TY - CONF A1 - Völzke, Holger T1 - Update on the German Spent Fuel Management Strategy N2 - The presentation provides an update on the German spent fuel Management strategy in the areas of transportation, dry interim storage and deep geological disposal. T2 - IAEA 18th TWG-NFCO CY - Online meeting DA - 07.09.2020 KW - Spent nuclear fuel KW - Final disposal KW - Safety KW - Transport and storage cask PY - 2020 AN - OPUS4-51616 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Simbruner, Kai T1 - Brittle failure analysis of spent fuel zirconium alloy claddings during long-term dry interim storage - Current approach at BAM N2 - Understanding the mechanical behaviour of fuel cladding is decisive for an extended dry interim storage period beyond the currently licensed period of 40 years in Germany and subsequent handling, transport, and disposal considering normal and accidental conditions. Since fuel cladding acts as the first barrier for spent fuel, its integrity should be demonstrated to enable safe and secure discharge and handling options after interim storage and subsequent transportation in preparation of final disposal. An established method for characterizing cladding material is the Ring Compression Test (RCT), where a small, cylindrical sample of the cladding tube is subjected to a compressive load. This test is a laboratory representation of a fuel rod load case and has shown a high susceptibility to failure under the occurrence of radial hydrides. Certain hydride morphologies and low temperature might even lead to brittle fracture at very small loads. The BAM research project BRUZL (Fracture mechanical analysis of spent fuel claddings during long-term dry interim storage) aims to study the results of quasi-static RCTs and to establish numerical models to gain an in-depth understanding of the stress state during the experiments. Using a fracture mechanical approach, cases of sudden failure during the RCT procedure are used to characterize the material behaviour and establish a failure criterion. As an integral part of the project BRUZL, quasi-static Ring Compression Tests have been performed to identify all experimental details, which might be helpful for numerical modelling. Unirradiated samples of the cladding material ZIRLO® have been subjected to hydrogen charging and a thermo-mechanical treatment for radial hydride reorientation. Sample preparation, testing procedures, and analysis results are presented. A numerical model has been established and an elastic-plastic material model was derived from as-received RCT samples by inverse finite element analyses. Cohesive zone modelling has been implemented to reproduce sudden load drops during RCT. The project is funded by the Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561. T2 - Kolloquium der Abteilung 3 CY - Online meeting DA - 05.11.2020 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2020 AN - OPUS4-51987 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger T1 - Anforderungen an Behälter für die Endlagerung hochradioaktiver Abfälle in Steinsalz, Tonstein und Kristallingestein N2 - Bei der Auswahl eines Standortes für ein Endlager für hochradioaktive Abfälle (Wärme entwickelnde radioaktive Abfälle und ausgediente Brennelemente) in Deutschland werden auf Grundlage des Standortauswahlgesetzes von 2017 die drei potenziellen Wirtsgesteine Steinsalz, Tonstein und Kristallingestein in Betracht gezogen. Die Auswahl eines konkreten Endlagerstandortes steht dabei in engem Zusammenhang mit den für die jeweiligen Wirtsgesteine geeignet erscheinender Endlagersystemen. Zentraler Bestandteil dieser Endlagersysteme sind die Endlagerbehälter, deren Gestaltung und Auslegung maßgeblich die Randbedingungen für die Transport- und Einlagerungstechnik vorgibt und zudem die Nachweise zur Betriebs- und Langzeitsicherheit entscheidend prägt. Daher ist es für die Behälterentwicklung erforderlich, die an Endlagerbehälter zu stellenden Anforderungen systematisch herzuleiten. Auf dieser Grundlage können anforderungsgerechte Endlagerbehälter zielgerichtet für die jeweils zu untersuchenden Endlagerstandorte in unterschiedlichen Wirtsgesteinen entwickelt werden. Dieser Beitrag fasst die wesentlichen Ergebnisse des von BAM und BGE TECHNOLOGY GmbH im Zeitraum 2017 bis 2020 bearbeiteten und vom Bundesministerium für Wirtschaft und Energie (BMWi) geförderten Forschungsvorhabens KoBrA zusammen, die in einem Abschlussbericht und vier zugehörigen Teilberichten umfassend dokumentiert sind. T2 - Tage der Standortauswahl CY - Freiberg, Germany DA - 11.02.2021 KW - Endlager KW - Behälter KW - Sicherheitsanforderungen KW - hochradioaktive Abfälle PY - 2021 AN - OPUS4-53115 LA - deu 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 - Weber, Mike A1 - Völzke, Holger A1 - Zencker, Uwe A1 - Hauer, Katharina A1 - Herz, A. A1 - Kreienmeyer, M. T1 - Aufbau eines Aufprallfundamentes mit reproduzierbaren Eigenschaften für Baumusterprüfungen von Abfallbehältern für das Endlager Konrad N2 - Abgeleitet aus den in Teil 3 der „Systemanalyse Konrad“ [1] definierten Lastannahmen für den „untertägigen Anlagenbereich“ müssen Verpackungen, die der Abfallbehälterklasse I (ABK I) mit dem zusätzlichen Merkmal der störfallfesten (sf) Verpackung oder der ABK II zugeordnet werden, gemäß den „Endlagerungsbedingungen, Endlager Konrad“ [2], nach einem „…Fall aus 5 m Höhe auf eine unnachgiebige Unterlage…“ bestimmten Anforderungen genügen. Eine nähere Beschreibung der Prüfeinrichtung sowie die Definition detaillierter Anforderungen an das Aufprallfundament erfolgen in der „Produktkontrolle, Endlager Konrad“ [3]: „Die Prüfeinrichtung, bestehend aus einer Hebeeinrichtung und einem Fundament ist so zu gestalten, dass der Fall des Prüfmusters in jeder beliebigen Aufprallorientierung aus 5 m Höhe, gemessen von der Unterkante des Prüfmusters zum Fundament, möglich ist. Das Fundament ist aus Beton der Mindestgüte B 35 herzustellen und muss so beschaffen und verankert sein, dass keine Zerstörung in zahlreiche kleinere Bruchstücke erfolgt und keine horizontale Verschiebung des Fundamentes oder seiner Teile auftritt. Maße und Masse des Fundaments müssen so beschaffen sein, dass der Einfluss des Baugrundes unter dem Fundament auf den Aufprallwiderstand vernachlässigbar ist.“ Im Rahmen der Bauartprüfung für Behälter zur Endlagerung radioaktiver Abfälle mit zu vernachlässigender Wärmeentwicklung im Endlager Konrad wird die Bundesanstalt für Materialforschung und -prüfung (BAM) von der Bundesgesellschaft für Endlagerung (BGE) als Sachverständige hinzugezogen und ist in der Regel auch mit der Durchführung der notwendigen Fallprüfungen auf dem Testgelände Technische Sicherheit der BAM (BAM TTS) einschließlich der Bereitstellung der benötigten Prüfeinrichtungen sowie des regelkonformen Aufprallfundamentes beauftragt. Der Prozess, beginnend von der Gewährleistung einer von allen Beteiligten als gültig bewerteten Fallprüfung bis hin zur abschließenden Bewertung und Interpretation der Versuchsergebnisse erfordert dabei u.a. eine sorgfältige Planung und Fertigung der für das Endlager Konrad repräsentativen Fundamentplatte und deren Ankopplung an einen Untergrund im Sinne der „Regulations for the Safe Transport of Radioactive Material“ [4], [5] wie er zum Beispiel auf den beiden Fallversuchsanlagen des BAM Testgelände Technische Sicherheit (BAM TTS) zur Verfügung steht. Um diese Anforderungen zu erfüllen muss zum einen der Erhalt der Integrität der Fundamentplatte beim Fallversuch durch eine geeignete Auswahl und kraftflussgerechte Anordnung der Stabstahl-Bewehrung sowie deren Verankerung garantiert und damit eine unzulässige Energieaufnahme durch Risse sowie eine unzulässige „… Zerstörung (der Fundamentplatte) in zahlreiche kleinere Bruchstücke …“, wie in [3] gefordert, verhindert werden. Außerdem ist zu gewährleisten, dass der Beton die Anforderungen an die vorgeschriebene Mindestgüte bzw. -festigkeit zum Zeitpunkt der Fallprüfung erfüllt, die Betonfestigkeit jedoch zur Vermeidung unnötig verschärfter Prüfrandbedingungen eine zu definierende Obergrenze nicht überschreitet. Ausgehend von dem 2009 im Rahmen der KONTEC veröffentlichten Wissensstand [6] präsentiert die BAM im vorliegenden Beitrag die wesentlichen Ergebnisse der zwischenzeitlich bei der Planung und Durchführung von zahlreichen Fallprüfungen gewonnenen Erfahrungen bzgl. Spezifikation und Fertigung der für die Fallversuche notwendigen Fundamentplatten sowie deren Anbindung an den Untergrund. Neben den immer wieder notwendigen Anstrengungen zur Gewährleistung einer im Rahmen der Definition korrekten Betondruckfestigkeit am Tag der Fallprüfung wird die Anpassung der Bewehrungsführung an verschiedene Fallpositionen diskutiert. Zur Ergänzung des gültigen Regelwerks entsteht parallel zu diesem Dokument eine Fachnotiz, in der BGE und BAM die Vorgaben und Spezifikationen zur regelwerkskonformen Herstellung und Montage von Fundamentplatten für Fallprüfungen im Rahmen der Behälterbauartprüfungsverfahren für das Endlager Konrad zusammenfassen und veröffentlichen werden. Diese Fachnotiz soll detaillierte Vorgaben und Spezifikationen zur Planung, Beauftragung, Herstellung und abschließenden Prüfung eines solchen Fundamentes und dessen Anbindung an den Untergrund enthalten. Damit soll allen beteiligten Organisationen wie Antragstellern, Sachverständigen und BGE ein langfristig verlässlicher Leitfaden im Hinblick auf die Durchführung anforderungsgerechter Fallprüfungen unter definierten und reproduzierbaren Randbedingungen im Rahmen der Bauartprüfungsverfahren für das Endlager Konrad an die Hand gegeben werden. T2 - KONTEC 2021 CY - Dresden, Germany DA - 25.08.2021 KW - Endlager Konrad KW - Fallprüfung KW - Aufprallfundament PY - 2021 AN - OPUS4-53374 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Simbruner, Kai T1 - Embrittlement of Spent Fuel Claddings – Fracture Mechanical Evaluation and Failure Criteria N2 - The BAM research project BRUZL (Fracture mechanical analysis of spent fuel claddings during long-term dry interim storage) evaluates results of quasi-static Ring Compression Tests (RCT). Due to its simple realisation, the RCT is an established method for characterising the mechanical behaviour of fuel cladding, where a small tube sample is subjected to a compressive load and the relation between load and displacement is obtained. During tests at room temperature, unirradiated ZIRLO® samples with a distinct extend of radial hydrides have shown high susceptibility to sudden, brittle failure at relatively small loads. Hydride morphology and radial hydride metrics have been an integral part of the characterisation of degradation mechanisms of cladding. A software tool for the analysation of metallographic micrographs was developed. The program allows the automated calculation of hydride lengths, detection of continuous hydride structures, their statistical evaluation and visualisation. Numerical models have been established to reproduce experimental RCT results featuring an elastic-plastic material model, which was derived from inverse finite element analyses of as-received RCT samples. A cohesive zone model (CZM) is introduced to the model to reproduce load drops in the linear regime of the load vs. displacement curve. Using a traction-separation approach for the CZM, the cohesive parameters are fitted to match numerical and experimental results yielding the fracture mechanical material properties. Combining the three key aspects hydride morphology, fracture mechanics and RCT stress-state a failure criterion can be established as a result of the BRUZL project. The project is funded by the Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561. T2 - 5th GRS Workshop on Safety of Extended Dry Storage of Spent Nuclear Fuel CY - Online meeting DA - 09.06.2021 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2021 AN - OPUS4-53779 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Simbruner, Kai T1 - Embrittlement of Spent Fuel Claddings - Micromechanically Motivated Cohesive Zone Modelling N2 - Since cladding acts as the first barrier for nuclear fuel, it is key to understand the degrading mechanisms leading to cladding embrittlement. Zirconium alloy based cladding samples subjected to simulated pre-storage drying conditions have shown high susceptibility to brittle failure under the occurrence of hoop stress and low temperature as expected after long-term interim storage. The stress state associated with a compressive pinch-load scenario is reproduced by the Ring Compression Test (RCT), which results in sudden failure at relatively small mechanical loads with high sensitivity to the presence of radial hydrides. As part of the BAM research project BRUZL (Fracture mechanics analysis of spent fuel claddings during long-term dry interim storage), static RCTs were conducted on non-irradiated, pre-hydrided ZIRLO® ring-shaped samples that were subjected to a radial hydride treatment. All samples failed in macroscopically brittle fashion with sudden load drops and severe crack propagation through almost the entire wall thickness. In a post-RCT examination of the fracture surfaces using scanning electron microscopy, radial hydrides could be identified as the weakest link of the structure. However, the microscopic failure mechanism is much more complex due to cracks initiating at different axial positions in depth direction of the ring, overlapping cracks, and arresting cracks. A considerable portion of the fracture surfaces appeared to have undergone plastic deformation of the zirconium matrix resulting in void growth and coalescence associated with ductile failure. The geometry of the ductile patches was measured, and the ductile surface fraction was determined. Based on these observations, a statistical model was developed that can reproduce the composition of the fracture surface as a one-dimensional projection. A micromechanically motivated cohesive zone model (CZM) was implemented in existing RCT finite element analysis (FEA) models with different constitutive laws for brittle hydrides and ductile matrix respectively. In this work, FEA results are presented for non-irradiated ZIRLO® and high-burnup M5® claddings. The impact of varying hydride-matrix combinations is discussed. Acknowledgement: The project was funded by the former Federal Ministry for Economic Affairs and Energy (BMWi) under contract no. 1501561. T2 - 6th GRS Workshop on the Safety of Extended Dry Storage of Used Nuclear Fuel CY - Garching, Germany DA - 01.06.2022 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2022 AN - OPUS4-55023 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Simbruner, Kai T1 - Brittle Failure Analysis of High-Burnup PWR Fuel Cladding Alloys N2 - The general aim of this research is the development of methods for predicting mechanical behavior and identification of limiting conditions to prevent brittle failure of high-burnup (HBU) pressure water reactor (PWR) fuel cladding alloys. A finite element (FE) model of the ring compression test (RCT) was created to analyze the failure behavior of zirconium-based alloys with radial hydrides during the RCT. An elastic-plastic material model describes the zirconium alloy. The stress-strain curve needed for the elastic-plastic material model was derived by inverse finite element analyses. Cohesive zone modeling is used to reproduce sudden load drops during RCT loading. Based on the failure mechanism in non-irradiated ZIRLO® claddings, a micro-mechanical model was developed that distinguishes between brittle failure along hydrides and ductile failure of the zirconium matrix. Two different cohesive laws representing these types of failure are present in the same cohesive interface. The key differences between these constitutive laws are the cohesive strength, the stress at which damage initiates, and the cohesive energy, which is the damage energy dissipated by the cohesive zone. Statistically generated matrix-hydride distributions were mapped onto the cohesive elements and simulations with focus on the first load drop were performed. Computational results are in good agreement with the RCT results conducted on high-burnup M5® samples. It could be shown that crack initiation and propagation strongly depend on the specific configuration of hydrides and matrix material in the fracture area. T2 - 26th International Conference on Structural Mechanics in Reactor Technology - SMiRT 26 CY - Potsdam, Germany DA - 10.07.2022 KW - Cladding Embrittlement KW - Ring Compression Test KW - Cohesive Zone Modelling KW - Spent Nuclear Fuel PY - 2022 AN - OPUS4-55435 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gaddampally, Mohan Reddy T1 - Failure Analysis on Irradiated Claddings Subjected to Long-Term Dry Interim Storage N2 - Long-term dry interim storage may adversely affect the mechanical properties of spent fuel rods, possibly resulting in a reduced resilience during handling or transport after storage. Since the cladding is the first barrier for the spent fuel pellets, its integrity must be demonstrated until the end of interim storage and subsequent transportation. An established method for characterizing the cladding material is the ring compression test, in which a small, cylindrical sample of the cladding tube is subjected to a compressive load. This test is a laboratory representation of a load case where the fuel rod is crushed. Pre-storage drying and the early stage of interim storage can subject the cladding to higher temperatures and higher pressure induced tensile hoop stresses than those associated with in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate in zirconium-based alloys (Zircaloy) during slow cooling, which result in embrittlement of the cladding material and eventually a possible sudden failure of cladding integrity under additional mechanical loads. Especially long, continuous radial hydride structures and low temperature can cause severe embrittlement of claddings and finally failure by fracture even at small deformations. Therefore, the study of hydride morphology plays an important role in describing the brittle failure behaviour of the claddings. The focus of the presented research is on the development of appropriate numerical methods for predicting the mechanical behaviour and identification of limiting conditions to prevent brittle fracture of Zircaloy claddings. Typical hydride morphologies are shown. An iterative inverse analysis method is described for deriving the elastic-plastic material properties in the hoop direction of a ring-shaped sample. A modelling approach based on cohesive zones is explained which is able to reproduce the propagation of cracks initiated at radial hydrides in the zirconium matrix. The developed methods are applied to defueled samples of cladding alloy ZIRLO®, which were subjected to a thermo-mechanical treatment to reorient existing circumferential hydrides to radial hydrides. A selected sample showing sudden load drops during a quasi-static ring compression test is analysed by means of fracture mechanics for illustrative purposes. This project as part of the European Joint Programme on Radioactive Waste Management has received funding from the European Union’s Horizon 2020 research and innovation programme under grant agreement no. 847593. T2 - BAM-Kolloquium der Abteilung 3 CY - Berlin, Germany DA - 05.06.2023 KW - Cladding Embrittlement KW - Cohesive Zone Modelling KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2023 AN - OPUS4-57598 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Fracture Mechanics Analysis of Spent Fuel Claddings during Long-Term Dry Interim Storage N2 - The prevention of brittle fracture of spent fuel claddings during long-term dry interim storage is based on experimental investigations, numerical analyses and assessment methods for predicting the mechanical behavior and determining limiting conditions. The ring compression test (RCT) is an established experimental method for characterizing cladding material. Test results for various high-burnup pressure water reactor zirconium-based fuel cladding alloys (e.g., ZIRLO®, M5®) are publicly available. To reduce the effort associated with irradiated samples in hot cells, it is helpful to perform studies on unirradiated surrogate cladding material. Based on such experimental data, load-displacement curves were numerically analyzed for selected cladding materials. In the presence of radial hydrides, a sample may suddenly fail by fracture even at small deformations. Noticeable load drops in the RCT occur associated to unstable crack propagation through the radial hydride network. The failure mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth, and coalescence in the zirconium matrix, with ductile tearing patches connecting neighboring hydrides. The failure process was simulated by cohesive zones controlled by the fracture energy and the cohesive strength. A modeling approach is presented in which the radial hydride morphology is taken into account. Based on the developed fracture mechanics approach with cohesive zone modeling, not only the deformation behavior but also the failure behavior of irradiated as well as unirradiated claddings with radial hydrides under RCT loading conditions can be adequately described. T2 - 20th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 2022) CY - Juan-les-Pins, France DA - 11.06.2023 KW - Extended Interim Storage KW - Nuclear Fuel Cladding KW - Numerical Failure Analysis KW - Ring Compression Test KW - Spent Fuel Characterization PY - 2023 AN - OPUS4-59147 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - EURAD WP 8 Task 3 Overview: Behaviour of nuclear fuel and cladding after discharge N2 - Task 3 of Work package 8 (Spent Fuel Characterization) of the European Joint Programme on Radioactive Waste Management (EURAD) investigates the behaviour of nuclear fuel and cladding after discharge. The aim of the work is to understand and describe the be-haviour of spent nuclear fuel (SNF), irradiated cladding, fuel/cladding chemical interaction (FCCI) and ageing under conditions of extended interim storage, transportation and em-placement in a final disposal system. BAM contributes to the project as partner and leads Task 3. The presentation gives an overview of the project status, main achievements in experimental work and modelling studies, deviations from the plan, delays and challenges ahead. T2 - EURAD Work Package 8 (SFC) Annual Meeting CY - Wettingen, Switzerland DA - 31.10.2023 KW - Nuclear fuel KW - Cladding KW - Discharge PY - 2023 AN - OPUS4-59148 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gaddampally, Mohan Reddy T1 - Cohesive Zone Modelling Approach on Irradiated Claddings Subjected to Long-Term Dry Interim Storage N2 - Long-term dry interim storage may adversely affect the mechanical properties of spent fuel rods, possibly resulting in a reduced resilience during handling or transport after storage. Pre-storage drying and the early stage of interim storage can subject the cladding to higher temperatures and higher pressure induced tensile hoop stresses than those associated with in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate in zirconium-based alloys (Zircaloy) during slow cooling, which may result in embrittlement of the cladding material and eventually a sudden failure of cladding under additional mechanical loads. Especially long, continuous radial hydride structures and low temperature can cause severe embrittlement of claddings and finally failure by fracture even at small deformations. The focus of the presented research is on the development of appropriate numerical methods for predicting the mechanical behaviour and identification of limiting conditions to prevent brittle fracture of Zircaloy claddings. An iterative inverse analysis method is used for deriving the elastic-plastic material properties in the hoop direction of a ring-shaped sample. A modelling approach based on cohesive zones is explained which can reproduce the propagation of cracks initiated at radial hydrides in the zirconium matrix. The developed methods are applied to defueled samples of cladding alloy ZIRLO®, which were subjected to a thermo-mechanical treatment to reorient existing circumferential hydrides to radial hydrides. A selected sample showing sudden load drops during a quasi-static ring compression test is analysed by means of fracture mechanics for illustrative purposes. T2 - 7th GRS Workshop on the Safety of Extended Dry Storage of Spent Nuclear Fuel CY - Garching, Germany DA - 24.05.2023 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2023 AN - OPUS4-58553 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike T1 - Design of a Drop Test Target with Reproducible Properties for Konrad Prototype Testing N2 - Beginning in 1976 the former iron ore mine Konrad was geologically investigated concerning its suitability as a location for a German final repository for low and intermediate level waste. After a comprehensive licensing procedure it was approved and is now being prepared and equipped for the planned start of storage in 2027. At the end of the 1980s, the requirements for waste containers for storage in this repository were defined for the first time on the basis of an incident analysis of the on-site handling procedures, beginning with the unloading of the packages after the arrival at the site up to the final positioning in the underground storage galleries. A 5-meter drop onto the rock ground of the mine was identified as the covering case for high mechanical requirements (ABK II container class). In contrast to the 9-meter drop according to the IAEA Regulations for the Safe Transport of Radioactive Material, the 5-meter drop is performed onto a hard but not essentially unyielding target. The container is typically not protected by an impact limiter. The requirements for a potential test facility are described in the regulations for the Konrad repository. The mechanical strength of the target is defined as a concrete strength equivalent to the identified properties of the rock ground of the mine. Since 1991 BAM has consistently used precast reinforced concrete slabs as target for drop tests in the framework of licensing procedures as well as in research projects. While the original design fulfilled the requirement for the integrity of the concrete slab in most cases, it failed when drop tests onto an edge of a container were performed. A redesigned concrete slab developed in a research project and suggested as a reference target in 2009 has been successfully used in Konrad licensing procedures since then. The paper gives a brief overview of the historical development and it describes BAM’s efforts and approaches to continuously guarantee a concrete slab of defined quality and to provide a test setup for valid drop tests from the applicants and the authorities view. T2 - ASME 2023 Pressure Vessels & Piping Conference (PVP2023) CY - Atlanta, GA, USA DA - 16.07.2023 KW - Final repository Konrad KW - Drop test KW - Target PY - 2023 AN - OPUS4-58565 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Spent Fuel Characterization - Current Activities at BAM N2 - The European Joint Programme on Radioactive Waste Management (EURAD) is working on Spent Fuel Characterization (SFC) in its work package (WP) 8. Inspired by the EURAD activities, the International Atomic Energy Agency (IAEA) established an international Coordinated Research Project (CRP) on SFC. The EURAD WP SFC participants are collaborating as a team on the IAEA CRP on SFC. The EURAD WP SFC project consists of four tasks. Task 3 investigates the behaviour of nuclear fuel and cladding after discharge. The aim of these activities is to understand and describe the evolution of the cladding-pellet system and its ageing under conditions of extended interim storage, transportation and emplacement in a final disposal system. At a Consultancy Meeting, BAM as contributor to Task 3 presented current results of the failure analysis of irradiated ZIRLO® claddings under conditions of the Ring Compression Test. T2 - IAEA Consultancy Meeting on the Coordinated Research Project on Spent Fuel Characterization CY - Online meeting DA - 12.09.2023 KW - Nuclear Fuel Cladding KW - Numerical Failure Analysis KW - Ring Compression Test KW - Spent Fuel Characterization KW - Extended Interim Storage PY - 2023 AN - OPUS4-58276 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - 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 - Zencker, Uwe T1 - Brittle Failure Limits of Spent Fuel Claddings Subjected to Long-Term Dry Interim Storage Conditions N2 - The mechanical properties of spent fuel claddings can be adversely affected under the conditions of long-term dry interim storage, so that the failure limits may be reached in case of mechanical loads during handling or transport after storage. Pre-storage drying and the early stage of interim storage can expose the cladding to higher temperatures and higher tensile hoop stresses than those associated with in-reactor operation and pool storage. During slow cooling of a cladding tube under internal pressure, radial hydrides may precipitate in zirconium-based cladding alloys. This can lead to embrittlement of the material and sudden failure of the cladding integrity under mechanical stress. In order to prevent brittle failure, numerical methods are being developed to predict the mechanical behaviour and identify limiting conditions. Experimental investigations, numerical analyses and evaluation methods are discussed. An established experimental method for characterising cladding materials is the Ring Compression Test (RCT). Some test results on irradiated cladding tubes after operation in pressure water reactors are publicly available. However, it is helpful to carry out studies on unirradiated surrogate claddings with similar material properties to reduce the effort associated with irradiated samples in hot cells and to perform material tests with a wider range of parameters. On the basis of such experimental data, load-displacement curves have been numerically analysed for a selection of cladding materials. Radial hydrides can cause a sample to break suddenly due to fracture even at low deformation. Noticeable load drops in the RCT are caused by unstable crack propagation through the radial hydride network. The failure mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth, and coalescence in the zirconium matrix, with ductile tearing patches connecting adjacent hydrides. The cohesive zone approach was used to simulate the failure process taking into account the radial hydride morphology. The developed method can adequately describe both the deformation and failure behaviour of irradiated as well as unirradiated claddings of zirconium-based alloys with radial hydrides under RCT conditions. Limiting conditions can be expressed in terms of fracture energy and cohesive strength. T2 - IAEA International Conference on the Management of Spent Fuel from Nuclear Power Plants - Meeting the Moment CY - Vienna, Austria DA - 10.06.2024 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2024 AN - OPUS4-60324 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Fracture Mechanics Analysis of Spent Fuel Claddings during Long-Term Dry Interim Storage N2 - The 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 Third Research Coordination Meeting on Spent Fuel Characterization, BAM as contributor to Task 3 reported on the activities and status of its research work. T2 - IAEA 3rd Research Coordination Meeting of the Coordinated Research Project on Spent Fuel Characterization CY - Cockermouth, UK DA - 01.07.2024 KW - Nuclear fuel KW - Cladding KW - Discharge PY - 2024 AN - OPUS4-60668 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Failure Analysis on Irradiated Claddings Subjected to Long-Term Dry Interim Storage N2 - Long-term dry interim storage may adversely affect the mechanical properties of spent fuel rods, possibly resulting in a reduced resilience during handling or transport after storage. The cladding is the first barrier for the spent fuel pellets. An established method for characterising the cladding material is the ring compression test (RCT), in which a small, cylindrical sample of the cladding tube is subjected to a compressive load. Radial hydrides may precipitate in zirconium-based alloys (Zircaloy) under pre-storage drying and during slow cooling, which result in embrittlement of the cladding material and eventually a possible sudden failure of cladding integrity under additional mechanical loads. The focus of the presented research is on the development of appropriate nu-merical methods for predicting the mechanical behaviour and identification of limiting conditions to prevent brittle fracture of Zircaloy claddings. A modelling approach based on cohesive zones is ex-plained which is able to reproduce the propagation of cracks initiated at radial hydrides in the zir-conium matrix. The developed methods are applied to defueled samples of cladding alloy ZIRLO®, which were subjected to a thermo-mechanical treatment to reorient existing circumferential hy-drides to radial hydrides. A selected sample showing sudden load drops during a quasi-static ring compression test is analysed by means of fracture mechanics for illustrative purposes. Based on the developed fracture mechanics approach, not only the deformation behaviour but also the fail-ure behaviour of irradiated as well as unirradiated Zircaloy claddings with radial hydrides under RCT loading conditions can be adequately described. T2 - 27th International Conference on Structural Mechanics in Reactor Technology - SMiRT 27 CY - Yokohama, Japan DA - 03.03.2024 KW - Cladding Embrittlement KW - Cohesive Zone Modelling KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2024 AN - OPUS4-60672 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike T1 - Consequences of an imperfectly mounted reinforcement cage in a generical cylindrical concrete container during mechanical specimen tests N2 - By highest court decision in 2007 the issued license for the Konrad repository (a former iron ore mine) for low and intermediate level radioactive waste was finally confirmed. Since then, the site is improved, prepared and equipped for the currently planned start of waste package emplacement by 2029. The Bundesanstalt fuer Materialforschung und -pruefung (BAM, Federal Institute for Materials Research and Testing) is regularly contracted by the Bundesgesellschaft für Endlagerung (BGE, Federal Company for Radioactive Waste Disposal) as the responsible authority for waste product control, waste container design testing, and waste package acceptance. In this context BAM is mainly responsible for design testing of various container types including the evaluation quality assurance measures for container manufacturing. In the written Konrad disposal requirements [1] [2], all acceptance criteria for potential final disposal containers are defined. In addition to general requirements concerning container design, materials, dimensions and others, specific mechanical load scenarios the casks have to withstand are derived from an analysis of the on-site handling and emplacement procedures. Drop tests from different heights under consideration of the most damaging drop orientation and a stacking test are two examples for safety demonstrations that have to be performed during the container design testing procedure to qualify the respective container type. The containers, whether cubic or cylindrical containers, can be made from different materials like ductile cast iron, reinforced concrete or as welded construction from steel plates and profiles. In case of concrete containers the reinforcement has the function to absorb tensile forces the concrete is not able to withstand by its own. Because of manufacturing imperfections of cylindrical concrete waste containers manufactured already decades ago, the reinforcement cages are not always exactly, axisymmetrically positioned when they were poured. This inspired BAM to perform various ABAQUS finite-element (FE) simulations [3] by using the example of a simplified cylindrical container design with generic dimensions and pure elastic material properties. As outcome the stress conditions in the container during the drop test impact under different drop positions and during the stacking test were investigated and are explained and illustrated in this contribution. Different reinforcement cage positions were calculated and compared. On the basis of the calculated tensile and compressive stress distributions suggestions are derived on how a cylindrical concrete container with an imperfectly placed reinforcement cage has to be oriented for a drop or stacking test to ensure the most damaging test scenario. T2 - ASME 2024 Pressure Vessels & Piping Conference CY - Bellevue, Washington, USA DA - 28.07.2024 KW - Container design testing procedure KW - Reinforced concrete containers KW - Drop test KW - Stacking test KW - Imperfectly mounted reinforcement cage KW - Finite-element-analysis PY - 2024 AN - OPUS4-62560 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Analysis of Zirconium Hydride Structures in Spent Fuel Claddings Subjected to Storage Conditions N2 - Pre-storage drying and the early stage of storage can subject spent fuel claddings to higher temperatures and higher tensile hoop stresses than those associated with in-reactor operation. Under these conditions, existing circumferential hydrides in zirconium-based alloys can be partly reoriented into radial ones that may result in embrittlement of the cladding material. As a consequence, the failure limits may be reduced under mechanical loads caused by handling or during transport after storage. The potential occurrence of brittle fracture strongly depends on the specific location, size, and orientation of the zirconium hydrides. The morphology of such hydride structures was experimentally investigated in unirradiated, hydrogen-loaded samples made of a zirconium-based alloy. The hydrogen was introduced into the cladding sample by means of a cathodic charging process. The loading process was followed by a thermal treatment to precipitate the hydrides in the circumferential direction and then a thermomechanical treatment to reorient the hydrides in the radial direction. Metallographic methods were used to describe and classify the hydride morphology. The resulting metallographic micrographs were evaluated using suitable radial hydride metrics. The determination of the maximum length of a hydride structure was carried out in two different ways, each with its own advantages and disadvantages. Both methods consider neighboring hydrides as continuous according to the definition of the Radial Hydride Continuity Factor (RHCF). The first option is to measure the Euclidean distance between the two most distant pixels of the hydride structure in the image analysis, whereby the actual maximum size of the hydride structure is determined. The second possibility is to measure the maximum size of a hydride or a hydride structure within a certain arc length in the radial direction of the cladding tube. This value is of particular interest because the ratio of the maximum radial size of a hydride structure to the wall thickness of the cladding tube is used for the safety assessment of spent fuel claddings with radial hydrides using the RHCF. T2 - 21st International Symposium on the Packaging and Transportation of Radioactive Materials - PATRAM 2025 CY - San Antonio, TX, USA DA - 27.07.2025 KW - Extended Interim Storage KW - Nuclear Fuel Cladding KW - Zirconium Hydride Structure KW - Ring Compression Test KW - Micrograph Analysis PY - 2025 AN - OPUS4-63876 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe T1 - Failure Analysis on Spent Fuel Claddings Subjected to Long-Term Dry Interim Storage N2 - Long-term dry interim storage may adversely affect the mechanical properties of spent fuel claddings, possibly resulting in a reduced resilience during handling or transport after storage. Pre-storage drying and the early stage of interim storage can subject the cladding to higher temperatures and higher tensile hoop stresses than those associated with in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate in zirconium-based alloys (Zircaloy) during slow cooling, which result in embrittlement of the cladding material and eventually a possible sudden failure of cladding integrity under additional mechanical loads. Especially long, continuous radial hydride structures and low temperature can cause severe embrittlement of claddings and finally failure by fracture even at small deformations. Therefore, the study of hydride morphology plays an important role in describing the brittle failure behaviour of the claddings. The focus of the presented research is on the development of appropriate numerical methods for predicting the mechanical behaviour and identification of limiting conditions to prevent brittle fracture of Zircaloy claddings. An established experimental method for characterising cladding materials is the Ring Compression Test (RCT). Some test results on irradiated cladding tubes after operation in pressure water reactors are publicly available. However, it is helpful to carry out studies on unirradiated surrogate claddings with similar material properties to reduce the effort associated with irradiated samples in hot cells and to perform material tests with a wider range of parameters. On the basis of such experimental data, load-displacement curves have been numerically analysed for a selection of cladding materials. Radial hydrides can cause a sample to break suddenly. Noticeable load drops in the RCT are caused by unstable crack propagation through the radial hydride network. The failure mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth, and coalescence in the zirconium matrix, with ductile tearing patches connecting adjacent hydrides. The cohesive zone approach was used to simulate the failure process taking into account the radial hydride morphology. Experimental investigations, numerical analyses and evaluation methods for both the deformation and failure behaviour of irradiated as well as unirradiated claddings of zirconium-based alloys with radial hydrides are discussed. Limiting conditions can be expressed in terms of fracture energy and cohesive strength. T2 - IAEA Technical Meeting on the Behaviour of Spent Fuel and Cladding During Storage and the Performance of Spent Fuel Storage Systems CY - Seoul, South Korea DA - 23.06.2025 KW - Extended Interim Storage KW - Nuclear Fuel Cladding KW - Numerical Failure Analysis KW - Ring Compression Test KW - Spent Fuel Characterization PY - 2025 AN - OPUS4-63597 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger T1 - Innovative Disposal Container Materials: Improved Durability and Manufacturing Feasibility N2 - Deep geological disposal of high-level radioactive waste (HLW) is the internationally preferred final waste management option to prevent the biosphere from radiologically relevant release of radioactive nuclides for at least 100,000 to 1 million years. Current repository concepts are based on a multi-barrier system consisting of geological, geo-technical and technical barriers as an engineered barrier system (EBS). Considered host rocks are e.g. crystalline rocks, claystone, or rock salt. In all cases disposal containers are a crucial technical barrier to safely enclose the HLW, either spent nuclear fuel (SNF) or vitrified high-level waste (VHLW) from reprocessing. Depending on the repository concept and the respective safety case disposal containers guarantee the safe confinement of their radioactive inventory not just during handling, emplacement, and potential retrieval but also during the post-closure phase of the repository for a certain period. Required container lifetimes usually vary between 1,000 and 100,000 years depending on the selected host rock and repository concept. During container operation and their long-term disposal, they are subject to various mechanical, thermal, radiological, and geo-chemical/biological loads. Thus, mechanical stability and corrosion resistance in the long-term are among the most crucial challenges. Disposal container components lifetimes are typically calculated based on their time-dependent corrosion behaviour, with the implicit assumption that the design remains structurally stable for the required period. Even though, the durability of several component materials subjected to corrosion processes have been previously studied in detail, the interaction of mechanical processes and corrosion calls for further study, and assessment of the impact of joint degradation modes on component lifetimes will result in a more robust and defensible safety case. Besides, as corrosion occurs in a thin interfacial surface layer between the component outermost surface and the environment, specific R&D work is required to understand the long-term performance controlled by the entire engineered barrier system (EBS) and to feed simulation and extrapolations tools by representative experimental data for validation. Another approach to tackle the corrosion issue is to prevent and/or minimize it, thereby ensuring an even safer disposal. In this approach, materials much less prone to corrosion (e.g., copper, ceramic materials) can be used to fabricate the containers. Alternatively, protective coatings (e.g. copper coating) can be applied to current selected reference materials like steels. Both ways require validation of the materials durability under realistic, accelerated field conditions. The InCoManD work package 9 (Innovative and new Container/canister materials under disposal field conditions), part of the EURAD-2 European Partnership on Radioactive Waste Management (2024 – 2029), will address these issues through a collaborative project involving many countries across Europe (and beyond) with a shared goal. Building on the results of the ConCorD work package 15 (Container Corrosion under Disposal Conditions), part of the EURAD-1 programme (2019 - 2024), the InCoManD work package specifically aims to: (i) provide a better understanding of material degradation mechanisms, (ii) define optimised and innovative material solutions, (iii) develop comprehensive predictive models and common methodologies to enhance confidence in the results produced by each partner, and (iv) train new scientists in this field. T2 - safeND 2025 CY - Berlin, Germany DA - 17.09.2025 KW - Radioactive Waste Disposal KW - Disposal Container KW - Deep Geological Repository KW - Engineered Barrier System PY - 2025 DO - https://doi.org/10.5194/safend2025-40 AN - OPUS4-64254 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kömmling, Anja T1 - Experiences and perspectives for the design evaluation of Konrad containers N2 - The former iron ore mine Konrad near Salzgitter is the approved final repository for radioactive waste with negligible heat production in Germany and should go into operation in the 2030s. The waste comes from operation and decommissioning of nuclear power plants, as well as from medicine, research and industry using radiation sources or radioactive materials. These types of waste are comparable to international low- and certain types of intermediate-level waste. The waste includes a large variety of products, materials and properties. The containers for such waste have the primary purpose of enabling handling and storing of the waste by safely enclosing the radioactive inventory and shielding of radiation. Due to the large variety of waste forms to be disposed of in the Konrad repository, different container types are specified to account for the different requirements related to each waste type. Each container design for waste disposal in Konrad has to be approved by BGE on the basis of the “final disposal conditions” (Endlagerungsbedingungen) and the respective “product control” (Produktkontrolle). For demonstrating the container’s compliance with the regulatory framework, the waste producers or container manufacturers have to apply for design approval on the basis of a comprehensive safety assessment including all relevant reports about analytical, numerical and experimental safety demonstrations and quality assurance measures for container manufacturing and operation. Usually, BGE commissions independent experts like BAM to evaluate these documents and the design testing (e.g. drop or fire tests). While the basic assessment principle is to verify the container design and quality to be in line with all regulatory requirements, there are often challenges associated with the container’s safety evaluation. As containers of different types have already been produced in large numbers and partially loaded with waste before the regulatory framework became effective (so-called “old” containers), they are considered in the regulations as well and safety assessments have to be provided in an equivalent way. As the requirements for the documentation of manufacturing quality were not yet known at the time these containers were manufactured, this is often challenging in practice. On the other hand, containers requiring the highest safety level (ABK II, sf) due to their nuclear content sometimes cause significant challenges and very long approval processes concerning safety demonstrations for severe accidental conditions like a 5 m drop without impact limiter onto a nearly unyielding target or a one hour 800 °C fire scenario. Furthermore, as the regulatory framework has not been updated for several decades, some requirements and the respective safety assessment methods have to be interpreted under consideration of the current state of knowledge in science and technology. T2 - safeND conference 2025 CY - Berlin, Germany DA - 17.09.2025 KW - Radioactive waste KW - Final disposal KW - Waste container KW - Design evaluation PY - 2025 AN - OPUS4-64989 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Banerjee, Ayantika T1 - Enhancing Requirements for HLW Disposal Containers and Developing Transparent Safety Evaluation Concepts in Collaboration between BAM and BASE (Research Project KAnnE) N2 - The research project KAnnE aims to identify, concretize, and quantify requirements for high-level radioactive waste (HLW) disposal containers in Germany, considering potential types of host rock and repository concepts. The primary goal is to establish a transparent and comprehensible regulatory framework that accounts for different host rocks and operational phases, ensuring the safe long-term containment of radioactive waste. Given the current general legal formulation of these requirements, this study provides a systematic approach to defining them more precisely by integrating scientific and regulatory perspectives. An extensive review is conducted to gather relevant information about safety and design requirements for disposal containers in the respective types of host rock. The analysis considers host rock properties and conditions, operational repository conditions, and their impact on the container design, including key factors such as structural integrity, corrosion resistance, and shielding. Additionally, design specifications, boundary conditions, and expected loads (thermal, mechanical, radiological, chemical, and biological) are systematically reviewed, with quantified values provided where available. The findings support a comprehensive evaluation of both quantitative and qualitative container requirements concerning potential German host rock types, waste forms, and repository concepts. By establishing a clearer basis for defining these requirements, this project serves as a foundation for further refinement, with a more detailed structuring of requirements currently in progress. Such regulatory basis allows a more targeted approach to container development and ensures a transparent evaluation within the repository licensing procedure. Additionally, it supports decision making on the most favorable container design by considering repository conditions that provide the highest level of safety over one million years. This contribution explains the goals and concept of the research project KAnnE, presents its preliminary findings and conclusions, and provides an outlook on the final project phase. T2 - safeND conference 2025 CY - Berlin, Germany DA - 17.09.2025 KW - Radioactive waste KW - Final disposal KW - Waste container KW - Requirements KW - Regulation PY - 2025 AN - OPUS4-64990 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zaghdoudi, Maha 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 - Weber, Mike 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 - Völzke, Holger T1 - Research at BAM on Metal Seal Performance and Fuel Rod Integrity N2 - The presentation provides an update on preliminary results from research projects in the area of long term performance of metal seals and fuel rod integrity as safety relevant components of spent fuel transport and storage casks for spent nuclear fuel. T2 - Extended Storage Collaboration Program (ESCP) Winter Meeting 2020 CY - Online meeting DA - 09.11.2020 KW - Metal seal KW - Safety KW - Interim storage KW - Spent nuclear fuel KW - Fuel rod PY - 2020 AN - OPUS4-51615 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zaghdoudi, Maha 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 - Jaunich, Matthias T1 - An update on metal seal tests performed at BAM and implications for interim storage N2 - The investigation of the long-term performance of sealing systems employed in containers for radioactive waste and spent nuclear fuel is one research focus area for division 3.4 “Safety of Storage Containers” at the Bundesanstalt für Materialforschung und -prüfung. Metallic seals are in use in interim storage casks for used fuel and high active waste. Their change in performance over time is of high importance for maintaining safe enclosure. Therefore, investigations on these systems were started at BAM to get a general understanding of the relevant processes. Our investigations comprise investigations on different parameters which influence the seal performance and the main part is focussed on the time and temperature dependent behaviour. In this contribution an update on the performed tests and their respective results will be given in respect to the overview presented in 2020. The understanding of the long-term behaviour of the sealing systems and the performance evaluation during interim storage and subsequent transportation are our goals. T2 - SEDS-Workshop: Sicherheit der längerfristigen trockenen Zwischenlagerung CY - Garching, Germany DA - 15.05.2024 KW - Metal seals KW - Ageing KW - Leakage rate KW - Creep PY - 2024 AN - OPUS4-60640 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zaghdoudi, Maha T1 - Overview of ongoing research and future prospects on polyethylene neutron shielding materials at BAM N2 - The extension of the interim storage period of radioactive waste before disposal will cause additional challenges for the nuclear waste management in Germany, so that an extensive knowledge of the long-term performance of casks, including their components and inventories, will be required for future extended storage licenses. Ultra-high and high molecular weight polyethylenes ((U)HMW-PE) are used for neutron shielding purposes in casks for storage and transport of spent fuel and high-level waste due to their extremely high hydrogen content. During their service life of several decades as cask components, the PE materials are exposed to neutron and gamma radiation from the radioactive inventory of the casks, mechanical assembling stresses and temperature. All these combined effects affect the material properties of such components which in turn may be crucial for some possible accident scenarios. At the Bundesanstalt für Materialforschung und -prüfung (BAM), the effects of high temperature exposure in combination with subsequent or previous irradiation were investigated with a comprehensive aging program including thermal aging at 125 °C for different aging periods up to 5 years and irradiation with doses ranging from 50 to 600 kGy. This contribution provides an overview of the ongoing research related to the structural changes of (U)HMW-PE induced by gamma irradiation and high temperature exposure and focuses on current research perspectives at BAM with regard to the prediction of the dynamic behavior of the material during extended interim storage in case of an accident scenario. First results of the coupled effect of temperature, radiation and mechanical loading will be presented. The effect of microstructural changes induced by gamma irradiation and high temperature on the mechanical behavior of (U)HMW-PE will be assessed. T2 - PATRAM22 CY - Juan-les-Pins, France DA - 11.06.2023 KW - (U)HMWPE KW - Ageing KW - Irradiation KW - WAXD KW - SHPB PY - 2023 AN - OPUS4-57708 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kömmling, Anja T1 - Aging of elastomer O-rings and PE neutron shielding materials for radioactive waste containers N2 - Our institution BAM in Berlin, Germany is concerned with research and testing of materials in the context of safety in chemistry and technology. Our working group is involved in the licensing procedures of casks for radioactive waste. Besides, we’re doing research on aging and lifetime prediction of elastomer O-rings and investigate degradation and thermal expansion of PE neutron shielding materials. T2 - Polymers in nuclear applications CY - Online meeting DA - 01.12.2021 KW - Rubber KW - Polyethylene KW - Thermal expansion KW - Lifetime KW - Degradation PY - 2021 UR - https://energiforsk.se/media/30631/bam_material_aging_analysis.pdf AN - OPUS4-54191 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Liu, Y. T1 - Cooperation in Advanced Technologies Associated with the Back End of Nuclear Fuel Cycle N2 - A Memorandum of Understanding (MOU) between Argonne National Laboratory (Argonne) and the German Bundesanstalt für Materialforschung und -prüfung (BAM) was signed in October 2014. Its objectives are to promote cooperation among scientists and specialists at Argonne and BAM and establish a framework for collaboration in advanced technologies associated with the back end of the nuclear fuel cycle. Collaborative activities involving Argonne and BAM may be implemented through the promotion of joint research activities and scientific workshops and conferences; exchange of technical information; and visits by scientists, specialists, and graduate, postgraduate, and Ph.D. students. The MOU was renewed in January 2020 for another five years. Highlights of the progress and achievements are provided in identified topical areas for collaboration on ageing management guidance, storage, transportation and disposal R&D, joint conference activities, and conclude with the benefits of the MOU in fostering Argonne and BAM collaboration. T2 - 26th International Conference on Structural Mechanics in Reactor Technology - SMiRT 26 CY - Potsdam, Germany DA - 10.07.2022 KW - Nuclear fuel cycle KW - Back end KW - Spent fuel performance KW - Storage KW - Transportation PY - 2022 AN - OPUS4-55862 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Sagradov, Ilja T1 - Long-term tests on metal seals using continuous force measurement in compressed state N2 - In Germany, spent fuel and heat generating radioactive waste is stored in casks for transport and storage (DPCs). The storage licenses were granted for up to 40 years. Due to the delay in identifying a final repository site for such radioactive waste, the storage periods are to be extended significantly. The objective of our research work is to collect additional material data for metal seals in order to be able to better assess the long-term safety of the sealing system of the casks for a period of up to 100 years. A principal safety consideration for the DPCs is the long-term performance of bolted lid sealing systems utilizing metal seals to ensure the safe enclosure of the radioactive inventory during interim storage. To assess the sealing behavior, it is necessary to examine the aging-dependent residual restoring force for the metal seals and the usable resilience. By investigating the change in these values as a function of time and temperature, further insight can be gained into the assessment of long-term safety for extended interim storage. In the present study, the restoring force of metal seals with an outer aluminum jacket and a torus diameter of 9.9 mm was measured continuously. For this purpose, a specially designed device equipped with a load cell was used under isothermal conditions. As part of the investigation, the level and rate of decrease of the restoring seal force were recorded over test periods of one and two months. The seal temperature was set to 35 °C, 75 °C, 100 °C, 125 °C, and 150 °C, respectively. A direct correlation was observed between the temperature of the seal, the level of the restoring force after compressing and heating the seal, and the drop-off characteristics of the restoring force. It was found that in our test device the force level immediately after reaching the respective temperature is lower for higher seal temperatures, while the drop rate of the restoring force is higher. This paper describes the experimental setting and discusses first results in comparison with our long-term discontinuous measurements. T2 - 21st International Symposium on the Packaging and Transportation of Radioactive Materials - PATRAM 2025 CY - San Antonio, TX, USA DA - 27.07.2025 KW - Long-Term Safety Assessment KW - Extended Interim Storage KW - Metal Seal Aging KW - Restoring Force Degradation KW - Temperature-Dependent Performance PY - 2025 AN - OPUS4-63880 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias T1 - Investigations at BAM on Fuel Cladding Integrity and DPC Seal Performance N2 - The presentation provides an update on preliminary results from research projects in the area of long term performance of metal seals and fuel rod integrity as safety relevant components of spent fuel transport and storage casks for spent nuclear fuel. T2 - Extended Storage Collaboration Program (ESCP) winter meeting 2022 CY - Charlotte, NC, USA DA - 07.11.2022 KW - Metal seal KW - Fuel cladding KW - Ring compression test KW - Brittle failure PY - 2022 AN - OPUS4-56400 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias T1 - Overview of metal seal tests performed at BAM and implications for extended interim storage N2 - The investigation of the long-term performance of sealing systems employed in containers for radioactive waste and spent nuclear fuel is one research focus area for division 3.4 “Safety of Storage Containers” at the Bundesanstalt für Materialforschung und -prüfung. The behaviour of metallic seals, which are employed in interim storage casks for used fuel and high active waste, over time is of high importance for ensuring safe enclosure. Therefore, investigations on these systems were started at BAM to get a general understanding of the relevant processes. Our investigations comprise investigations on different parameters which influence the seal performance and the main part is focussed on the time and temperature dependent behaviour. In this contribution an overview of the performed tests and their respective results will be given. A focus will lie on a comparison between seals with large torus diameter in comparison to small diameters. Our aim is to understand the long-term behaviour of the sealing systems for evaluation of their performance during possible extended interim storage and subsequent transportation. T2 - 4th Workshop on Safety of Extended Dry Storage of Spent Nuclear Fuel CY - Online meeting DA - 03.06.2020 KW - Seal performance KW - Metal seal KW - Leakage rate PY - 2020 AN - OPUS4-50843 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike T1 - A numerical approach to correlate compression stress relaxation and compression set of elastomer O-rings with tightness 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 AN - OPUS4-54423 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zaghdoudi, Maha T1 - Numerical modelling of compression stress relaxation and compression set of elastomer O-rings during aging N2 - Elastomer seals are used in many industrial applications due to their excellent mechanical properties at a wide range of temperatures. Their high versatility and recovery potential under several load conditions make them well suitable for the application in containers designed for transport, storage and/or disposal of radioactive materials. In containers for low and intermediate level radioactive waste, elastomer seals are used as barrier seals, and as auxiliary seals in storage and transportation casks (dual purpose casks) for heat generating radioactive waste, such as spent fuel and high-level waste. While a seal exchange at defined intervals is typical in many conventional applications, it is impossible or at least hard to perform when principles of minimization of radiation exposure have to be considered and prohibit an unnecessary cask handling. An extensive knowledge of the change of the elastomer’s properties during aging and the availability of reliable end-of-lifetime criteria to guarantee the permanent safe enclosure of the radioactive material is mandatory. As BAM is involved in most of the national cask licensing procedures and in the evaluation of cask-related long-term safety issues, great efforts have been already made and are still planned to scientifically support this task. Compression stress relaxation and compression set were identified as key indicators of elastomer long-term performance and quantitatively investigated in comprehensive test programs. Among other representative types of elastomers, specimens made from ethylene propylene diene rubber (EPDM) were tested before, during and after aging to capture the most important of their complex mechanical properties. In the presented study, exemplary results were used to simulate the compression stress relaxation and the compression set of elastomer O-rings during aging. Regarding the influence of temperature, the time temperature superposition principle is applied in the relaxation analysis of elastomer O-rings. The proposed model is implemented in the commercial finite element software ABAQUS/Standard® [1] with a sequential temperature displacement coupling. Numerical results match the experimental compression stress relaxation measurements well. The prediction of compression set values after long-term aging shows a relatively good agreement with the experimental results. Nevertheless, all input parameters derived from the specimen tests, additional assumptions concerning boundary conditions and modeling strategy are discussed with regard to the identified slight discrepancies. The possibility to extend the finite element model to represent the O-ring seal’s ability to recover after a (fast) partial release is taken into account T2 - ASME 2020 Pressure Vessels & Piping Conference (PVP2020) CY - Online meeting DA - 03.08.2020 KW - Stress relaxation KW - Ageing KW - Compression Set KW - O-ring KW - Sequential analysis KW - Simulation PY - 2020 AN - OPUS4-52135 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias T1 - Elastomer seals - overview and development N2 - The talk summarizes the current and future investigations of elastomer seal materials in division 3.4. T2 - IRSN-BAM Symposium Safety of Transport and Storage Packages CY - Online meeting DA - 19.11.2020 KW - Seal KW - Degradation KW - Ageing PY - 2020 AN - OPUS4-51751 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias T1 - Aging investigations of metal seals used in containers for interim storage of spent fuel N2 - In Germany dual purpose containers with double lid systems are used for the transport and storage of spent fuel and high level radioactive waste. The bolted lids are equipped with metal seals (e.g. Helicoflex®) to ensure the safe enclosure of the inventory for the required time period. The used metal seals have a layered structure consisting of three components. In the center a helical spring is positioned that is surrounded by two C-shaped jackets. The helical spring is mainly responsible for generation of the required restoring force. The inner jacket is made of stainless steel and homogenizes the restoring force generated by the helical spring. The outer jacket is made of silver or aluminum. Both metals are soft in comparison to the contact partners (lid and container body). During bolting of the lid the seal is compressed between lid and container body. The generated restoring force of the helical spring causes a plastic deformation of the outer jacket which adapts to the surfaces of the lid and the container body. Hence, leakage paths are closed and the sealing function is caused. Typical durations for existing interim storage licenses in Germany are 40 years, but due to the challenges in selection and construction of a final repository it can be expected that they have to be extended to longer periods. An extension of license periods requires knowledge of the long-term behavior of the seals. To meet this challenge long-term investigations have been started at Bundesanstalt für Materialforschung und –prüfung (BAM) in 2009. These tests focus on seals assembled in test flanges which are stored at temperatures ranging from room temperature to 150 °C for accelerated aging. The aged seals are tested repeatedly after certain aging steps and the performance is evaluated in respect to the leakage rate, the remaining seal force, and the useable resilience upon decompression. In this contribution an update on the running investigations in respect to earlier publications (Grelle et al. 2019, Goral et al. 2023) will be given and the implications of the results for resilient long-term safety will be presented. New activities will be addressed and their expected outcome be discussed. T2 - PVP2024, Pressure Vessels & Piping Conference CY - Bellevue, WA, USA DA - 29.07.2024 KW - Metal seal KW - Aging KW - Interim storage PY - 2024 AN - OPUS4-61355 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias T1 - Low temperature properties of rubber seals N2 - Understanding the failure mechanisms that lead to leakage of rubber seals at low temperatures is important. Rubber properties are highly temperature dependent. Changes in behaviour at low temperatures can result from the rubber-glass transition and for some rubbers by partial crystallisation. Experimental results on the low temperature behaviour of different rubber seals were discussed to help understand seal failure processes. T2 - Rubber in Engineering Group (RIEG) webinar an Elastomers at Low Temperatures CY - Online meeting DA - 11.09.2020 KW - Low temperature KW - Rubber seal KW - Compression set KW - Glass transition PY - 2020 UR - https://www.iom3.org/rubber-engineering-group/event/elastomers-low-temperatures AN - OPUS4-51306 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias T1 - Ageing under Hydrogen: effects on O-ring seals N2 - In this work we present results for three typical sealing materials (EPDM, HNBR, FKM) after up to 100 days ageing at high temperature under hydrogen. To allow for a better evaluation of the occurring changes the results are compared with effects of ageing in air on the material properties. Despite the common assumption that ageing under hydrogen atmosphere is expected to be less severe for the material in com-parison to air ageing, this is not seen for every material. T2 - Technomer 2023 CY - Chemnitz, Germany DA - 09.11.2023 KW - Hydrogen KW - Ageing KW - Seal KW - Compression set PY - 2023 AN - OPUS4-58818 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias T1 - Comparison of Ageing Behavior of O-Ring Seals Under Hydrogen and Air N2 - Elastomeric seals are essential components in the infrastructure which prevent leakage of gas and ensure proper function of technical devices and are therefore highly safety relevant. For proper function a remaining resilience and tolerance to pressure changes is required. The ageing of elastomers is an undesirable but inevitable process leading to a limited lifetime of rubber products such as seals. Independent of the application, as e.g. automotive, piping or container applications, a long lifetime or at least a timely exchange of a seal before occurrence of critical failure is desired. Hydrogen is a key element for the energy turnaround and therefore the compatibility of seals with hydrogen is essential. In addition to the effect of hydrogen uptake and release, which might also cause destructive effects such as rapid gas decompression, ageing at high temperature under hydrogen atmosphere is of interest too. Most of the existing work is addressing the performance of new materials and comprises only very limited investigations concerning long-term use and the behaviour of aged materials in contact with hydrogen. As ageing can lead to substantial changes of material properties, it must be evaluated whether these changes are beneficial or deteriorating for the component function. In this work we present and compare results of the characterization of three sealing materials (EPDM, HNBR, FKM) after ageing at high temperature under hydrogen and air. Despite the common assumption that ageing under hydrogen atmosphere should be less severe for the material in comparison to air ageing, this is not the case for every material. T2 - ASME PVP2025, Pressure Vessels & Piping Conference CY - Montreal, Canada DA - 20.07.2025 KW - Aging KW - Seal KW - Compression set KW - Atomic force microscopy PY - 2025 AN - OPUS4-63888 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias T1 - Elastomerdichtungen für Wasserstoffanwendungen - Vergleich des Alterungsverhaltens in Luft und Wasserstoff N2 - In diesem Beitrag soll ein Fokus auf das Alterungsverhalten von Elastomerdichtungen unter Wasserstoff gelegt werden. Hierfür wurden Untersuchungen bei einem mittleren Wasserstoffdruck über Alterungszeiten von bis zu 100 Tagen durchgeführt. Das Alterungsverhalten wird im Vergleich der Alterungseffekte unter thermooxidativen Bedingungen betrachtet und bewertet. Die Annahme, dass der Alterungseffekt unter Wasserstoff stets weniger ausgeprägt ist als an Luft bestätigt sich nicht durch die Ergebnisse. T2 - 4. O-Ring-Forum der Firma ISGATEC CY - Bad Schönborn, Germany DA - 15.05.2025 KW - Wasserstoff KW - Dichtung KW - Thermische Alterung KW - Elastomer PY - 2025 AN - OPUS4-63135 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -