety evaluation. Beyond that, non-destructive and destructive material test devices and experts are available. Equipment and application of all kinds of typical measurement categories can be offered for testing campaigns. In recent years we performed testing of full-scale type B package models with complex handling and preparation procedures. The results were contributed for different package design approval procedures. Type A packages mainly designed for medical related transport purposes, were continuously tested according to the transport regulations over recent years as well. Moreover, we work on research topics with relevance to package safety. The mechanical behaviour of lid closure systems under transport and storage conditions and the thermal behaviour of impact limiters were recently of special importance for the assessment competencies of BAM and were investigated under use of our test facilities. The paper describes the test facilities and capabilities for package design safety evaluation at BAM and shows examples from our recent work. T2 - PATRAM 2022, 20th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Juan-les-Pins, France DA - 11.06.2023 KW - IAEA KW - Fire test KW - Drop testing KW - Transport KW - Package PY - 2023 SP - 1 EP - 12 AN - OPUS4-57967 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - 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 - Zencker, Uwe A1 - Gaddampally, Mohan Reddy A1 - Völzke, Holger T1 - Fracture Mechanics Analysis of Spent Fuel Claddings during Long-Term Dry Interim Storage N2 - The prevention of brittle fracture of spent fuel claddings during long-term dry interim storage is based on experimental investigations, numerical analyses and assessment methods for predicting the mechanical behavior and determining limiting conditions. The ring compression test (RCT) is an established experimental method for characterizing cladding material. Test results for various high-burnup pressure water reactor zirconium-based fuel cladding alloys (e.g., ZIRLO®, M5®) are publicly available. To reduce the effort associated with irradiated samples in hot cells, it is helpful to perform studies on unirradiated surrogate cladding material. Based on such experimental data, load-displacement curves were numerically analyzed for selected cladding materials. In the presence of radial hydrides, a sample may suddenly fail by fracture even at small deformations. Noticeable load drops in the RCT occur associated to unstable crack propagation through the radial hydride network. The failure mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth, and coalescence in the zirconium matrix, with ductile tearing patches connecting neighboring hydrides. The failure process was simulated by cohesive zones controlled by the fracture energy and the cohesive strength. A modeling approach is presented in which the radial hydride morphology is taken into account. Based on the developed fracture mechanics approach with cohesive zone modeling, not only the deformation behavior but also the failure behavior of irradiated as well as unirradiated claddings with radial hydrides under RCT loading conditions can be adequately described. T2 - 20th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 2022) CY - Juan-les-Pins, France DA - 11.06.2023 KW - Extended Interim Storage KW - Nuclear Fuel Cladding KW - Numerical Failure Analysis KW - Ring Compression Test KW - Spent Fuel Characterization PY - 2023 SP - 1 EP - 8 AN - OPUS4-59146 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Fracture Mechanics Analysis of Spent Fuel Claddings during Long-Term Dry Interim Storage N2 - The prevention of brittle fracture of spent fuel claddings during long-term dry interim storage is based on experimental investigations, numerical analyses and assessment methods for predicting the mechanical behavior and determining limiting conditions. The ring compression test (RCT) is an established experimental method for characterizing cladding material. Test results for various high-burnup pressure water reactor zirconium-based fuel cladding alloys (e.g., ZIRLO®, M5®) are publicly available. To reduce the effort associated with irradiated samples in hot cells, it is helpful to perform studies on unirradiated surrogate cladding material. Based on such experimental data, load-displacement curves were numerically analyzed for selected cladding materials. In the presence of radial hydrides, a sample may suddenly fail by fracture even at small deformations. Noticeable load drops in the RCT occur associated to unstable crack propagation through the radial hydride network. The failure mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth, and coalescence in the zirconium matrix, with ductile tearing patches connecting neighboring hydrides. The failure process was simulated by cohesive zones controlled by the fracture energy and the cohesive strength. A modeling approach is presented in which the radial hydride morphology is taken into account. Based on the developed fracture mechanics approach with cohesive zone modeling, not only the deformation behavior but also the failure behavior of irradiated as well as unirradiated claddings with radial hydrides under RCT loading conditions can be adequately described. T2 - 20th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 2022) CY - Juan-les-Pins, France DA - 11.06.2023 KW - Extended Interim Storage KW - Nuclear Fuel Cladding KW - Numerical Failure Analysis KW - Ring Compression Test KW - Spent Fuel Characterization PY - 2023 AN - OPUS4-59147 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - EURAD WP 8 Task 3 Overview: Behaviour of nuclear fuel and cladding after discharge N2 - Task 3 of Work package 8 (Spent Fuel Characterization) of the European Joint Programme on Radioactive Waste Management (EURAD) investigates the behaviour of nuclear fuel and cladding after discharge. The aim of the work is to understand and describe the be-haviour of spent nuclear fuel (SNF), irradiated cladding, fuel/cladding chemical interaction (FCCI) and ageing under conditions of extended interim storage, transportation and em-placement in a final disposal system. BAM contributes to the project as partner and leads Task 3. The presentation gives an overview of the project status, main achievements in experimental work and modelling studies, deviations from the plan, delays and challenges ahead. T2 - EURAD Work Package 8 (SFC) Annual Meeting CY - Wettingen, Switzerland DA - 31.10.2023 KW - Nuclear fuel KW - Cladding KW - Discharge PY - 2023 AN - OPUS4-59148 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 - 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 - 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 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 - 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 - 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 - Weber, Mike A1 - Völzke, Holger A1 - Zencker, Uwe A1 - Herz, A. A1 - Kreienmeyer, M. T1 - Design of a Drop Test Target with Reproducible Properties for Konrad Prototype Testing N2 - Beginning in 1976 the former iron ore mine Konrad was geologically investigated concerning its suitability as a location for a German final repository for low and intermediate level waste. After a comprehensive licensing procedure it was approved and is now being prepared and equipped for the planned start of storage in 2027. At the end of the 1980s, the requirements for waste containers for storage in this repository were defined for the first time on the basis of an incident analysis of the on-site handling procedures, beginning with the unloading of the packages after the arrival at the site up to the final positioning in the underground storage galleries. A 5-meter drop onto the rock ground of the mine was identified as the covering case for high mechanical requirements (ABK II container class). In contrast to the 9-meter drop according to the IAEA Regulations for the Safe Transport of Radioactive Material, the 5-meter drop is performed onto a hard but not essentially unyielding target. The container is typically not protected by an impact limiter. The requirements for a potential test facility are described in the regulations for the Konrad repository. The mechanical strength of the target is defined as a concrete strength equivalent to the identified properties of the rock ground of the mine. Since 1991 BAM has consistently used precast reinforced concrete slabs as target for drop tests in the framework of licensing procedures as well as in research projects. While the original design fulfilled the requirement for the integrity of the concrete slab in most cases, it failed when drop tests onto an edge of a container were performed. A redesigned concrete slab developed in a research project and suggested as a reference target in 2009 has been successfully used in Konrad licensing procedures since then. The paper gives a brief overview of the historical development and it describes BAM’s efforts and approaches to continuously guarantee a concrete slab of defined quality and to provide a test setup for valid drop tests from the applicants and the authorities view. T2 - PATRAM 2022, 20th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Juan-les-Pins, France DA - 11.06.2023 KW - Target KW - Drop test KW - Final repository Konrad PY - 2023 SP - 1 EP - 11 AN - OPUS4-58563 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike T1 - Design of a Drop Test Target with Reproducible Properties for Konrad Prototype Testing N2 - Beginning in 1976 the former iron ore mine Konrad was geologically investigated concerning its suitability as a location for a German final repository for low and intermediate level waste. After a comprehensive licensing procedure it was approved and is now being prepared and equipped for the planned start of storage in 2027. At the end of the 1980s, the requirements for waste containers for storage in this repository were defined for the first time on the basis of an incident analysis of the on-site handling procedures, beginning with the unloading of the packages after the arrival at the site up to the final positioning in the underground storage galleries. A 5-meter drop onto the rock ground of the mine was identified as the covering case for high mechanical requirements (ABK II container class). In contrast to the 9-meter drop according to the IAEA Regulations for the Safe Transport of Radioactive Material, the 5-meter drop is performed onto a hard but not essentially unyielding target. The container is typically not protected by an impact limiter. The requirements for a potential test facility are described in the regulations for the Konrad repository. The mechanical strength of the target is defined as a concrete strength equivalent to the identified properties of the rock ground of the mine. Since 1991 BAM has consistently used precast reinforced concrete slabs as target for drop tests in the framework of licensing procedures as well as in research projects. While the original design fulfilled the requirement for the integrity of the concrete slab in most cases, it failed when drop tests onto an edge of a container were performed. A redesigned concrete slab developed in a research project and suggested as a reference target in 2009 has been successfully used in Konrad licensing procedures since then. The paper gives a brief overview of the historical development and it describes BAM’s efforts and approaches to continuously guarantee a concrete slab of defined quality and to provide a test setup for valid drop tests from the applicants and the authorities view. T2 - PATRAM 2022, 20th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Juan-les-Pins, France DA - 11.06.2023 KW - Final repository Konrad KW - Drop test KW - Target PY - 2023 AN - OPUS4-58564 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike T1 - Design of a Drop Test Target with Reproducible Properties for Konrad Prototype Testing N2 - Beginning in 1976 the former iron ore mine Konrad was geologically investigated concerning its suitability as a location for a German final repository for low and intermediate level waste. After a comprehensive licensing procedure it was approved and is now being prepared and equipped for the planned start of storage in 2027. At the end of the 1980s, the requirements for waste containers for storage in this repository were defined for the first time on the basis of an incident analysis of the on-site handling procedures, beginning with the unloading of the packages after the arrival at the site up to the final positioning in the underground storage galleries. A 5-meter drop onto the rock ground of the mine was identified as the covering case for high mechanical requirements (ABK II container class). In contrast to the 9-meter drop according to the IAEA Regulations for the Safe Transport of Radioactive Material, the 5-meter drop is performed onto a hard but not essentially unyielding target. The container is typically not protected by an impact limiter. The requirements for a potential test facility are described in the regulations for the Konrad repository. The mechanical strength of the target is defined as a concrete strength equivalent to the identified properties of the rock ground of the mine. Since 1991 BAM has consistently used precast reinforced concrete slabs as target for drop tests in the framework of licensing procedures as well as in research projects. While the original design fulfilled the requirement for the integrity of the concrete slab in most cases, it failed when drop tests onto an edge of a container were performed. A redesigned concrete slab developed in a research project and suggested as a reference target in 2009 has been successfully used in Konrad licensing procedures since then. The paper gives a brief overview of the historical development and it describes BAM’s efforts and approaches to continuously guarantee a concrete slab of defined quality and to provide a test setup for valid drop tests from the applicants and the authorities view. T2 - ASME 2023 Pressure Vessels & Piping Conference (PVP2023) CY - Atlanta, GA, USA DA - 16.07.2023 KW - Final repository Konrad KW - Drop test KW - Target PY - 2023 AN - OPUS4-58565 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Spent Fuel Characterization - Current Activities at BAM N2 - The European Joint Programme on Radioactive Waste Management (EURAD) is working on Spent Fuel Characterization (SFC) in its work package (WP) 8. Inspired by the EURAD activities, the International Atomic Energy Agency (IAEA) established an international Coordinated Research Project (CRP) on SFC. The EURAD WP SFC participants are collaborating as a team on the IAEA CRP on SFC. The EURAD WP SFC project consists of four tasks. Task 3 investigates the behaviour of nuclear fuel and cladding after discharge. The aim of these activities is to understand and describe the evolution of the cladding-pellet system and its ageing under conditions of extended interim storage, transportation and emplacement in a final disposal system. At a Consultancy Meeting, BAM as contributor to Task 3 presented current results of the failure analysis of irradiated ZIRLO® claddings under conditions of the Ring Compression Test. T2 - IAEA Consultancy Meeting on the Coordinated Research Project on Spent Fuel Characterization CY - Online meeting DA - 12.09.2023 KW - Nuclear Fuel Cladding KW - Numerical Failure Analysis KW - Ring Compression Test KW - Spent Fuel Characterization KW - Extended Interim Storage PY - 2023 AN - OPUS4-58276 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Rolle, Annette A1 - Neumeyer, Tino A1 - Ballheimer, Viktor A1 - Wille, Frank T1 - Investigations of Aged Metal Seals for Transport Package Safety Assessment N2 - Acceptable limits for activity release from transport casks for high-level radioactive material specified in the IAEA regulations must be kept by the integrity of cask body and the cask sealing system. BAM as the German competent authority for mechanical, thermal and containment assessment of packages liable for approval verifies the activity release compliance with the regulatory limits. One of the fundamental aspects in assessment is the specification of conservative package design leakage rates. To ensure the required package tightness for both, storage, and transport of the cask before and after storage usually metal seals of the Helicoflex® Type are used. Due to the long-term use the seal behavior is influenced by temperature and time. The mechanical and thermal loadings associated with the routine, normal and accident conditions of transport specified in the regulations can have a significant effect on the leak tightness of the sealing system. Whereas the safety for application of new, non- aged Helicoflex® seals is verified sufficiently, there are still technical data gaps concerning the efficiency of aged Helicoflex® seals. BAM performed experiments to learn more about the sealing efficiency of aged Helicoflex® seals with Aluminum and Silver outer jackets. The seals were compressed in test-flanges and for artificial ageing the complete flange systems were stored in an oven for several month at a high temperature. During the compression and decompression tests after the aging, load-deformation characteristics of the seals, and leakage rates were measured. With these tests a load situation was simulated, which can occur in the regulatory drop test of the cask: Under high impact loads the bolted lid can lift a little for a short moment, allowing a little movement of the seal, so that the contact area can change before compressing again. The poster presentation will show details about test conditions and first results. T2 - Interdisciplinary research symposium on the safety of nuclear disposal practices safeND2023 CY - Berlin, Germany DA - 13.09.2023 KW - Radioactive material KW - Sealing KW - Ageing KW - Leaktightnes PY - 2023 AN - OPUS4-58436 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 - JOUR A1 - Sjöland, A. A1 - Christensen, P. A1 - Zetterström Evins, L. A1 - Bosbach, D. A1 - Duro, L. A1 - Farnan, I. A1 - Metz, V. A1 - Zencker, Uwe A1 - Ruiz-Hervias, J. A1 - Rodriguez-Villagra, N. A1 - Kiraly, M. A1 - Schillebeeckx, P. A1 - Rochman, D. A1 - Seidl, M. A1 - Dagan, R. A1 - Verwerft, M. A1 - Herranz Puebla, L. E. A1 - Hordynskyi, D. A1 - Feria, F. A1 - Vlassopoulos, E. T1 - Spent nuclear fuel management, characterisation, and dissolution behaviour: progress and achievement from SFC and DisCo N2 - SFC is a work package in Eurad that investigates issues related to the properties of the spent nuclear fuel in the back-end of the nuclear fuel cycle. Decay heat, nuclide inventory, and fuel integrity (mechanical and otherwise), and not least the related uncertainties, are among the primary focal points of SFC. These have very significant importance for the safety and operational aspect of the back-end. One consequence is the operation economy of the back-end, where deeper understanding and quantification allow for significant optimization, meaning that significant parts of the costs can be reduced. In this paper, SFC is described, and examples of results are presented at about half-time of the work package, which will finish in 2024. The DisCo project started in 2017 and finished in November 2021 and was funded under the Horizon 2020 Euratom program. It investigated if the properties of modern fuel types, namely doped fuel, and MOX, cause any significant difference in the dissolution behavior of the fuel matrix compared with standard fuels. Spent nuclear fuel experiments were complemented with studies on model materials as well as the development of models describing the solid state, the dissolution process, and reactive transport in the near field. This research has improved the understanding of processes occurring at the interface between spent nuclear fuel and aqueous solution, such as redox reactions. Overall, the results show that from a long-term fuel matrix dissolution point of view, there is no significant difference between MOX fuel, Cr+Al-doped fuel, and standard fuels. KW - Spent nuclear fuel management KW - Spent fuel characterization KW - Dissolution behaviour PY - 2023 UR - https://nbn-resolving.org/urn:nbn:de:kobv:b43-572461 DO - https://doi.org/10.1051/epjn/2022029 SN - 2491-9292 VL - 9 SP - 1 EP - 12 PB - EDP Sciences CY - Les Ulis Cedex, France AN - OPUS4-57246 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger T1 - Container requirements for high-level radioactive waste disposal in rock salt, claystone, and crystalline rock – Outcomes of the research project KoBrA N2 - Suitable containers for receiving high-level radioactive waste are a key elements of every repository system concerning different host rocks like rock salt, claystone and crystalline rock as being considered in Germany. The disposal container design is decisive regarding boundary conditions for the transport, emplacement techniques, and the operational and long-term safety assessment. The systematic derivation of the requirements to be placed on disposal containers forms the basis for a targeted, comprehensible and transparent development of containers that meet the requirements. The presentation summarizes the outcomes of the research project KoBrA. T2 - IGD-TP Symposium CY - Zurich, Switzerland DA - 20.09.2022 KW - Disposal KW - Container KW - Radioactive waste KW - Safety PY - 2022 AN - OPUS4-57021 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - 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 - RPRT A1 - Zencker, Uwe A1 - Aguado, Carlos A1 - Arkoma, Asko A1 - Bertsch, Johannes A1 - Cristóbal Beneyto, Miguel A1 - Feria Marquez, Francisco A1 - Herranz, Luis Enrique A1 - Ivanov, Ivan A1 - Király, Márton A1 - Nasyrow, Ramil A1 - Papaioannou, Dimitrios A1 - Ruiz-Hervias, Jesus A1 - Vlassopoulos, Efstathios T1 - Thermo-mechanical-chemical properties of unirradiated and irradiated samples of spent nuclear fuel rod segments and cladding N2 - The report addresses the experimental characterisation, thermo-mechanical modelling and performance of unirradiated and irradiated samples of spent nuclear fuel rod segments and cladding. Hydrogen is relocated in nuclear fuel claddings by diffusion along temperature, stress, and concentration gradients, and towards a liner if existent. The precipitation of hydrides during cool-down at the beginning of the storage is influenced by mechanical stress. A strong accumulation of hydrides and a disadvantageous hydride alignment in relation to the stress state can affect the mechanical properties and compromise the fuel cladding integrity. Neutron radiography is used to determine spatially resolved hydrogen and hydride distributions in claddings. A liner in duplex claddings attracts hydrogen, which will then no longer be available to form detrimental hydrides in the bulk part of the cladding. Individual three-point bending and gravitational impact tests on filled pressurised spent nuclear fuel rod segments were conducted in hot-cell facilities. The material behaviour under three-point bending in the presence of hydrides and at different temperatures was systematically investigated in quasi-static laboratory tests on unirradiated rod segments with and without surrogate pellets until rupture. Ring compression tests were carried out on irradiated and unirradiated samples at different temperatures and both with and without radial hydrides. With radial hydrides, brittle failure is possible even at low deformation and simultaneously low temperature. Crack initiation will occur on the radial hydride with the highest fracture mechanics load, which depends on the hydride size in the regions with the highest hoop stresses. Due to the high hydrogen content used in creep tests conducted, precipitation hardening of the hydrides in the zirconium alloy significantly reduces the creep rate, as it strengthens the material and inhibits the dislocation sliding. The use of finite element analyses on mechanical tests to determine the cladding properties using a numerical optimisation method based on a large number of automatically generated sensitivity calculations is demonstrated. The brittle fracture of cladding samples with radial hydrides in the ring compression test can be described with a cohesive zone model if the cohesive parameters are selected appropriately and the hydride morphology is specified. Progress was made in the verification, validation and enhancement of spent fuel performance codes by the compilation of a representative validation database (related to rod internal pressure, fission gas release, rod void volume, and in-clad hydrogen distribution), the enhancement of CIEMAT’s FRAPCON-xt code for rod internal pressure prediction at high burnup, the enhancement of CIEMAT’s in-clad hydrogen performance subroutine HYDCLAD with more phenomenological modelling, the extension of INL’s BISON code with a cladding creep law for dry storage, and a benchmark with FRAPCON-xt and BISON simulating in-reactor and dry storage scenarios. KW - Spent Nuclear Fuel KW - Cladding KW - Experimental Characterisation KW - Modelling KW - Fuel Performance Code PY - 2024 SP - 1 EP - 102 PB - European Joint Programme EURAD AN - OPUS4-61008 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 -