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 behaviour of spent nuclear fuel (SNF), irradiated cladding, fuel/cladding chemical interaction (FCCI) and ageing under conditions of extended interim storage, transportation and emplacement 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 - Kalmar, Sweden DA - 19.09.2022 KW - Nuclear fuel KW - Cladding KW - Discharge PY - 2022 AN - OPUS4-56120 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - EURAD WP 8 (SFC) Task 3 - Update: Behaviour of nuclear fuel and cladding after discharge 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 the Second Research Coordination Meeting on Spent Fuel Characterization, BAM as leader of Task 3 reported on the status of the research work. T2 - IAEA 2nd Research Coordination Meeting of the Coordinated Research Project on Spent Fuel Characterization CY - Kalmar, Sweden DA - 20.09.2022 KW - Nuclear fuel KW - Cladding KW - Discharge PY - 2022 AN - OPUS4-56121 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias A1 - Neumann, Martin A1 - Linnemann, Konrad A1 - Wille, Frank T1 - Fracture Mechanical Investigations on a Welding Seam of a Thick-Walled Transport Package N2 - Untersuchung einer Schweißnaht mit verschiedenen Codes. Vergleiche von experimentellen und numerischen Ergebnissen T2 - Technical Exchange IRSN – BAM: Transport & Storage of Packages for Radioactive Material CY - Cadarache, France DA - 13.10.2022 KW - Drop test KW - Fracture initiation KW - Transport package PY - 2022 AN - OPUS4-56055 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Rolle, Annette A1 - Neumeyer, Tino A1 - Wille, Frank T1 - Safety assessment of aged metal seals N2 - With our experiments we want to simulate a load situation which can result from a cask vertical drop test. The lid can lift a little for a very short moment and that can perhaps result in a little movement of the seal, a little rotation or a little lateral movement, so that the contact section can change. The aim of the project is to learn more about the sealing efficiency after the seal is compressed again. What we know from former component tests with non-aged seals is, that the leakage rate measured after one or more decompression /compression cycles with seal movement becomes significant higher. And this result is considered for specification of design leakage rates for release calculation in safety cases. In the recent years we all have learned more about the significant influence of temperature and time on seal behavior and so the plan was to do similar experiments with aged seals, just to get an idea about seal behavior and achievable leakage rates. What we did in detail is: We compressed Helicoflex-seals of both design types, with Aluminum and Silver outer jackets, in testflanges and aged them in an oven at a temperature of 125°C for 3 months to produce seal properties comparable with properties after several years use at more typical operating temperatures between 90 and 100°C. After this ageing procedure we opened the flanges completely, moved the seal a little to vary the contact area and compressed the flanges again. What we measured during all compression and decompression cycles was the load, the deformation and -as long as possible- the leakage rates. T2 - Workshop IRSN/BAM CY - Cadarache, France DA - 12.10.2022 KW - Transport KW - Radioactive KW - Seal behavior KW - Ageing PY - 2022 AN - OPUS4-56141 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kömmling, Anja A1 - Wolff, Dietmar A1 - Jaunich, Matthias T1 - Analysis of O-ring seal failure in the context of radioactive waste containers N2 - At BAM, which is a federal institute for materials research and testing in Germany, it is one of our tasks to evaluate the safety of casks designed for transport and/or storage of radioactive material. This includes assessment of elastomeric seals applied in the casks. Besides examining the low-temperature behaviour and irradiation effects of elastomeric seals, it is our goal to estimate the service lifetime of the seals with regard to the requirements for long-term safety (40 years and more) of the containers. Therefore, we started an accelerated ageing programme with selected rubbers often used for seals (HNBR, EPDM and FKM) which are aged at four different temperatures (75 °C, 100 °C, 125 °C and 150 °C) up to five years. In order to assess sealability, O-rings were aged in compression by 25 % (corresponding to the compression during service) between plates as well as in flanges that allow leakage rate measurements. For comparison, uncompressed O-rings were aged as well. Further methods characterising seal performance are compression stress relaxation (CSR) reflecting the loss of sealing force of a compressed seal over time, and compression set (CS) which represents the recovery behaviour of a seal after release from compression. Additionally, material properties such as hardness, elastic modulus, glass transition temperature and viscoelastic loss factor as well as relaxation and recovery behaviour are examined in order to understand the underlying ageing mechanisms in each material. For obtaining results closely related to practical conditions, O-rings with a full-scale cord diameter of 10 mm were aged. However, this set-up can lead to heterogeneous aging caused by diffusion-limited oxidation (DLO) effects, resulting in distorted bulk properties such as compression stress relaxation and compression set. However, if DLO-affected data is excluded, extrapolations of CS data are possible using time-temperature shifts and Arrhenius graphs. For selecting an appropriate end-of-lifetime criterion, leakage rate measurements were performed, since leakage rate is the only characteristic directly correlated to the performance of the sealing system. A significant increase in leakage rate was considered as the end of the lifetime. However, the O-rings remained leak tight under static conditions and even exhibited an improved, i.e. decreased leakage rate while other properties already indicated strong deterioration. These tests were extended by seal tests with fast partial decompression. Overall an update will be given on the current investigations and results and the planned activities. T2 - Kautschuk Herbst Kolloquium 2022 CY - Online meeting DA - 08.11.2022 KW - Seal KW - Ageing KW - Compression set PY - 2022 AN - OPUS4-56397 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Kömmling, Anja A1 - Grelle, Tobias A1 - Jaunich, Matthias A1 - Goral, Milan A1 - Wolff, Dietmar T1 - Three-dimensional thermal expansion of neat and irradiated (U)HMWPE materials at elevated temperatures JF - Polymer Testing N2 - The thermal expansion of polymeric parts can be an issue in many applications where the available space is limited, or exact dimensions of the part are required. For this study, a device was designed and built that allowed measuring the thermal expansion simultaneously in all three spatial directions on cubic samples with real-scale dimensions (78 mm edge length). The results are shown between 25 °C and 125 °C for two PE materials, one HMWPE and one tempered UHMWPE, for non-irradiated samples as well as cubes that have been irradiated with 100 and 400 kGy. The results measured with the new device were very similar to those measured with conventional thermo-mechanical analysis equipment and to literature data of UHMWPE. The HMWPE material shows a much larger thermal expansion coefficient in one direction compared to the other two directions during the first heating due to frozen stresses from the pressing step during material manufacturing. These stresses are mostly released by the expansion during the first heating, so that the expansion during the second heating is more uniform. The overall volumetric expansion is the same for both heating runs. By contrast, the tempered UHMWPE material shows no significant difference between first and second heating run, as the stresses from processing could already relax in the tempering step. The irradiation treatment does not affect the values significantly for the given test set-up. KW - Lupolen KW - Ultra high molecular weight polyethylene KW - GUR KW - Coefficient of thermal expansion KW - High temperature PY - 2022 UR - https://nbn-resolving.org/urn:nbn:de:kobv:b43-563987 DO - https://doi.org/10.1016/j.polymertesting.2022.107841 SN - 0142-9418 VL - 117 SP - 1 EP - 8 PB - Elsevier Ltd. AN - OPUS4-56398 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias A1 - Zencker, Uwe A1 - Wolff, Dietmar T1 - Investigations at BAM on Fuel Cladding Integrity and DPC Seal Performance N2 - The presentation provides an update on preliminary results from research projects in the area of long term performance of metal seals and fuel rod integrity as safety relevant components of spent fuel transport and storage casks for spent nuclear fuel. T2 - Extended Storage Collaboration Program (ESCP) winter meeting 2022 CY - Charlotte, NC, USA DA - 07.11.2022 KW - Metal seal KW - Fuel cladding KW - Ring compression test KW - Brittle failure PY - 2022 AN - OPUS4-56400 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - 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 T2 - Proceedings of the ASME 2021 Pressure Vessels & Piping Conference (PVP2021) 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 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 - 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 - Neumann, Martin A1 - Gleim, Tobias A1 - Gradt, Thomas T1 - Friction coefficients for wood-wood and wood-steel contacts N2 - The friction at the interfaces between pieces of wood and steel is of crucial importance for the impact and energy absorption in impact limiters of transport casks for radioactive material. Here, the friction coefficient for the combinations wood-wood and wood-steel was measured in the temperature range between -40°C and 90°C. Results show decreasing friction with increasing temperature, ranging from 0.43 at -40°C to 0.22 for 90°C for wood-steel combinations and from 0.3 at -40°C to 0.24 at 90°C for a wood-wood combination. T2 - Sandia-BAM Workshop CY - Berlin, Germany DA - 21.11.2022 KW - Wood KW - Friction KW - Transport cask PY - 2022 AN - OPUS4-56342 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gröke, Carsten A1 - Neumann, Martin A1 - Komann, Steffen A1 - Wille, Frank T1 - Requirements for Package Design Testing and Quality Assurance Measures for Packages not Requiring Competent Authority Design Approval N2 - Übersicht über die gefahrgutrechtlichen Anforderungen zur Bauartprüfung und dem Qualitätsmanagement für Verpackungen für nicht zulassungspflichtige Versandstücke. T2 - KONTEC 2019 CY - Dresden, Germany DA - 27.03.2019 KW - Bauartprüfung KW - Qualitätsmanagement KW - Nicht zulassungspflichtig PY - 2019 AN - OPUS4-56352 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Zaghdoudi, Maha A1 - Kömmling, Anja A1 - Jaunich, Matthias A1 - Wolff, Dietmar T1 - Oxidative ageing of elastomers: Experiment and modelling JF - Continuum mechanics and thermodynamics N2 - During an extensive test programme at the Bundesanstalt für Materialforschung und prüfung, material property changes of EPDM O-rings were investigated at different ageing times and two ageing temperatures of 125∘C and 150∘C. To exclude possible diffusion-limited oxidation (DLO) effects that can distort the data, IRHD microhardness measurements were taken over the cross section of compressed O-rings. Continuous stress relaxation measurements were taken on samples free of DLO effects. The additional effect of physical processes to irreversible chemical ones during a long-term thermal exposure is quantified by the analysis of compression set measurements under various test conditions. By combining the different experimental methods, characteristic times relative to the degradation processes were determined. On the basis of experimental data, a microphysically motivated model that takes into account reversible and irreversible processes was developed. The parameter identification strategy of the material model is based on our experimental investigations on homogeneously aged elastomer O-rings. The simulated results are in good agreement with the experiments. KW - Compression stress relaxation KW - Compression set KW - IRHD microhardness KW - Modelling PY - 2022 UR - https://nbn-resolving.org/urn:nbn:de:kobv:b43-545910 DO - https://doi.org/10.1007/s00161-022-01093-9 SN - 1432-0959 SP - 1 EP - 9 PB - Springer CY - Berlin AN - OPUS4-54591 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Zaghdoudi, Maha A1 - Kömmling, Anja A1 - Jaunich, Matthias A1 - Wolff, Dietmar T1 - A numerical approach to correlate compression stress relaxation and compression set of elastomer O-rings with tightness T2 - Proceedings of the ASME 2021 Pressure Vessels & Piping Conference (PVP2021) N2 - The excellent mechanical properties of elastomer seals at a wide range of temperatures as well as their high versatility and recovery potential under several load conditions make these materials well suitable for the application in containers designed for transport and disposal of negligible heat generating radioactive waste. While a seal exchange at defined intervals is typical in many conventional applications, it is impossible or at least hard to perform when principles of minimization of radiation exposure have to be considered which prohibit an avoidable cask handling. An extensive knowledge of the change of the elastomer properties during aging and the availability of reliable end-of-lifetime criteria to guarantee the safe enclosure of the radioactive material for the required time are mandatory. As BAM is involved in most of the national cask licensing procedures and in the evaluation of cask-related long-term safety issues, great efforts have already been made and are still ongoing to scientifically support this task. Among other representative types of elastomers, specimen made from ethylene propylene diene rubber (EPDM) were tested before, during and after aging to capture the with respect to application most important of their complex mechanical properties. Exemplary results of these investigations were used to calibrate material models implemented in the commercial finite element software ABAQUS/Standard®. The finite element model already presented in previous works uses a sequential temperature displacement coupling. The calculated compression stress relaxation (CSR) and compression set (CS) values do satisfactorily match the experimental results. In many investigations performed at BAM both values (CSR and CS) were identified as key indicators of elastomer’s long-term performance. However, the possibility to correlate these equivalent indicators with performance values such as tightness and leakage rate, measurable in the mounted state, is an important goal of our future work. In the presented study the ABAQUS® feature of “pressure penetration” is introduced in the suggested finite element model for this purpose. It provides the possibility to simulate the penetration of a gas into a possible gap between flange and O-ring causing an opening of a leakage path. Three dimensional and axis-symmetric finite element models were generated to represent flat and grooved flanges of different dimensions. The sensitivity of the feature to several input parameters is investigated and the observed behavior of the O-ring is correlated with the results of performed leakage tests. T2 - ASME 2021 Pressure Vessels & Piping Conference (PVP2021) CY - Online meeting DA - 13.07.2021 KW - Elastomer KW - Tightness KW - Leakage KW - Compression set KW - Compression stress relaxation PY - 2021 SN - 978-0-7918-8535-2 DO - https://doi.org/10.1115/PVP2021-61976 SP - 1 EP - 8 PB - American Society of Mechanical Engineers (ASME) CY - New York, NY, USA AN - OPUS4-54422 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Qiao, Linan A1 - Nagelschmidt, Sven A1 - Herbrich, Uwe A1 - Keller, Christian T1 - Introduction of a Power Law Time-Temperature Equivalent Formulation for the Description of Thermorheologically Simple and Complex Behavior JF - Materials N2 - Abstract: In this work, a conceptual framework is suggested for analyzing thermorheologically simple and complex behavior by using just one approach. Therefore, the linear relation between master time and real time which is required in terms of the time-temperature superposition principle was enhanced to a nonlinear equivalent relation. Furthermore, we evaluate whether there is any relation among well-known existing time-temperature equivalent formulations which makes it possible to generalize different existing formulations. For this purpose, as an example, the power law formulation was used for the definition of the master time. The method introduced here also contributes a further framework for a unification of established time-temperature equivalent formulations, for example the time-temperature superposition principle and time-temperature parameter models. Results show, with additional normalization conditions, most of the developed time-temperature parameter models can be treated as special cases of the new formulation. In the aspect of the arrow of time, the new defined master time is a bended arrow of time, which can help to understand the corresponding physical meaning of the suggested method. KW - bended arrow of time KW - time-temperature superposition principle KW - time-temperature equivalent formulation PY - 2022 UR - https://nbn-resolving.org/urn:nbn:de:kobv:b43-543800 DO - https://doi.org/10.3390/ma15030726 VL - 15 IS - 3 SP - 1 EP - 11 PB - MDPI AN - OPUS4-54380 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kömmling, Anja A1 - Jaunich, Matthias A1 - Wolff, Dietmar T1 - Aging of elastomer O-rings and PE neutron shielding materials for radioactive waste containers N2 - Our institution BAM in Berlin, Germany is concerned with research and testing of materials in the context of safety in chemistry and technology. Our working group is involved in the licensing procedures of casks for radioactive waste. Besides, we’re doing research on aging and lifetime prediction of elastomer O-rings and investigate degradation and thermal expansion of PE neutron shielding materials. T2 - Polymers in nuclear applications CY - Online meeting DA - 01.12.2021 KW - Rubber KW - Polyethylene KW - Thermal expansion KW - Lifetime KW - Degradation PY - 2021 UR - https://energiforsk.se/media/30631/bam_material_aging_analysis.pdf AN - OPUS4-54191 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - 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 - Zaghdoudi, Maha A1 - Kömmling, Anja A1 - Jaunich, Matthias A1 - Wolff, Dietmar T1 - Oxidative ageing of elastomers: experiment and modelling N2 - During an extensive test programme at the Bundesanstalt für Materialforschung und prüfung, material property changes of EPDM O-rings were investigated at different ageing times and an ageing temperature of 125 °C. To exclude possible diffusion limited oxidation (DLO) effects that can distort the data, IRHD micro hardness measurements were performed over the cross-section of compressed O-rings. Continuous stress relaxation measurements were conducted on samples free of DLO effects. The additional effect of physical processes to irreversible chemical ones during a long-term thermal exposure is quantified by the analysis of compression set measurements under various test conditions. By combining the different experimental methods, characteristic times relative to the degradation processes were determined. On the basis of experimental data, a microphysically motivated model that takes into account reversible and irreversible processes was developed. The parameter identification strategy of the material model is based on our experimental investigations on homogeneously aged elastomer O-rings. The simulated results are in good agreement with the experiments. T2 - ACEX-14th International Conference on Advanced Computational Engineering and Experimenting CY - Malta DA - 04.07.2021 KW - Compression stress relaxation KW - Compression set KW - IRHD micro hardness KW - Modelling PY - 2021 AN - OPUS4-52968 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Linnemann, Konrad A1 - Ballheimer, Viktor A1 - Sterthaus, Jens A1 - Rolle, Annette A1 - Wille, Frank A1 - Vlassopoulos, Efstathios A1 - Papaioannou, Dimitrios T1 - Numerical Simulation of Spent Fuel Segments under Transport Loads T2 - 17th International High-Level Radioactive Waste Management Conference (IHLRWM 2019) N2 - Packages for the transport of spent nuclear fuel shall meet the International Atomic Energy Agency regulations to ensure safety under different transport conditions. The physical state of spent fuel and the fuel rod cladding as well as the geometric configuration of fuel assemblies are important inputs for the evaluation of package capabilities under these conditions. Generally, the mechanical behavior of high burn-up spent fuel assemblies under transport conditions shall be analyzed with regard to the assumptions which are used in the containment and criticality safety analysis. Considering the complexity of the interactions between the fuel rods as well as between the fuel assemblies, basket, and cask containment, the exact mechanical analysis of such phenomena is nearly impossible. The gaps in Information concerning the material properties of cladding and pellet behavior, especially for the high burn-up fuel, make the analysis more complicated additionally. As a result, enveloping analytical approaches are usually used by BAM within the safety assessment of packages approved for transport of spent nuclear fuel. To justify the safety margins of such approaches additional analyses are necessary. In this paper, numerical simulations of a spent fuel assembly Segment are presented. The segment modeled represents the part of a generalized BWR fuel assembly between two spacers. Dynamic and quasi-static finite element calculations are performed to simulate the spent fuel behavior under regulatory defined accident conditions of transport. Beam elements are used for the modeling of the fuel rods representing the compound consisting of claddings and fuel pellets. The dynamic load applied is gathered from an experimental drop test with a spent fuel cask performed at BAM. A hot cell bending test performed at JRC Karlsruhe is the basis for obtaining the material behavior of the fuel rods. The material properties are determined by simulating the test setup of JRC and optimizing the results to fit the experimental load deflection curve. The simulations of the fuel assembly segment are used to get a better understanding about the loads on fuel rods under accident conditions of transport. T2 - 17th International High-Level Radioactive Waste Management Conference (IHLRWM 2019) CY - Knoxville, Tennessee, USA DA - 14.04.2019 KW - Spent Nuclear Fuel KW - Finite Element Simulation KW - Transport packages PY - 2019 SN - 978-1-51088-669-8 SP - 1 EP - 7 AN - OPUS4-52046 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Linnemann, Konrad A1 - Ballheimer, Viktor A1 - Sterthaus, Jens A1 - Rolle, Annette A1 - Wille, Frank A1 - Vlassopoulos, Efstathios A1 - Papaioannou, Dimitrios T1 - Numerical simulation of spent fuel segments under transport loads N2 - Packages for the transport of spent nuclear fuel shall meet the International Atomic Energy Agency regulations to ensure safety under different Transport conditions. The physical state of spent fuel and the fuel rod cladding as well as the geometric configuration of fuel assemblies are important inputs for the evaluation of package capabilities under these conditions. Generally, the mechanical behavior of high burn-up spent fuel assemblies under Transport conditions shall be analyzed with regard to the assumptions which are used in the containment and criticality safety analysis. In view of the complexity of the interactions between the fuel rods as well as between the fuel assemblies, basket, and cask containment, the exact mechanical analysis of such phenomena is nearly impossible. The gaps in information concerning the material properties of cladding and pellet behavior, especially for the high burn-up fuel, make the analysis more complicated additionally. As a result, enveloping analytical approaches are usually used by BAM within the safety assessment of packages approved for transport of spent nuclear fuel. To justify the safety margins of such approaches additional analyses are necessary. In this paper, numerical simulations of a segment of a spent fuel assembly are presented. The segment modeled represents the part of a generalized BWR fuel assembly between two spacers. Explicit dynamic finite element calculations are performed to simulate the spent fuel behavior under regulatory defined accident conditions of transport. A beam element formulation is used for the modeling of the fuel rods representing the compound consisting of claddings and fuel pellets. The load applied is gathered from experimental drop tests with spent fuel casks performed at BAM. A hot cell bending test performed at JRC Karlsruhe is the basis for obtaining the material behavior of the fuel rods. The material properties are determined by simulating the test setup of JRC and optimizing the results to fit the experimental load deflection curve. The simulations of the fuel Assembly segment are used to get a better understanding about the loads on fuel rods under accident conditions of transport. T2 - 17th International High-Level Radioactive Waste Management Conference (IHLRWM 2019) CY - Knoxville, Tennessee, USA DA - 14.04.2019 KW - Spent Nuclear Fuel KW - Finite Element Simulation KW - Transport packages PY - 2019 AN - OPUS4-52047 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin A1 - Quercetti, Thomas A1 - Musolff, André A1 - Gleim, Tobias A1 - Nehrig, Marko A1 - Wille, Frank 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 A1 - Quercetti, Thomas A1 - Sterthaus, Jens A1 - Ballheimer, Viktor T1 - Prestressed bolt connections under lateral displacements N2 - The containment system of transport packages for spent nuclear fuel and high-level waste usually includes bolted lids with metal gaskets. The specified transport condition imply high loading on the lids and the bolt connections of the package. The response of the lid systems on these load conditions is generally investigated by drop tests or numerically. BAM has started a research project to get a better understanding about the behavior of prestressed bolt connection under loadings typical for drop tests. T2 - 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 - Schubert, Sven A1 - Reichardt, Adrian A1 - Wille, Frank A1 - Komann, Steffen A1 - Neumann, Martin T1 - Introduction of an ageing management approach for packages for the transport of radioactive materials T2 - Proceedings of the INMM & ESARDA Joint Annual Meeting N2 - With integration of the new para 613A into SSR-6 [1] the consideration of ageing mechanisms is now obligatory for the design of packages and their approval. In addition, para 809(f) of SSR-6 [1] requires consideration of the effects of storage on ageing mechanisms, safety analyses and operation and maintenance instructions. German competent authorities Bundesanstalt für Materialforschung und -prüfung (BAM) and Bundesamt für die Sicherheit der nuklearen Entsorgung (BASE) are considering the aspect of ageing in approval procedures. Ageing assessment is mainly focused on dual purpose casks (DPC) package designs which are long-term stored in interim storage facilities. For these package designs, the evaluation of ageing management is now mandatory for the maintenance of the package design approvals with a validity period of 5 years and beyond. The ageing management includes amongst others a gap analysis, the assessment of ageing effects and operational experiences during operation and interim storage. BAM works on the compilation of a guideline for implementation of paras 613A, 809(f) and 809(k) for packages requiring competent authority approval at the application procedure in Germany. The paper describes essential items of ageing mechanisms and will give a foresight to the ageing management evaluation by BAM. T2 - INMM & ESARDA Joint Annual Meeting CY - Online meeting DA - 30.08.2021 KW - Dual purpose casks KW - Ageing KW - SSR-6 KW - Interim storage KW - Ageing mechanisms KW - Transport of radaioactive materials PY - 2021 UR - https://www.abstractsonline.com/pp8/#!/10383/presentation/964 SP - 1 EP - 6 PB - Institute of Nuclear Materials Management (INMM) CY - Mount Laurel, NJ AN - OPUS4-53180 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Reichardt, Adrian A1 - Schubert, Sven A1 - Wille, Frank A1 - Komann, Steffen A1 - Neumann, Martin T1 - Introduction of an ageing management approach for packages for the transport of radioactive materials N2 - With integration of the new para 613A into SSR-6 [1] the consideration of ageing mechanisms is now obligatory for the design of packages and their approval. In addition, para 809(f) of SSR-6 [1] requires consideration of the effects of storage on ageing mechanisms, safety analyses and operation and maintenance instructions. German competent authorities Bundesanstalt für Materialforschung und -prüfung (BAM) and Bundesamt für die Sicherheit der nuklearen Entsorgung (BASE) are considering the aspect of ageing in approval procedures. Ageing assessment is mainly focused on dual purpose casks (DPC) package designs which are long-term stored in interim storage facilities. For these package designs, the evaluation of ageing management is now mandatory for the maintenance of the package design approvals with a validity period of 5 years and beyond. The ageing management includes amongst others a gap analysis, the assessment of ageing effects and operational experiences during operation and interim storage. BAM works on the compilation of a guideline for implementation of paras 613A, 809(f) and 809(k) for packages requiring competent authority approval at the application procedure in Germany. The paper describes essential items of ageing mechanisms and will give a foresight to the ageing management evaluation by BAM. T2 - INMM & ESARDA Joint Annual Meeting CY - Online Meeting DA - 30.08.2021 KW - Ageing KW - Dual purpose casks KW - Transport of radaioactive materials KW - Ageing mechanisms KW - Interim storage KW - SSR-6 PY - 2021 AN - OPUS4-53197 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zaghdoudi, Maha A1 - Jaunich, Matthias A1 - Kömmling, Anja A1 - Wolff, Dietmar T1 - How to test the Arrhenius validity of elastomers during ageing N2 - The main task of the division "Safety of Storage Containers" at BAM is the safety assessment of storage containers for radioactive waste. A sufficient sealing of bolted lid systems for the safe containment of the waste for transport, storage and disposal is very important. Extensive knowledge of the change of the elastomer’s properties during ageing and the availability of reliable end-of-lifetime criteria is mandatory to guarantee long term safe enclosure. In a long-term test programme over 5 years we have studied the degradation and the change of sealing properties of several elastomers, including EPDM, at four different ageing temperatures (75 °C, 100 °C, 125 °C and 150 °C). Compression stress relaxation (CSR) and compression set (CS) experiments were carried out. It has been found that when the data does not cover a sufficient time frame necessary for the evolution of the degradation of a chosen property, a curvature in the Arrhenius relationship is observed. For CSR, the curvature was observed for samples aged up to 186 days. As for CS experiments, the curvature was detected for sample aged up to 2 years. To cover the possible lack of experimental long-term data at low temperatures, a numerical model for CSR was developed and longer ageing times for the simulation were adopted. A degradation-rate based model for the evolution of degradative processes is proposed. The main advantage of the model is the possibility to quickly validate the interpolation at lower temperatures within the range of slower chemical processes without forcing an Arrhenius straight-line extrapolation. The model was also applied to CS experiments and validated by the 5 years experimental results where the curvature was gone, and the degradation property has followed an Arrhenius relationship. In the figure the contribution of the two degradative processes are shown over CS and CSR respectively. T2 - 2nd advanced material science world congress CY - Online meeting DA - 14.06.2021 KW - Arrhenius KW - Processes KW - Compression set KW - Compression stress relaxation KW - Modelling PY - 2021 AN - OPUS4-52825 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Reichardt, Adrian A1 - Schubert, Sven A1 - Komann, Steffen A1 - Wille, Frank A1 - Neumann, Martin T1 - Introduction of An Ageing Management Approach for Dual Purpose Casks N2 - With integration of the new para 613A into SSR-6 [1] the consideration of ageing mechanisms is now obligatory for the design of packages and their approval. In addition, para 809(f) of SSR-6 [1] requires consideration of the effects of storage on ageing mechanisms, safety analyses and operation and maintenance instructions. German competent authorities Bundesanstalt für Materialforschung und -prüfung (BAM) and Bundesamt für die Sicherheit der nuklearen Entsorgung (BASE) are considering the aspect of ageing in approval procedures. Ageing assessment is mainly focused on dual purpose casks (DPC) package designs which are long-term stored in interim storage facilities. T2 - BAM-IRSN Symposium 2021 CY - Online meeting DA - 08.09.2021 KW - Ageing KW - DPC KW - Interim storage KW - Ageing management PY - 2021 AN - OPUS4-53287 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias T1 - Polyurethane Foam in Impact Limiters - Experimental and Numerical Analysis N2 - Transport containers with radioactive material are usually shipped in Germany with wooden-filled impact limiters. Alternative energy-absorbing materials for this purpose can be polyurethane foam, for example. In order to perform adequate simulations with PU foam, experimental and numerical investigations must be carried out. With the help of a series of experimental test runs, a material model is developed, tested and compared in LS-Dyna. T2 - Technical Exchange IRSN – BAM: Transport & Storage of Packages for Radioactive Material CY - Berlin, Germany DA - 08.09.2021 KW - Polyurethane Foam, Impact Limiters, PU-Foam PY - 2021 AN - OPUS4-53261 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias T1 - Polyurethane Foam for Impact Limiters - Theory and Experimental Data N2 - Transport containers with radioactive material are usually shipped in Germany with wooden-filled impact limiters. Alternative energy-absorbing materials for this purpose can be polyurethane foam, for example. In order to adequately determine the characteristics for a selected PU foam, various experimental investigations must be carried out. Classically, PU foam specimens are tested under uniaxial compression, confined compression and a tensile test. Using the experimental data and selected material models from the literature, the parameters can be determined to describe and apply material models. T2 - Technical Exchange Sandia – BAM: Transport & Storage of Packages for Radioactive Material CY - Online meeting DA - 31.08.2021 KW - Polyurethane Foam KW - Impact Limiters KW - PU-Foam PY - 2021 AN - OPUS4-53262 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Spent Fuel Characterization - Current Activities in Germany N2 - The characterization of spent fuel and radioactive waste is essential in the evaluation of the ability of these materials to meet acceptance criteria. Prior to determining a method for spent fuel characterization, it is necessary to identify the type of data required to support safety and other analyses for the storage, transportation and final disposal of spent fuel. BAM’s understanding and capabilities in the field of spent fuel characterization are presented for the scope of a coordinated research project on the topic. T2 - IAEA Technical Meeting on Spent Fuel Characterization for Management of Spent Fuel in the Back End of the Fuel Cycle CY - Vienna, Austria DA - 12.11.2019 KW - Nuclear fuel cycle KW - Back end KW - Spent fuel characterization KW - Storage KW - Transportation PY - 2019 AN - OPUS4-55866 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Simbruner, K. T1 - Embrittlement of spent fuel claddings during long-term dry interim storage - Current approach at BAM N2 - The integrity of fuel rods and cladding tubes must be demonstrated throughout the dry interim storage of spent nuclear fuel in Germany. For at least a relevant number of casks, an extension of interim storage between 40 and 100 years is foreseeable. The drying procedure for the fuel rods in conjunction with the thermomechanical storage conditions may change the microstructure of the cladding tubes and lead to their possible embrittlement. The failure probability might increase under mechanical load by handling procedures (transport after storage) on potentially embrittled cladding tubes of fuel rods especially at low temperatures after long-term storage. Results of ring compression tests on unirradiated pre-hydrided as well as irradiated samples with radial hydrides are investigated by finite element analyses to be able to describe the failure process and to identify failure criteria. T2 - EPRI Extended Storage Collaboration Program (ESCP) Winter 2019 Meeting CY - Charlotte, NC, USA DA - 04.11.2019 KW - Fuel rod KW - Cladding KW - Radial hydrides KW - Ring compression test KW - Failure analysis PY - 2019 AN - OPUS4-56005 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Simbruner, Kai T1 - Brittle failure of spent fuel claddings during long-term dry interim storage N2 - In order to demonstrate the safety of both spent fuel and the storage system, a good understanding of the processes that might cause deterioration is required. The International Atomic Energy Agency (IAEA) continued the coordinated research on Spent Fuel Performance Assessment and Research (SPAR) into a fourth phase starting in 2015 to evaluate fuel and materials performance under wet and dry storage and to assess the impact of interim storage on associated spent fuel management activities (such as handling and transport). BAM’s understanding and experiences in the field of long-term storage of spent fuel and especially spent fuel characterization are presented to assist the preparation of the final project report. T2 - IAEA CRP SPAR-IV Third Research Coordination Meeting (RCM3) CY - Buenos Aires, Argentina DA - 07.10.2019 KW - Fuel rod KW - Cladding KW - Radial hydrides KW - Ring compression test KW - Failure analysis PY - 2019 AN - OPUS4-56006 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 - 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 - Gleim, Tobias A1 - Komann, Steffen A1 - Neumann, Martin A1 - Linnemann, Konrad A1 - Wille, Frank T1 - Fracture Mechanical Analyses of a Welding Seam of a Thick-Walled Transport Package N2 - Transport packages shall satisfy various safety criteria regarding mechanical, thermal and radiation phenomena. Typical requirements focusing mechanical aspects are usually drop tests in accordance with IAEA regulations [1]. The drop tests are usually carried out experimentally and, as an additional measure, finite element analyses (FEA) are performed. A specific part of the investigations presented is the evaluation of the welding seam connecting cask shell and cask bottom. Experimental results and FEA are presented and compared. The evaluation of the welding seam performed includes a variety of aspects. In addition to the experimental and analytical stresses determined, different standards are used to investigate a possible crack initiation. Several destructive and non-destructive tests are performed for quality assurance in the manufacturing process as well as for different input parameters. The necessary monitoring and non-destructive measurement methods to define the boundary conditions of the standards are introduced. Taking into account all required parameters, the welding seam is examined and evaluated using the failure assessment diagrams (FAD) of the respective standards. It can be shown under the given boundary conditions that considering the experimental data, the welding seam is in the context of crack initiation below the enveloping curve in the acceptable region. More critical drop tests to be conducted are proposed and need to be investigated in future work. T2 - Pressure Vessels & Piping Conference® 2022 CY - Las Vegas, NV, USA DA - 17.07.2022 KW - Drop test KW - Fracture initiation KW - Transport package PY - 2022 AN - OPUS4-55374 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias A1 - Neumann, Martin A1 - Linnemann, Konrad A1 - Komann, Steffen A1 - Wille, Frank T1 - Fracture Mechanical Analyses of a Welding Seam of a Thick-Walled Transport Package T2 - Proceedings of the ASME 2022 Pressure Vessels & Piping Conference N2 - Transport packages shall satisfy various safety criteria regarding mechanical, thermal and radiation phenomena. Typical requirements focusing mechanical aspects are usually drop tests in accordance with IAEA regulations. The drop tests are usually carried out experimentally and, as an additional measure, finite element analyses (FEA) are performed. A specific part of the investigations presented is the evaluation of the welding seam connecting cask shell and cask bottom. Experimental results and FEA are presented and compared. The evaluation of the welding seam performed includes a variety of aspects. In addition to the experimental and analytical stresses determined, different standards are used to investigate a possible crack initiation. Several destructive and non-destructive tests are performed for quality assurance in the manufacturing process as well as for different input parameters. The necessary monitoring and non-destructive measurement methods to define the boundary conditions of the standards are introduced. Taking into account all required parameters, the welding seam is examined and evaluated using the failure assessment diagrams (FAD) of the respective standards. It can be shown under the given boundary conditions that considering the experimental data, the welding seam is in the context of crack initiation below the enveloping curve in the acceptable region. More critical drop tests to be conducted are proposed and need to be investigated in future work. T2 - Pressure Vessels & Piping Conference® 2022 CY - Las Vegas, NV, USA DA - 17.07.2022 KW - Transport package KW - Drop test KW - Fracture initiation PY - 2022 SP - 1 EP - 9 AN - OPUS4-55375 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - 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 - SMIRT26 - 26th International Conference on Structural Mechanics in Reactor Technology CY - Potsdam, Germany DA - 10.07.2022 KW - Transport packages KW - Bolt connections KW - Finite element analysis KW - Experimental investigation PY - 2022 AN - OPUS4-55426 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Linnemann, Konrad A1 - Komann, Steffen A1 - Wille, Frank A1 - Reiche, I. A1 - Ramsay, J. T1 - New Sco-III regulations to ship large objects as surface contaminated objects T2 - Proceedings der Kerntechnik 2022 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 - Weber, Mike A1 - Zaghdoudi, Maha A1 - Brandt, Guido A1 - Kömmling, Anja A1 - Jaunich, Matthias A1 - Wolff, Dietmar T1 - Coefficients of Friction in Dependence on Aging State of Elastomers – Experimental Identification and Numerical Simulation of the Experiment T2 - Proceedings of the ASME 2022 Pressure Vessels & Piping Conference (PVP2022) N2 - Elastomer seals are mounted as barrier seals in lid systems of containers designed for transport and disposal of negligible heat generating radioactive waste and as auxiliary seals in spent fuel storage and transportation casks (dual purpose casks (DPC)). When the behavior of mounted seals under normal and hypothetical accident conditions of disposal and transport is to be simulated, a comprehensive knowledge of their complex mechanical properties at every state of aging is necessary. In previous works, BAM’s efforts in experimental investigations on specimen artificially aged at different temperatures and times and the implementation of the found results in finite element material models were presented. Additionally, our approaches to reproduce the aging process itself and to extrapolate the results of artificially accelerated aging to longer times were presented. Numerical simulations have shown that the behavior of the seal during mounting and one-sided pressurizing and the resulting performance values such as leakage rate strongly depend on the coefficient of friction (COF) between flange and seal. The friction coefficient, in turn, depends on the aging state of the elastomer material as several publications suggest (see below). Dynamic COF between an exemplary ethylene propylene diene rubber (EPDM) material and a stainless steel ball were determined by using a self-designed linear oscillation tribometer. Unaged and artificially aged EPDM specimen stored for 30 days and 100 days at a temperature of 150 °C were tested. A stainless steel ball (d=10 mm) is brought in contact with the specimen’s surface und loaded by normal forces of 2.5 N, 5 N, 10 N and 20 N. During a reciprocating movement of the EPDM sheet, the horizontal force/friction force is continuously measured, and the COF can be derived. It is well known that friction is a complex phenomenon especially in soft materials. It cannot be excluded that the measured friction force is influenced by additional force components, resulting from the ball’s grooving through the elastomer’s surface. This force depends on the penetration depth of the ball and on the resistance of the elastomer in its different states of aging. The latter results from microstructural changes i.e., chain scission and additional crosslinking that occur during aging which in turn influence the softening or hardening of the material. A finite element (FE) ABAQUS® model was developed to reproduce the measurement process. It should help to better understand the physical mechanisms and to quantify the percentage of measured forces resulting from real friction on the one hand and forces resulting from unintended side effects that could falsify the result on the other hand. The behavior of the elastomer in its different states of aging is reproduced by a FE material model already presented in previous works of BAM. T2 - ASME 2022 Pressure Vessels & Piping Conference (PVP2022) CY - Las Vegas, NV, USA DA - 17.02.2022 KW - Numerical Simulation KW - Radioactive Waste KW - Elastomers KW - Aging KW - Seal Behavior KW - Leakage Rate KW - Coefficient of Friction KW - Experiment PY - 2022 SP - 1 EP - 8 PB - American Society of Mechanical Engineers (ASME) CY - New York, NY, USA AN - OPUS4-57093 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Zaghdoudi, Maha A1 - Brandt, Guido A1 - Kömmling, Anja A1 - Jaunich, Matthias A1 - Wolff, Dietmar T1 - Coefficients of Friction in Dependence on Aging State of Elastomers – Experimental Identification and Numerical Simulation of the Experiment N2 - Elastomer seals are mounted as barrier seals in lid systems of containers designed for transport and disposal of negligible heat generating radioactive waste and as auxiliary seals in spent fuel storage and transportation casks (dual purpose casks (DPC)). When the behavior of mounted seals under normal and hypothetical accident conditions of disposal and transport is to be simulated, a comprehensive knowledge of their complex mechanical properties at every state of aging is necessary. In previous works, BAM’s efforts in experimental investigations on specimen artificially aged at different temperatures and times and the implementation of the found results in finite element material models were presented. Additionally, our approaches to reproduce the aging process itself and to extrapolate the results of artificially accelerated aging to longer times were presented. Numerical simulations have shown that the behavior of the seal during mounting and one-sided pressurizing and the resulting performance values such as leakage rate strongly depend on the coefficient of friction (COF) between flange and seal. The friction coefficient, in turn, depends on the aging state of the elastomer material as several publications suggest (see below). Dynamic COF between an exemplary ethylene propylene diene rubber (EPDM) material and a stainless steel ball were determined by using a self-designed linear oscillation tribometer. Unaged and artificially aged EPDM specimen stored for 30 days and 100 days at a temperature of 150 °C were tested. A stainless steel ball (d=10 mm) is brought in contact with the specimen’s surface und loaded by normal forces of 2.5 N, 5 N, 10 N and 20 N. During a reciprocating movement of the EPDM sheet, the horizontal force/friction force is continuously measured, and the COF can be derived. It is well known that friction is a complex phenomenon especially in soft materials. It cannot be excluded that the measured friction force is influenced by additional force components, resulting from the ball’s grooving through the elastomer’s surface. This force depends on the penetration depth of the ball and on the resistance of the elastomer in its different states of aging. The latter results from microstructural changes i.e., chain scission and additional crosslinking that occur during aging which in turn influence the softening or hardening of the material. A finite element (FE) ABAQUS® model was developed to reproduce the measurement process. It should help to better understand the physical mechanisms and to quantify the percentage of measured forces resulting from real friction on the one hand and forces resulting from unintended side effects that could falsify the result on the other hand. The behavior of the elastomer in its different states of aging is reproduced by a FE material model already presented in previous works of BAM. T2 - ASME 2022 Pressure Vessels & Piping Conference (PVP2022) CY - Las Vegas, NV, USA DA - 17.07.2022 KW - Radioactive Waste KW - Elastomers KW - Aging KW - Experiment KW - Seal Behavior KW - Numerical Simulation KW - Leakage Rate KW - Coefficient of Friction PY - 2022 AN - OPUS4-57095 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Gröke, Carsten A1 - Wille, Frank T1 - Assessment of quality management for transport packages not requiring authority design approval T2 - Waste Management Conference 2018 N2 - The majority of transports of radioactive materials are carried out in packages which don’t need a package design approval of a competent authority. Low active radioactive materials are transported in such kind of packages e.g. in the medical and pharmaceutical industry and in the nuclear industry as well. In Germany the decision to phase out nuclear energy leads to a strong demand for packages to transport low and middle active radioactive waste due to the dismantling and decommissioning of nuclear power plants. According to IAEA regulations the “non-competent authority approved package types” are the excepted packages and the industrial packages of Type IP-1, IP-2 and IP-3 and of Type A. For the packages of Type IP-2, IP-3 and Type A an assessment by the German competent authority is required for the quality management for the design, manufacture, testing, documentation, use, maintenance and inspection. In general a compliance audit of the manufacturer of the packaging is required during this assessment procedure. T2 - Waste Management Conference 2018 CY - Phoenix, USA DA - 18.3.2018 KW - Radioactive material KW - Quality management system KW - Transport packages PY - 2018 SP - Article 18194, 1 EP - 7 AN - OPUS4-44170 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Nehrig, Marko A1 - Erenberg, Marina A1 - Feldkamp, Martin A1 - Bletzer, Claus A1 - Musolff, André A1 - Wille, Frank T1 - Large Scale IAEA Thermal Test with Wood filled Impact Limiters N2 - Packages for the transport of radioactive material are often equipped with impact limiters consisting of wood. Mostly this wood is encapsulated by steel sheets. The impact limiters are needed to ensure that the transport casks meet the IAEA safety requirements. According to the IAEA safety requirements a package has to withstand consecutively severe mechanical tests followed by a thermal test. The mechanical tests have to produce maximum damage concerning the thermal test. Following this, the impact limiters may have serious pre-damage when the thermal tests begins. The IAEA safety requirements state that during and following the fire test, the specimen shall not be artificially cooled and any combustion of materials of the package shall be permitted to proceed naturally. Small scale fire tests with wood filled metal drums by BAM and works of the French Institute for Radiological Protection and Nuclear Safety (IRSN) showed that pre-damaged steel encapsulated wooden structures could start smoldering initiated by the thermal test. These processes supply additional energy to the cask which should be considered within the safety assessment of the package. As not much is known about smoldering processes in encapsulated wooden structures with a reduced oxygen supply the need for a test was identified. To investigate the influence of a smoldering impact limiter concerning the amount of energy supplied to the cask in dependence of the time BAM conducted a large scale impact limiter thermal test. For that, a pre-damaged impact limiter with a diameter of 2,3 m was mounted on a water tank simulating a cask. A complex system of a regulated pump, a heater, a cooler, a slide valve, a flow meter and numerous thermocouples were installed and connected to a control unit to ensure all needed operating conditions. After a pre-heating compared to typical SNF decay-heat, the 30 min lasting fire phase of the thermal test was started. After that, the expected and initiated smoldering began. The results of the large scale test are presented in this poster. Systematic small scale tests will follow to identify the influence of different parameters, e.g. moisture content and scale effects. The tests took place at BAM Test Site for Technical Safety (TTS) with its various possibilities for mechanical and thermal tests. The results of these tests will have direct influence in the safety assessment of transport cask for the transport of radioactive material. T2 - SNL BAM Workshop CY - Albuquerque, NM, USA DA - 14.03.2018 KW - Type-B Package KW - Thermal Test PY - 2018 AN - OPUS4-45313 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CHAP A1 - Nehrig, Marko A1 - Wille, Frank ED - Saegusa, Toshiari ED - Sert, Gilles ED - Völzke, Holger ED - Wille, Frank T1 - Internal Pressure build-up of Waste Packages with Wet Contents Under Fire Impact T2 - Basic of Transport and Storage of Radioactive Materials N2 - Intermediate level waste often contains residual water or other liquids. The liquids can be decomposed due to radiolysis and can also vaporize especially under fire impact. These pressure generating processes are described. One example for the pressure determination inside a closed cask is discussed using two different approaches: analytical and CFD KW - Wet Content KW - Pressure build-up KW - Type-B Package PY - 2018 SN - 978-981-3234-03-1 SP - 195 EP - 209 PB - World Scientific Publishing CY - Singapore ET - Erste AN - OPUS4-45301 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Nehrig, Marko A1 - Erenberg, Marina A1 - Feldkamp, Martin A1 - Bletzer, Claus A1 - Musolff, André A1 - Wille, Frank T1 - Large Scale IAEA Thermal Test with Wood filled Impact Limiters N2 - Packages for the transport of radioactive material are often equipped with impact limiters consisting of wood. Mostly this wood is encapsulated by steel sheets. The impact limiters are needed to ensure that the transport casks meet the IAEA safety requirements. According to the IAEA safety requirements a package has to withstand consecutively severe mechanical tests followed by a thermal test. The mechanical tests have to produce maximum damage concerning the thermal test. Following this, the impact limiters may have serious pre-damage when the thermal tests begins. The IAEA safety requirements state that during and following the fire test, the specimen shall not be artificially cooled and any combustion of materials of the package shall be permitted to proceed naturally. Small scale fire tests with wood filled metal drums by BAM and works of the French Institute for Radiological Protection and Nuclear Safety (IRSN) showed that pre-damaged steel encapsulated wooden structures could start smoldering initiated by the thermal test. These processes supply additional energy to the cask which should be considered within the safety assessment of the package. As not much is known about smoldering processes in encapsulated wooden structures with a reduced oxygen supply the need for a test was identified. To investigate the influence of a smoldering impact limiter concerning the amount of energy supplied to the cask in dependence of the time BAM conducted a large scale impact limiter thermal test. For that, a pre-damaged impact limiter with a diameter of 2,3 m was mounted on a water tank simulating a cask. A complex system of a regulated pump, a heater, a cooler, a slide valve, a flow meter and numerous thermocouples were installed and connected to a control unit to ensure all needed operating conditions. After a pre-heating compared to typical SNF decay-heat, the 30 min lasting fire phase of the thermal test was started. After that, the expected and initiated smoldering began. The results of the large scale test are presented in this poster. Systematic small scale tests will follow to identify the influence of different parameters, e.g. moisture content and scale effects. The tests took place at BAM Test Site for Technical Safety (TTS) with its various possibilities for mechanical and thermal tests. The results of these tests will have direct influence in the safety assessment of transport cask for the transport of radioactive material. T2 - 2018 WM Symposia CY - Phoenix, AZ, USA DA - 18.03.2018 KW - Type-B Package KW - Accident Conditions of Transport KW - Thermal Test PY - 2018 AN - OPUS4-45310 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Nehrig, Marko A1 - Erenberg, Marina A1 - Feldkamp, Martin A1 - Bletzer, Claus A1 - Musolff, André A1 - Wille, Frank T1 - Large Scale IAEA Thermal Test with Wood filled Impact Limiters T2 - WM 2018 Symposia (Proceedings) N2 - Packages for the transport of radioactive material are often equipped with impact limiters consisting of wood. Mostly this wood is encapsulated by steel sheets. The impact limiters are needed to ensure that the transport casks meet the IAEA safety requirements. According to the IAEA safety requirements a package has to withstand consecutively severe mechanical tests followed by a thermal test. The mechanical tests have to produce maximum damage concerning the thermal test. Following this, the impact limiters may have serious pre-damage when the thermal tests begins. The IAEA safety requirements state that during and following the fire test, the specimen shall not be artificially cooled and any combustion of materials of the package shall be permitted to proceed naturally. Small scale fire tests with wood filled metal drums by BAM and works of the French Institute for Radiological Protection and Nuclear Safety (IRSN) showed that pre-damaged steel encapsulated wooden structures could start smoldering initiated by the thermal test. These processes supply additional energy to the cask which should be considered within the safety assessment of the package. As not much is known about smoldering processes in encapsulated wooden structures with a reduced oxygen supply the need for a test was identified. To investigate the influence of a smoldering impact limiter concerning the amount of energy supplied to the cask in dependence of the time BAM conducted a large scale impact limiter thermal test. For that, a pre-damaged impact limiter with a diameter of 2,3 m was mounted on a water tank simulating a cask. A complex system of a regulated pump, a heater, a cooler, a slide valve, a flow meter and numerous thermocouples were installed and connected to a control unit to ensure all needed operating conditions. After a pre-heating compared to typical SNF decay-heat, the 30 min lasting fire phase of the thermal test was started. After that, the expected and initiated smoldering began. The results of the large scale test are presented in this poster. Systematic small scale tests will follow to identify the influence of different parameters, e.g. moisture content and scale effects. The tests took place at BAM Test Site for Technical Safety (TTS) with its various possibilities for mechanical and thermal tests. The results of these tests will have direct influence in the safety assessment of transport cask for the transport of radioactive material T2 - 2018 WM Symposia CY - Phoenix, AZ, USA DA - 18.03.2018 KW - Type-B Package KW - Thermal Test PY - 2018 SP - Paper 18257, 1 EP - 11 AN - OPUS4-45311 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Bletzer, Claus T1 - CONFIRE Temperture development during fire test of a final storagr container Typ V N2 - This presentation focusses on the thermal test of a typical storage container design for the German final storage repository KONRAD. The preparation of the container, the conduction of the fire test itself and results like the temperature evolution and as-sumptions for thermal material properties are shown. As thermal test final storage containers have to be exposed to a fire of 800 °C (1472 °F) for 1 hour according to the final storage conditions provided by the federal com-pany for radioactive waste disposal (BGE). The here discussed cask contains a steel covered concrete layer of a thickness of 100/150 mm. Heating up concrete is a highly nonlinear process since the vaporization of residual water consumes high amounts of thermal energy and produce vast amounts of steam. The constantly measured temperatures provide the basis to adjust common models of thermal properties for concrete esp. for FEAs. Finally in a later step this will allow predictions of similar designed container with different sizes. T2 - SNL BAM Workshop CY - Albuquerque, NM, USA DA - 14.03.2018 KW - KONRAD KW - Fire testing KW - Concrete KW - Final storage KW - Thermal properties PY - 2018 AN - OPUS4-45427 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Scheidemann, Robert A1 - Weber, Mike A1 - Quercetti, Thomas T1 - Approach for determining the length of the bar in compliance with the IAEA 1m puncture test N2 - The paper focuses on the preliminary determination of the bar length and the experimental performance of the 1m-puncture test in compliance with the guidelines of the IAEA under the given boundary conditions. Following aspects have to be considered concerning the determination of an appropriate length of the bar to obtain maximum damage to the specimen: the design of the package, its drop orientation and the impact point as well as a pre-damage of the package resulting from a previous 9m drop test. According to the Regulations the minimum length of the bar is defined as 20 cm, which has to be adapted to the outer surface of the package, especially to packages with large impact limiters. In this context, two main aspects are important: First, the realisation of a maximum puncture load applied to the package while no other component of the package has contact with the impact target. Second, the reduction of the risk of buckling and maximising the stiffness of the bar by a length which is as short as possible. In order to optimize the bar length, finite element calculations are often done by simulating the global behaviour of the package during the puncture test. The evaluation of the conducted puncture test regarding IAEA compliance is done by analysing deceleration measurements. The paper presents a possible approach to determine the length of the bar and the construction of a form-fitted connection to the target showing various examples of puncture tests. T2 - RAMTRANSPORT 2018 CY - London, UK DA - 16.05.2018 KW - Puncture test KW - Bar length KW - Numerical simulation PY - 2018 AN - OPUS4-45027 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Jaunich, Matthias T1 - The average number of references per article published in Polymer Testing over the years JF - Polymer Testing N2 - This Editorial is focussed on the changes in the number of references per article published in Polymer Testing from 1980 until today. To set in into context with other changes a comparison with the impact factor of Polymer Testing is presented. KW - Reference KW - Editorial KW - Impact factor KW - Changes with time PY - 2018 DO - https://doi.org/10.1016/j.polymertesting.2018.05.034 SN - 0142-9418 VL - 68 SP - A1 EP - A2 PB - Elsevier AN - OPUS4-45212 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias T1 - Investigation of the low-temperature performance of rubber seals N2 - For many sealing applications rubbers are applied due to their special material properties and easy use. In many applications as e.g. in aviation, traffic or process technology the material can be exposed to low temperatures during operation. Therefore, it is important to know the material behaviour at those low temperatures to determine the temperature range that allows a safe operation of the seal. In this work, we focus on the behaviour of elastomer seals at low temperatures with regard to potential decrease of leak-tightness or catastrophic seal failure of O-ring rubber seals. This is required as material properties of rubbers are strongly temperature dependent but their temperature application range is not always clearly defined. Based on previous investigations which considered the physical material properties and the seal behaviour under purely static conditions we widen the focus on the sealing performance after a fast partial relief of compressed seals and additional materials. For the investigations, different typical rubber seal materials were used as e.g. fluorocarbon (FKM), ethylene propylene diene (EPDM) and hydrogenated acrylonitrile-butadiene (HNBR) rubber. The correlation of the physical material properties with seal performance is studied. This includes as well purely static sealing applications as the performance during/after fast partial relief of seals at low temperatures. T2 - Polymertec 2018 CY - Merseburg, Germany DA - 13.06.2018 KW - Seal KW - Low temperature KW - Leakage PY - 2018 AN - OPUS4-45214 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Müller, Lars A1 - Schönfelder, Thorsten A1 - Komann, Steffen A1 - Ballheimer, Viktor A1 - Wille, Frank ED - Müller, Lars T1 - Assessment experience on packages loaded with damaged spent nuclear fuel for transport after storage T2 - 11th International Conference on the Transport, Storage and Disposal of Radioactive Materials N2 - In 2017 the first German package approval certificate was issued for a dual purpose cask (DPC) design with encapsulated damaged spent nuclear fuel. At the Bundesanstalt für Materialforschung und -prüfung (BAM) a comprehensive assessment procedure was carried out with respect to the mechanical and thermal design, the containment design and quality assurance for manufacturing and operation. Main objective of this procedure was to verify the Package Design Safety Report (PDSR) fulfils the requirements according to the IAEA regulations SSR-6. Until now only standard spent nuclear fuel assemblies were designated for interim storage and transports. Due to nuclear phase out in Germany all other kinds of SNF in particular damaged fuel has to be packed. Therefore specific requirements have to be considered in accordance with international experiences written in IAEA technical reports. In Germany damaged spent nuclear fuel (DSNF) needs a tight encapsulation with special encapsulations and clearly defined properties. Due to the limited amount of DSNF these encapsulations are designed for storage and transport in existing packages. From the assessment experience it has been seen, corresponding PDSR need an extensively expansion to cover the design of these encapsulations and their influences on the package. Then such well-defined encapsulations can be handled like standard fuel assemblies. The main difference to standard package components is, encapsulations with permanent closure achieve their specified condition not after manufacturing but only during operation after loading and closing. Thus specific handling instruction and test procedures are necessary especially for welding, where BAM is able to survey the quality of this first part of operation. T2 - 11th International Conference on the Transport, Storage and Disposal of Radioactive Materials CY - London, UK DA - 15.05.2018 KW - Assessment KW - Dual purpose cask KW - Spent nuclear fuel PY - 2018 SP - Paper 18524, 1 EP - 8 AN - OPUS4-45258 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Moutarde, Marianne A1 - Apel, Andreas A1 - Ballheimer, Viktor A1 - Sterthaus, Jens T1 - Recent developments in standards and IAEA guidance material for package load attachment points N2 - For transport package design and operation according to IAEA regulations, the package shall be securely stowed and its retention system shall be capable to withstand load conditions of routine transport. The supporting IAEA Advisory Material SSG-26 provides information how to do that. Up to now package designers in different countries use other load factors for the design of attachment points than those specified in the IAEA guidance material. In particular the acceleration values vary between different countries and lead to difficulties during the validation of foreign approval certificates. Therefore the IAEA started a discussion process to review the existing guidance text. An international working group was constituted in 2013. Representatives came from different stakeholders, e.g. transport operators, competent authorities and modal organizations. The discussions concluded especially on the transport conditions which has to be considered for stowage design, including on the one hand the relevance of the load factors used for strength and fatigue analysis and on the other hand the criteria which have to be considered for the attachment points. Another standard with relevance to the load attachment of packages is ISO 10276. This standard deals with trunnion design, manufacturing and operational aspects. The regular standard revision phase started in 2017. An expert group discussed new state-of-the-art technology and proposed revised text for the ISO standard for international discussion. The paper describes relevant tie-down aspects, gives background argumentation, and tries to support harmonized application of the revised IAEA guidance material and future ISO 10276 standard. T2 - 11th International Conference on the Transport, Storage and Disposal of Radioactive Materials CY - London, UK DA - 15.05.2018 KW - Acceleration KW - Package design KW - ISO KW - Ladungssicherung PY - 2018 AN - OPUS4-45293 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Moutarde, M. A1 - Apel, Andreas A1 - Ballheimer, Viktor A1 - Sterthaus, Jens T1 - Recent developments in standards and IAEA guidance material for package load attachment points T2 - Proceedings of the 11th International Conference on the Transport, Storage and Disposal of Radioactive Materials N2 - For transport package design and operation according to IAEA regulations, the package shall be securely stowed and its retention system shall be capable to withstand load conditions of routine transport. The supporting IAEA Advisory Material SSG-26 provides information how to do that. Up to now package designers in different countries use other load factors for the design of attachment points than those specified in the IAEA guidance material. In particular the acceleration values vary between different countries and lead to difficulties during the validation of foreign approval certificates. Therefore the IAEA started a discussion process to review the existing guidance text. An international working group was constituted in 2013. Representatives came from different stakeholders, e.g. transport operators, competent authorities and modal organizations. The discussions concluded especially on the transport conditions which has to be considered for stowage design, including on the one hand the relevance of the load factors used for strength and fatigue analysis and on the other hand the criteria which have to be considered for the attachment points. Another standard with relevance to the load attachment of packages is ISO 10276. This standard deals with trunnion design, manufacturing and operational aspects. The regular standard revision phase started in 2017. An expert group discussed new state-of-the-art technology and proposed revised text for the ISO standard for international discussion. The paper describes relevant tie-down aspects, gives background argumentation, and tries to support harmonized application of the revised IAEA guidance material and future ISO 10276 standard. T2 - 11th International Conference on the Transport, Storage and Disposal of Radioactive Materials CY - London, UK DA - 15.05.2018 KW - Verzögerungswerte KW - Lastanschlag KW - Trunnion design KW - ISO KW - Retention PY - 2018 SP - 18511, 1 EP - 8 AN - OPUS4-45294 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Debruyne, M. A1 - Eckert, B. A1 - Wille, Frank A1 - Gauthier, F. A1 - Le Bars, I. A1 - Cordier, N. A1 - Jouve, A.-C. T1 - Assessment of safety demonstrations relative to packages containing UF6 T2 - Proceedings of 11th International Conference on the Transport, Storage and Disposal of Radioactive Materials N2 - Specific attention should be paid on safety demonstrations transmitted by applicants in the case of approval request for the package designs containing enriched UF6. Concerning the shipment of enriched UF6, the package designs consist in general of a filled 30B cylinder surrounded by an overpack. The description of the content, considering the UF6 origin, i.e. natural or reprocessed, shall be clearly justified especially when the UF6 isotopic composition exceeds the limits specified in ASTM standards. Concerning the containment of the UF6, the applicant shall demonstrate in all conditions of Transport the leak-tightness of the valve and plug of the cylinders filled with enriched UF6. In this regard, when justifications are based on numerical calculations, the absence of contact between These components of the cylinder and the internal surfaces of the overpack after the regulatory drop tests shall be shown. In particular, absence of contact between the valve and any other component of the packaging shall be confirmed to respect the current IAEA regulations [3]. If complementary calculations show a contact between the plug and the internal surfaces of the overpack, additional tests are required to confirm that the strength resulting from this contact will not affect the plug leak-tightness. It can be noticed that the future revision of the IAEA regulations will include additional provision in case of contact of the plug with any other component of the packaging. In addition, the applicant shall demonstrate that the melting temperature of the valve, including the tinned joint, will not be exceeded during the regulatory fire test. Furthermore, the representativeness of the ballast used to simulate the behaviour of the UF6 loaded within the cylinder shall be justified if drop tests are performed. Finally, specific provisions relative to the use of plugs and the maintenance of cylinders should be included in the safety analysis reports. T2 - 11th International Conference on the Transport, Storage and Disposal of Radioactive Materials CY - London, UK DA - 15.05.2018 KW - Uranhexaflourid KW - Package safety KW - Valve KW - Drop testing KW - Radioactive material PY - 2018 SP - 18523, 1 EP - 8 AN - OPUS4-45295 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -