TY - CONF A1 - Völzke, Holger A1 - Wieser, Günter A1 - Zencker, Uwe A1 - Droste, Bernhard T1 - Mechanical and thermal safety analyses and demonstrations for cubic DCI - multi-purpose containers T2 - 4th International conference on transportation for the nuclear industry CY - Bournemouth, Dorset, UK DA - 1997-05-13 PY - 1997 IS - Session 2, Paper 5 SP - 1 EP - 13 AN - OPUS4-21665 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Völzke, Holger A1 - Droste, Bernhard T1 - Entwicklung von Beurteilungsmethoden für Transport- und Lagerbehälter mit erhöhten metallischen Reststoffanteilen (EBER) - FEM-Berechnungen zum Aufprall kubischer Gußcontainer T2 - V. Stilllegungskolloquium Hannover und 4. Statusbericht "Stilllegung und Rückbau kerntechnischer Anlagen" CY - Hannover, Deutschland DA - 1997-06-24 PY - 1997 SP - 264 EP - 275 AN - OPUS4-21659 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Protz, Christian A1 - Völzke, Holger A1 - Ellouz, Manel A1 - Kasparek, Eva Maria A1 - Zencker, Uwe A1 - Grünewald, Heike A1 - Gründer, Klaus-Peter T1 - ConDrop - Development of numerical simulation approaches for drop test analyses of steel sheet containers T2 - 53rd INMM Annual meeting CY - Orlando, FL, USA DA - 2012-07-15 KW - INMM 53rd annual meeeting PY - 2012 SP - 1 EP - 7(?) AN - OPUS4-26417 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Qiao, Linan A1 - Zencker, Uwe A1 - Völzke, Holger A1 - Wille, Frank A1 - Musolff, André ED - Topping, B.H.V. T1 - Validation of numerical simulation models for transport and storage casks using drop test results N2 - The safety assessment of new designs for transport and storage casks for radioactive materials is a challenging task accomplished using different methods such as prototype tests, model tests, calculations and analogy reflections. At BAM (Federal Institute for Materials Research and Testing), the test procedures for the mechanical IAEA (International Atomic Energy Agency) test conditions often start with preliminary finite element (FE) calculations mostly with a small-scale cask model for verification of the proposed test cask instrumentation and test plan. On that basis the extensive test cask instrumentation is applied and checked. After that, a series of drop tests consisting of different test sequences is performed. Following the drop tests, numerical post-analyses are carried out. These analyses offer the possibility of a detailed calculation and assessment of stresses and strains in the entire test cask construction. The calculation results have to be carefully compared with the measurement data over the impact history to find out all relevant parameters for a realistic simulation of the impact scenario. The desired ideal boundary test conditions often cannot be met exactly during the drop tests. Therefore, the numerical post-analyses are carried out by using the real boundary conditions of the drop tests. The objective is to find a validated model, where the results of the numerical simulations satisfactorily meet the experimental results. Under test conditions according to the IAEA transport regulations, casks are usually equipped with impact limiters and dropped onto a so-called unyielding target. In general, it is difficult to verify a complex FE model by using results from only one drop test because of the complex impact process and the complex structure of such packages. After each drop test, numerical post-analyses should be carried out. Only if all drop tests were simulated successfully by using the same FE model under different test conditions, it is possible to obtain a validated numerical model for further investigations. In this case the results of the numerical simulations meet satisfactorily the experimental results. In this paper a study is presented, where the influence of different components on the cask loading is investigated systematically. T2 - 11th International conference on computational structures technology CY - Dubrovnik, Croatia DA - 2012-09-04 KW - Impact KW - Simulation KW - Cask KW - Dop test KW - Finite element model KW - Validation KW - Dynamics KW - Transport and storage cask PY - 2012 DO - https://doi.org/10.4203/ccp.99.273 SN - 1759-3433 IS - Paper 273 SP - 1 EP - 12 PB - Civil Comp Press AN - OPUS4-26537 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Qiao, Linan A1 - Kasparek, Eva Maria A1 - Völzke, Holger T1 - Modeling of cylindrical casks under horizontal drop test conditions T2 - International High-Level Radioactive Waste Management Conference 2013 (IHLRWM 2013) CY - Albuquerque, NM, USA DA - 2013-04-28 PY - 2013 AN - OPUS4-28472 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Qiao, Linan A1 - Völzke, Holger T1 - Fracture mechanical analysis of a cylindrical cast iron cask N2 - The safety evaluation of cask components made of ductile cast iron includes investigations to prevent brittle fracture. Generally, ductile cast iron is endangered by brittle fracture especially at low temperatures (down to -40°C) and in combination with existing crack-like material defects. An applicable method is the assessment of fracture resistance using fracture mechanics according to the IAEA guidelines. The approach is based on the prevention of fracture initiation. For application of these principles for drop loads, account must be taken both of dynamic stresses within the component and dynamic material behavior. Basically, the dynamic stress intensity factor of postulated pre-existing crack-like defects is compared with the dynamic fracture toughness of the material. Applicable numerical and experimental methods for the safety assessment of cask components are demonstrated for the case of an artificially pre-cracked cylindrical cast iron cask which undergoes dynamic loading conditions as result of the hard impact between the cask and a concrete target. The proposed evaluation procedure is a combination of numerical and experimental steps. Exemplarily, the calculated stress intensity factor is compared with measured fracture toughness values from single edge notched bending specimens. T2 - 19th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 2019) CY - New Orleans, LA, USA DA - 04.08.2019 KW - Ductile Cast Iron KW - Brittle Fracture KW - Cylindrical Cask PY - 2019 AN - OPUS4-48915 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Simbruner, Kai A1 - Völzke, Holger T1 - Brittle failure of spent fuel claddings under long-term dry interim storage conditions – Preliminary analysis N2 - The evaluation of cladding integrity is a major issue to be demonstrated in Germany for extended interim storage periods up to 100 years and subsequent transportation considering operational and accidental conditions with respect to reactor operation, cask drying and dry interim storage. The chemical reaction between the zirconium fuel cladding and the cooling water in water-cooled reactors produces hydrogen and zirconium oxide. Hydrogen diffuses into the cladding and precipitates as zirconium hydrides when the solubility limit is reached, preferably oriented in hoop direction. At high temperatures during vacuum drying procedures, the hydrides can dissolve. Over a succeeding period of slow cooling with existing hoop stress the hydrides precipitate again, but partly reoriented along the radial direction of the cladding. This change of microstructure in combination with a decreasing temperature (0.5...2 K/year) during (extended) interim storage and additional mechanical load by handling procedures or under accident conditions could lead to a potential cladding embrittlement and consequently increased failure probability. The current research project BRUZL (Fracture mechanical analysis of spent fuel claddings under long-term dry interim storage conditions) has been launched by BAM to investigate potential sudden brittle failure of spent fuel claddings at small deformation under long-term dry interim storage conditions and is presented. T2 - IAEA International Conference on the Management of Spent Fuel from Nuclear Power Reactors CY - Vienna, Austria DA - 24.06.2019 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2019 AN - OPUS4-48868 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Simbruner, Kai A1 - Völzke, Holger T1 - Brittle failure of spent fuel claddings under long-term dry interim storage conditions – Preliminary analysis N2 - The evaluation of cladding integrity is a major issue to be demonstrated in Germany for extended interim storage periods up to 100 years and subsequent transportation considering operational and accidental conditions with respect to reactor operation, cask drying and dry interim storage. The chemical reaction between the zirconium fuel cladding and the cooling water in water-cooled reactors produces hydrogen and zirconium oxide. Hydrogen diffuses into the cladding and precipitates as zirconium hydrides when the solubility limit is reached, preferably oriented in hoop direction. At high temperatures during vacuum drying procedures, the hydrides can dissolve. Over a succeeding period of slow cooling with existing hoop stress the hydrides precipitate again, but partly reoriented along the radial direction of the cladding. This change of microstructure in combination with a decreasing temperature (0.5...2 K/year) during (extended) interim storage and additional mechanical load by handling procedures or under accident conditions could lead to a potential cladding embrittlement and consequently increased failure probability. The current research project BRUZL (Fracture mechanical analysis of spent fuel claddings under long-term dry interim storage conditions) has been launched by BAM to investigate potential sudden brittle failure of spent fuel claddings at small deformation under long-term dry interim storage conditions and is presented. T2 - IAEA International Conference on the Management of Spent Fuel from Nuclear Power Reactors CY - Vienna, Austria DA - 24.06.2019 KW - Ageing Management KW - Cladding Embrittlement KW - Extended Storage KW - Ring Compression Test KW - Spent Nuclear Fuel PY - 2019 SP - Paper IAEA-CN-272/49, 1 EP - 8 AN - OPUS4-48869 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Qiao, Linan A1 - Völzke, Holger T1 - Fracture mechanical analysis of a cylindrical cast iron cask N2 - The safety evaluation of cask components made of ductile cast iron includes investigations to prevent brittle fracture. Generally, ductile cast iron is endangered by brittle fracture especially at low temperatures (down to -40°C) and in combination with existing crack-like material defects. An applicable method is the assessment of fracture resistance using fracture mechanics according to the IAEA guidelines. The approach is based on the prevention of fracture initiation. For application of these principles for drop loads, account must be taken both of dynamic stresses within the component and dynamic material behavior. Basically, the dynamic stress intensity factor of postulated pre-existing crack-like defects is compared with the dynamic fracture toughness of the material. Applicable numerical and experimental methods for the safety assessment of cask components are demonstrated for the case of an artificially pre-cracked cylindrical cast iron cask which undergoes dynamic loading conditions as result of the hard impact between the cask and a concrete target. The proposed evaluation procedure is a combination of numerical and experimental steps. Exemplarily, the calculated stress intensity factor is compared with measured fracture toughness values from single edge notched bending specimens. T2 - 19th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 2019) CY - New Orleans, LA, USA DA - 04.08.2019 KW - Ductile Cast Iron KW - Brittle Fracture KW - Cylindrical Cask PY - 2019 SP - Paper 1209, 1 EP - 7 AN - OPUS4-48914 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Liu, Yung A1 - Völzke, Holger A1 - Zencker, Uwe A1 - Jaunich, Matthias T1 - Cooperation in Advanced Technologies Associated with the Back End of Nuclear Fuel Cycle N2 - A Memorandum of Understanding (MOU) between Argonne National Laboratory (Argonne) and the German Bundesanstalt für Materialforschung und -prüfung (BAM) was signed in October 2014. Its objectives are to promote cooperation among scientists and specialists at Argonne and BAM and establish a framework for collaboration in advanced technologies associated with the back end of the nuclear fuel cycle. Collaborative activities involving Argonne and BAM may be implemented through the promotion of joint research activities and scientific workshops and conferences; exchange of technical information; and visits by scientists, specialists, and graduate, postgraduate, and Ph.D. students. The MOU was renewed in January 2020 for another five years. Highlights of the progress and achievements are provided in identified topical areas for collaboration on ageing management guidance, storage, transportation and disposal R&D, joint conference activities, and conclude with the benefits of the MOU in fostering Argonne and BAM collaboration. T2 - 26th International Conference on Structural Mechanics in Reactor Technology - SMiRT 26 CY - Potsdam, Germany DA - 10.07.2022 KW - Nuclear fuel cycle KW - Back end KW - Spent fuel performance KW - Storage KW - Transportation PY - 2022 SP - 1 EP - 10 AN - OPUS4-55861 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Völzke, Holger A1 - Zencker, Uwe A1 - Herz, A. A1 - Kreienmeyer, M. T1 - Design of a Drop Test Target with Reproducible Properties for Konrad Prototype Testing N2 - Beginning in 1976 the former iron ore mine Konrad was geologically investigated concerning its suitability as a location for a German final repository for low and intermediate level waste. After a comprehensive licensing procedure it was approved and is now being prepared and equipped for the planned start of storage in 2027. At the end of the 1980s, the requirements for waste containers for storage in this repository were defined for the first time on the basis of an incident analysis of the on-site handling procedures, beginning with the unloading of the packages after the arrival at the site up to the final positioning in the underground storage galleries. A 5-meter drop onto the rock ground of the mine was identified as the covering case for high mechanical requirements (ABK II container class). In contrast to the 9-meter drop according to the IAEA Regulations for the Safe Transport of Radioactive Material, the 5-meter drop is performed onto a hard but not essentially unyielding target. The container is typically not protected by an impact limiter. The requirements for a potential test facility are described in the regulations for the Konrad repository. The mechanical strength of the target is defined as a concrete strength equivalent to the identified properties of the rock ground of the mine. Since 1991 BAM has consistently used precast reinforced concrete slabs as target for drop tests in the framework of licensing procedures as well as in research projects. While the original design fulfilled the requirement for the integrity of the concrete slab in most cases, it failed when drop tests onto an edge of a container were performed. A redesigned concrete slab developed in a research project and suggested as a reference target in 2009 has been successfully used in Konrad licensing procedures since then. The paper gives a brief overview of the historical development and it describes BAM’s efforts and approaches to continuously guarantee a concrete slab of defined quality and to provide a test setup for valid drop tests from the applicants and the authorities view. T2 - PATRAM 2022, 20th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Juan-les-Pins, France DA - 11.06.2023 KW - Target KW - Drop test KW - Final repository Konrad PY - 2023 SP - 1 EP - 11 AN - OPUS4-58563 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike 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 - Zencker, Uwe A1 - Gaddampally, Mohan Reddy A1 - Völzke, Holger T1 - Fracture Mechanics Analysis of Spent Fuel Claddings during Long-Term Dry Interim Storage N2 - The prevention of brittle fracture of spent fuel claddings during long-term dry interim storage is based on experimental investigations, numerical analyses and assessment methods for predicting the mechanical behavior and determining limiting conditions. The ring compression test (RCT) is an established experimental method for characterizing cladding material. Test results for various high-burnup pressure water reactor zirconium-based fuel cladding alloys (e.g., ZIRLO®, M5®) are publicly available. To reduce the effort associated with irradiated samples in hot cells, it is helpful to perform studies on unirradiated surrogate cladding material. Based on such experimental data, load-displacement curves were numerically analyzed for selected cladding materials. In the presence of radial hydrides, a sample may suddenly fail by fracture even at small deformations. Noticeable load drops in the RCT occur associated to unstable crack propagation through the radial hydride network. The failure mechanism is quasi-cleavage in the hydrides and micro-void nucleation, growth, and coalescence in the zirconium matrix, with ductile tearing patches connecting neighboring hydrides. The failure process was simulated by cohesive zones controlled by the fracture energy and the cohesive strength. A modeling approach is presented in which the radial hydride morphology is taken into account. Based on the developed fracture mechanics approach with cohesive zone modeling, not only the deformation behavior but also the failure behavior of irradiated as well as unirradiated claddings with radial hydrides under RCT loading conditions can be adequately described. T2 - 20th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 2022) CY - Juan-les-Pins, France DA - 11.06.2023 KW - Extended Interim Storage KW - Nuclear Fuel Cladding KW - Numerical Failure Analysis KW - Ring Compression Test KW - Spent Fuel Characterization PY - 2023 SP - 1 EP - 8 AN - OPUS4-59146 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Weber, Mike A1 - Qiao, Linan A1 - Droste, Bernhard A1 - Völzke, Holger T1 - Mechanical safety analyses of cast iron containers for the KONRAD repository N2 - Within the last years BAM has carried out numerous drop tests with prototype casks made of ductile cast iron onto targets according to the requirements for final disposal of non-heat generating waste in the German KONRAD repository. The results have shown that the target specifications in the acceptance criteria have to be defined more accurately to get reproducible test results with high precision. Hence, a suitable test stand foundation was developed with much effort. The integrity of the upper concrete layer of this target must be preserved during a test. Recently the geometrical properties of a tested cubic cast iron container led to a concentration of the impact forces beneath the container walls. The target was damaged strongly with the consequence of inadmissible reduction of cask stresses. For that reason the target construction was modified. However, the basic design was not changed. A prefabricated concrete slab was still joined by a mortar layer to the IAEA target of the BAM drop test facility. In the course of the optimization of the test stand foundation the concrete slab dimensions and the reinforcement were enlarged. During the drop test repetition the target kept intact. Additionally, the mechanical behavior of the cast iron container and the target was analyzed by finite element calculations. This improved target construction is suggested as a reference target for drop tests with casks whose mass and base area are covered by the container types VI or VII respectively according to the KONRAD repository acceptance criteria. The measurements during the drop tests with cast iron casks have provided the strains on the cask surface at selected positions. This allows the verification of finite element simulations of drop tests which show the stress distribution also inside the component. In September 2008 a drop test was carried out with a cylindrical cast iron cask containing an artificial material defect which was designed under consideration of critical stress states in the cask body. This drop test could demonstrate the safety against failure by fracture of a cask made of a special cast iron with reduced fracture toughness. T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Ductile iron KW - Container design KW - Safety assessment KW - Fracture mechanics KW - Dynamic loading conditions PY - 2010 SP - 1 EP - 8 (Session T28 / Paper 220) AN - OPUS4-23844 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Bletzer, Claus Wilhelm A1 - Zencker, Uwe A1 - Völzke, Holger A1 - Koch, Frank T1 - Asymmetrische Temperaturprofile in Transport- und Lagerbehältern für radioaktive Stoffe T2 - Jahrestagung Kerntechnik CY - Dresden, Deutschland DA - 2009-05-12 KW - Temperaturprofil KW - CFD-Berechnung KW - Transportbehälter KW - Radioaktive Stoffe PY - 2009 SP - 1 EP - 5(?) CY - Dresden AN - OPUS4-19431 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Jaunich, Matthias A1 - Weber, Mike A1 - Kömmling, Anja A1 - Zencker, Uwe A1 - Wolff, Dietmar A1 - Völzke, Holger T1 - Performance of elastomer seals in transport and storage casks N2 - Elastomer seals are widely used as barrier seals in containers for low and intermediate level radioactive waste and for spent fuel transportation casks. In addition, they are also used for spent fuel storage and transportation casks (dual purpose casks (DPC)) as auxiliary seals to allow leakage rate measurements of metal barrier seals for demonstration of their proper assembling conditions. Depending on the area of use, the rubber materials have to demonstrate proper sealing performance with regard to mechanical, thermal, and environmental conditions as well as irradiation during the entire operation period. Concerning DPC, degradation effects should be limited in a way that, for example, effects from potentially released decomposition elements may not harm e.g. metal barrier seals. Leakage rate measurements should be possible also after long interim storage periods prior to subsequent transportation. Because of the complex requirements resulting from the various applications of containers for radioactive waste and spent nuclear fuel, BAM has initiated several test programmes for investigating the behaviour of elastomer seals. In this contribution the current status is described and first results are discussed. T2 - International conference on management of spent fuel from nuclear power reactors: An integrated approach to the back end of the fuel cycle CY - Vienna, Austria DA - 15.06.2015 KW - Ageing KW - Elastomer KW - Glass-rubber transition KW - Irradiation KW - Material model PY - 2015 SP - 1 EP - 8 AN - OPUS4-33553 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Zencker, Uwe A1 - Wolff, Dietmar A1 - Jaunich, Matthias A1 - Kömmling, Anja A1 - Völzke, Holger A1 - Schulze, Dietmar T1 - Numerical simulation of the behaviour of elastomer seals under consideration of time dependent effects N2 - Due to delays in the site-selection procedure to establish a deep geological repository for spent nuclear fuel and high level waste and in construction of the already licensed Konrad-repository for low and intermediate level waste without heat generation, extended periods of interim storage become more relevant in Germany. BAM is involved in most of the cask licensing procedures and especially responsible for the evaluation of cask-related long-term safety issues. The long-term performance of elastomer seals for lid Systems of transport and storage casks whether used as auxiliary seals in spent fuel caslcs or as primary seals for low and intermediate level waste packages is an important issue in this context (Jaunich, 2013; Jaunich, 2014; Kömmling, 2015). The polymeric structure of these seals causes a complex mechanical behaviour with time-dependent elasticity reduction and loss of elastic recovery. The paper presents first results of a comprehensive test Programme consisting of several static and dynamic mechanical short- and long-term tests which have been carried out at BAM on specimens made of representative types of elastomers, fluorocarbon rubber (FKM) and ethylene propylene diene rubber (EPDM). The investigation of the test results used to identify material models and their Parameters as well as the development of two finite element models for the numerical Simulation of tension and compression tests using the finite element code ABAQUS® are described. The calculation results are presented in comparison to the test results. The influence of important material and test parameters was investigated and discussed in sensitivity analyses. T2 - RAMTRANSPORT 2015 - International conference on the radioactive materials transport and storage CY - Oxford, UK DA - 19.05.2015 KW - Numerical analysis KW - Extended interim storage KW - Elastomer seals PY - 2015 SP - Session 4, Paper 28, 1 EP - 9 AN - OPUS4-33832 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Qiao, Linan A1 - Zencker, Uwe A1 - Kasparek, Eva Maria A1 - Völzke, Holger ED - Topping, B.H.V. ED - Iványi, P. T1 - Simulation of damping concrete under severe impact loads using a crushable foam model with damage mechanisms N2 - The deformation and damage behaviour of damping concrete under impact loading conditions is investigated experimentally and numerically. The material model is based on the assumption of crushable foam with volumetric or isotropic hardening combined with ductile and shear damage criteria. Model parameters are determined in static and dynamic compression tests of confined cubic specimens. The derived material model is validated by numerical simulation of penetration tests. The static and dynamic penetration of indenters into uniform as well as assembled bricks made of damping concrete is discussed. Finally, the successful calculation of a large-scale drop test with a heavy cask-like test object onto a realistic damping concrete foundation is demonstrated. T2 - 12th International conference on computational structures technology CY - Naples, Italy DA - 2014-09-02 KW - Damping concrete KW - Impact load KW - Crushable foam material model KW - Damage PY - 2014 SN - 978-1-905088-61-4 DO - https://doi.org/10.4203/ccp.106.233 SN - 1759-3433 SP - Paper 233, 1 EP - 16 PB - Civil Comp Press AN - OPUS4-31480 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kasparek, Eva Maria A1 - Völzke, Holger A1 - Scheidemann, Robert A1 - Zencker, Uwe A1 - Wolff, Dietmar T1 - Experimentelle und numerische Untersuchungen von stoßdämpfenden Werkstoffen für Behälter zur Entsorgung radioaktiver Stoffe N2 - Die mechanische Beanspruchung von Transport- und Lagerbehältern für radioaktive Stoffe gemäß den durch die IAEA vorgegebenen Prüfszenarien oder bei Unfällen, die während der Handhabung in einer kerntechnischen Anlage auftreten können, wird maßgeblich durch das Design der Stoßdämpfer und die Eigenschaften des jeweiligen Untergrundes beeinflusst. Da der Nachweis der Behälterintegrität und -dichtheit häufig an numerische Simulationen gekoppelt ist, sind für die sicherheitstechnische Auslegung der Konstruktion zuverlässige Informationen über das Energieabsorptionsvermögen der Werkstoffe und eine geeignete Umsetzung dieser Daten in FEM - Programme erforderlich. Die im Rahmen von Falluntersuchungen bei der BAM festgestellten, teilweise erheblichen Abweichungen zwischen experimentellen und berechneten Größen sind entsprechend darauf zurückzuführen, dass die zur Verfügung stehenden numerischen Verfahren das komplexe Verhalten der stoßdämpfenden Strukturen nicht ausreichend präzise wiedergeben. Hiervon besonders betroffen sind Bauteile aus Holz und Polyurethan-Schaum (PU-Schaum), die zur Konstruktion von Behälterstoßdämpfern für die Transportkonfiguration eingesetzt werden, sowie Dämpferbeton, der in ausgewählten Bereichen im Fundament von Standort-Zwischenlagern eingebaut wird. Um die zugehörigen Materialmodelle weiter zu entwickeln und die dafür erforderliche Datenbasis zu schaffen, hat die BAM vor knapp zwei Jahren das vom BMBF geförderte Forschungsvorhaben ENREA (Entwicklung rechnerischer Analysemethoden für stoßdämpfende Strukturen beim Anprall oder Absturz von Abfallgebinden) gestartet. Wesentlicher Bestandteil ist ein systematisches Versuchsprogramm, bei dem die drei genannten Werkstoffe statisch und dynamisch hohen Druckbelastungen ausgesetzt werden. Relevante Versuchsparameter sind, neben der Dehnrate, die Temperatur und Lagerungsbedingungen sowie die Orientierung des Werkstoffes hinsichtlich der Belastungsrichtung. Dieser Beitrag präsentiert die wichtigsten Ergebnisse aus der ersten Versuchsphase, die durch weggesteuerte Versuche mit konstanten Prüfgeschwindigkeiten von 0,02 mm/s und 3000 mm/s gekennzeichnet ist. Darüber hinaus wird die Verwertung dieser Daten zur Anpassung von Materialmodellen auf der Grundlage von Nachrechnungen gezeigt, wobei die Möglichkeiten und Grenzen der verwendeten Konzepte Umrissen werden. Aufgrund des Umfangs, den die Versuche mit Holzwerkstoffen einnehmen, und des erforderlichen Aufwandes für die numerische Modellierung werden die entsprechenden Maßnahmen und Ergebnisse in einem eigenständigen Beitrag auf der KONTEC vorgestellt. T2 - KONTEC 2011 - 10. Internationales Symposium "Konditionierung radioaktiver Betriebs- und Stilllegungsabfälle" CY - Dresden, Germany DA - 06.04.2011 KW - Stoßdämpfende Werkstoffe KW - Materialmodellierung KW - Versuchsprogramm KW - Stoßbelastung KW - Dehnratenabhängigkeit KW - Temperaturabhängigkeit KW - Polyurethan Schaum KW - Dämpferbeton PY - 2011 SP - 453 EP - 458 AN - OPUS4-23709 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Weber, Mike A1 - Qiao, Linan A1 - Droste, Bernhard A1 - Völzke, Holger T1 - Entwicklung von Beurteilungsmethoden für Transport- und Lagerbehälter mit erhöhten metallischen Reststoffanteilen (EBER III) T2 - 9. Intern. Symp. "Konditionierung radioaktiver Betriebs- und Stilllegungsabfälle" - KONTEC CY - Dresden, Germany DA - 2009-04-15 PY - 2009 AN - OPUS4-19251 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Weber, Mike A1 - Qiao, Linan A1 - Völzke, Holger A1 - Droste, Bernhard T1 - Safety analyses of cast iron containers hitting a KONRAD reference target T2 - Radioactive Materials Transport Conference CY - Manchester, England DA - 2009-05-13 PY - 2009 AN - OPUS4-19467 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Qiao, Linan A1 - Wieser, Günter A1 - Völzke, Holger T1 - Numerical safety assessment of a transport and storage cask for radioactive materials without impact limiters by the 0.3 m drop test onto an unyielding target T2 - 9th International Conference on Computational Structures Technology CST 2008 CY - Athens, Greece DA - 2008-09-02 PY - 2008 AN - OPUS4-17939 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Weber, Mike A1 - Kasparek, Eva Maria A1 - Völzke, Holger T1 - Reliability of Cask Designs under Mechanical Loads in Storage Facilities T2 - 12th International Probabilistic Safety Assessment and Management Conference (PSAM12) CY - Honolulu, HI, USA DA - 2014-06-22 PY - 2014 AN - OPUS4-31108 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Qiao, Linan A1 - Völzke, Holger T1 - Strategies for Numerical Modelling of Metal Gaskets in Transport and Storage T2 - Radioactive Materials Transport and Storage Conference and Exhibition (RAMTRANSPORT 2015) CY - Oxford, United Kingdom DA - 2015-05-19 PY - 2015 AN - OPUS4-33398 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Qiao, Linan A1 - Zencker, Uwe A1 - Völzke, Holger ED - Brandt, A. M. ED - Olek, J. ED - Glinicki, M. A. ED - Leung, C.K.Y. ED - Lis, J. T1 - Crushable foam model with pressure and stress-triaxiality dependent damage mechanisms for damping concrete N2 - Damping concrete with high energy absorption capability consists of a cement matrix with embedded small polystyrene balls and is investigated experimentally and numerically under high static compression with and without clamping. A material model is derived which describes the fundamental effects of damping concrete like non-linear elastic-plastic behaviour, volume change, volume strain dependent hardening as well as shear failure with an adequate description of damage initiation and evolution. The suggested material model is validated by simulation of penetration tests. T2 - BMC-11, International Symposium on Brittle Matrix Composites CY - Warsaw, Poland DA - 28.09.2015 KW - Damping concrete KW - Impact load KW - Crushable foam material model KW - Damage PY - 2015 SN - 978-83-89687-96-8 SP - 269 EP - 277 PB - Inst Fundamental Technological Research, Polish Acad Sciences CY - Warschau AN - OPUS4-34613 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Qiao, Linan A1 - Völzke, Holger T1 - Dynamic finite element analysis of cask handling accidents at storage sites N2 - The safety assessment of casks for radioactive material at interim storage facilities or in final repositories includes the investigation of possible handling accidents if clearly defined test conditions are not available from the regulations. Specific handling accidents usually are the drop of a cask onto the transport vehicle or the floor as well as the collision with the wall of the storage building or another cask. For such load cases an experimental demonstration of cask safety would be difficult. Therefore, numerical analyses of the entire load scenario are preferred. The lessons learnt from dynamic finite element analyses of accident scenarios with thick-walled cubical containers or cylindrical casks are presented. The dependency of calculation results on initial and boundary conditions, material models, and contact conditions is discussed. Parameter sets used should be verified by numerical simulation of experimentally investigated similar test scenarios. On the other hand, decisions have to be made whether a parameter or property is modeled in a realistic or conservative manner. For example, a very small variation of the initial impact angle of a container can cause significantly different stresses and strains. In sophisticated cases an investigation of simpler limit load scenarios could be advantageous instead of analyzing a very complicated load scenario. T2 - ASME 2015 Pressure vessels & piping conference - PVP2015 CY - Boston, Massachusetts, USA DA - 19.07.2015 KW - Finite element method KW - Simulation KW - Cask handling accident KW - Interim storage PY - 2015 SN - 978-0-7918-5702-1 SP - Paper 45606, 1 EP - 10 AN - OPUS4-34937 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Qiao, Linan A1 - Völzke, Holger T1 - Strategies for numerical modelling of metal gaskets in transport and storage casks N2 - Metal gaskets in the lid system of transport and storage casks for radioactive material have to guarantee leak-tightness and safe enclosure of the radioactive inventory under normal and accident conditions during transport, in case of accidents and for the longterm interim storage. For safety assessments by nurnerical simulation of the thermomechanical behaviour of a lid system with metal gaskets, the finite element method offers three options to use: i) special gasket elements, ii) complex three-dimensional modeHing with solid elements, and iii) a simplified axisymmetric approach. Gasket elements can be adjusted by many parameters, but they give only a global representation of the experimentally observed gasket behaviour. For calculations of the entire cask including the lid system with gaskets, nonetheless this approach is recommended. To investigate the hardly measurable impact behaviour in detail or to extrapolate the long-term behaviour, a local modeHing of all parts of a gasket and their interactions with effects like elastic-plastic deformation, creep, relaxation, and friction is necessary. The three-dimensional model can describe the change of contact area between outer jacket of the gasket and flange dependent on the load conditions, what overcomes an essential limitation of special gasket elements. To simplify the problern of investigating the underlying physical effects, an axisymmetric lid system can be modelled with axisymmetric finite elements. Usually an approximation is only necessary for the helical spring of a metal gasket. This paper explains basic ideas for an adequate finite element simulation of cask lid systems with metal gaskets and their thermo-mechanical behaviour under specific load scenarios. T2 - RAMTRANSPORT 2015 - International conference on the radioactive materials transport and storage CY - Oxford, UK DA - 19.05.2015 KW - Metal gasket KW - Finite element method KW - Simulation PY - 2015 SP - Paper RAM15-27, 1-9 AN - OPUS4-34938 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Qiao, Linan A1 - Völzke, Holger T1 - Dynamic Finite Element Analysis of Cask Handling Accidents at Storage Sites T2 - ASME 2015 Pressure Vessels & Piping Conference (PVP2015) CY - Boston, MA, USA DA - 2015-07-19 PY - 2015 AN - OPUS4-34856 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Qiao, Linan A1 - Völzke, Holger T1 - Influence of impact angle and real target properties on drop test results of cubic containers N2 - Drop test scenarios with cubic containers without impact limiters at interim storage sites or in a final repository have been investigated by numerical simulations. An ideally flat drop is impossible to conduct as a free fall of a container even under laboratory conditions. Dynamic stresses and strains inside the container structure are sensitive to the impact angle. Even very small impact angles cause remarkable changes in the experimental or numerical results when a flat bottom or wall of a container hits a flat target. For drop tests with transport packages the International Atomic Energy Agency (IAEA) regulations define an essentially unyielding target. In contrast, potential accident scenarios for storage containers are derived from site-specific safety analyses or acceptance criteria in Germany. Each interim storage site or repository has a yielding or so-called real target with individual structural and material properties. The real target acts as a kind of impact limiter. A more conservative container design is required if the impact limiting effect of the target is not considered. T2 - ASME 2017 Pressure Vessels and Piping Conference CY - Waikoloa, Hawaii, USA DA - 16.07.2017 KW - Finite element method KW - Simulation KW - Drop test KW - Impact angle KW - Real target KW - Yielding target PY - 2017 SN - 978-0-7918-5802-8 DO - https://doi.org/10.1115/PVP2017-65731 VL - 7 SP - Article UNSP V007T07A039, 1 EP - 9 AN - OPUS4-43631 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - RPRT A1 - Zencker, Uwe A1 - Simbruner, Kai A1 - Völzke, Holger T1 - Entwicklung eines bruchmechanischen Berechnungsansatzes zur Beschreibung des Festigkeitsverhaltens von Brennstabhüllrohren bei längerfristiger Zwischenlagerung (BRUZL) N2 - Es werden Methoden zur sicherheitstechnischen Bewertung des mechanischen Verhaltens von Brennstabhüllrohren entwickelt, um sprödes Versagen in Belastungsszenarien nach längerfristiger trockener Zwischenlagerung zu verhindern. Unbestrahlte Brennstabhüllrohrsegmente aus der Zirkoniumlegierung ZIRLO® wurden hydriert und thermomechanischen Behandlungen unterzogen, um Hüllrohrmaterial durch Ausscheidung radialer Hydride zu verspröden. Die Proben mit radialen Hydriden wurden im Ring Compression Test (RCT) bei Raumtemperatur untersucht. Bereits bei kleinen Probenverformungen wurden abrupte Kraftabfälle mit instabiler Rissausbreitung durch das Netzwerk radialer Hydride beobachtet. Die Risslängen erreichten in einigen Fällen bis zu 90 % der Wanddicke. Der mikromechanische Versagensmechanismus ist Quasi-Spaltbruch in den Hydriden und Porenbildung, -wachstum und -vereinigung in der Zirkoniummatrix an den duktilen Brücken zwischen benachbarten Hydriden. Es wurde ein Finite-Elemente-Modell des Ring Compression Tests erstellt, um das Versagensverhalten zirkoniumbasierter Legierungen mit radialen Hydriden im RCT zu analysieren. Die Fließkurve für das elastisch-plastische Materialmodell des Matrixmaterials wurde durch eine inverse Finite-Elemente-Analyse ermittelt. Kohäsivzonenmodelle wurden benutzt, um die plötzlichen Kraftabfälle zu reproduzieren. Es wurde ein mikromechanisches Modell mit statistisch generierten Matrix-Hydrid-Verteilungen entwickelt, das zwischen sprödem Versagen entlang der Hydride und duktilem Versagen der Zirkoniummatrix in der Kohäsivzone unterscheidet. Für das jeweilige Materialverhalten wird ein lokales zweiparametriges Versagenskriterium auf Grundlage der Kohäsionsfestigkeit und der Separationsenergie bei einer festgelegten Form des Kohäsivgesetzes vorgeschlagen. Geeignete Kohäsivparameter lassen sich mit einer inversen Finite-Elemente-Analyse des Versagensverhaltens von hydrierten Proben im Ring Compression Test bestimmen. In den Simulationen lag der Schwerpunkt auf dem ersten Kraftabfall. Die Berechnungsergebnisse stimmen gut mit den RCT-Ergebnissen überein. Es konnte gezeigt werden, dass die Rissinitiierung und -ausbreitung stark von der speziellen Anordnung der radialen Hydride und Matrixbrücken in der Bruchzone abhängt. Die numerische Modellbildung wurde an Versuchen mit bestrahlten Proben aus der Zirkoniumlegierung M5® validiert. KW - Zwischenlagerung KW - Brennstabhüllrohr KW - Versagensmechanismus KW - Kohäsivzonenmodell KW - Ring Compression Test PY - 2022 SP - 1 EP - 106 PB - Bundesanstalt für Materialforschung und -prüfung (BAM) CY - Berlin AN - OPUS4-56853 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - RPRT A1 - Caruso, Stefano A1 - Vlassopoulos, Efstathios A1 - Dagan, Ron A1 - Fiorito, Luca A1 - Herm, Michel A1 - Jansson, Peter A1 - Kromar, Marjan A1 - Király, Márton A1 - Leppanen, Jaakko A1 - Feria Marquez, Francisco A1 - Metz, Volker A1 - Papaioannou, Dimitrios A1 - Herranz, Luis Enrique A1 - Rochman, Dimitri A1 - Schillebeeckx, Peter A1 - Seidl, Marcus A1 - Hernandez Solis, Augusto A1 - Stankovskiy, Alexey A1 - Alvarez Velarde, Francisco A1 - Verwerft, Marc A1 - Rodriguez Villagra, Nieves A1 - Zencker, Uwe A1 - Žerovnik, Gasper T1 - EURAD - Work Package 8 - Deliverable 8.1 - State-of-the-art report N2 - A state-of-the-art (SOTA) review on characterisation of spent nuclear fuel (SNF) properties in terms of source term and inventory assessment (neutron, gamma-ray emission, decay heat, radionuclide inventory, elemental content) and in terms of out-of-core fuel performance (cladding performance and fuel integrity in view of the safety criteria for SNF interim storage, transport and canister packaging) using several numerical and experimental approaches and methodologies is presented. This SOTA report is a result of the spent fuel characterisation (SFC) work package as part of the European Joint Programme on Radioactive Waste Management (EURAD), which offers an overview of the status of knowledge in the field of SNF characterisation and assessment during the pre-disposal phase. The document aims to focus on the current safety-significant gaps and related challenges, providing a direct link to the goals of the mandated actors of EURAD. The report is expected to be used by all EURAD colleagues in their national programmes as a key resource for knowledge management programmes and to contribute to demonstrating and documenting the state-of-the-art. KW - Radioactive waste management KW - Spent fuel characterisation KW - Extended interim storage KW - Predisposal PY - 2022 UR - https://www.ejp-eurad.eu/publications/eurad-d81-state-art-report SP - 1 EP - 112 PB - Agence Nationale pour la Gestion des Déchets Radioactifs (ANDRA) CY - Châtenay-Malabry AN - OPUS4-59154 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - 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 - 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 - JOUR A1 - Sjöland, A. A1 - Christensen, P. A1 - Zetterström Evins, L. A1 - Bosbach, D. A1 - Duro, L. A1 - Farnan, I. A1 - Metz, V. A1 - Zencker, Uwe A1 - Ruiz-Hervias, J. A1 - Rodriguez-Villagra, N. A1 - Kiraly, M. A1 - Schillebeeckx, P. A1 - Rochman, D. A1 - Seidl, M. A1 - Dagan, R. A1 - Verwerft, M. A1 - Herranz Puebla, L. E. A1 - Hordynskyi, D. A1 - Feria, F. A1 - Vlassopoulos, E. T1 - Spent nuclear fuel management, characterisation, and dissolution behaviour: progress and achievement from SFC and DisCo N2 - SFC is a work package in Eurad that investigates issues related to the properties of the spent nuclear fuel in the back-end of the nuclear fuel cycle. Decay heat, nuclide inventory, and fuel integrity (mechanical and otherwise), and not least the related uncertainties, are among the primary focal points of SFC. These have very significant importance for the safety and operational aspect of the back-end. One consequence is the operation economy of the back-end, where deeper understanding and quantification allow for significant optimization, meaning that significant parts of the costs can be reduced. In this paper, SFC is described, and examples of results are presented at about half-time of the work package, which will finish in 2024. The DisCo project started in 2017 and finished in November 2021 and was funded under the Horizon 2020 Euratom program. It investigated if the properties of modern fuel types, namely doped fuel, and MOX, cause any significant difference in the dissolution behavior of the fuel matrix compared with standard fuels. Spent nuclear fuel experiments were complemented with studies on model materials as well as the development of models describing the solid state, the dissolution process, and reactive transport in the near field. This research has improved the understanding of processes occurring at the interface between spent nuclear fuel and aqueous solution, such as redox reactions. Overall, the results show that from a long-term fuel matrix dissolution point of view, there is no significant difference between MOX fuel, Cr+Al-doped fuel, and standard fuels. KW - Spent nuclear fuel management KW - Spent fuel characterization KW - Dissolution behaviour PY - 2023 UR - https://nbn-resolving.org/urn:nbn:de:kobv:b43-572461 DO - https://doi.org/10.1051/epjn/2022029 SN - 2491-9292 VL - 9 SP - 1 EP - 12 PB - EDP Sciences CY - Les Ulis Cedex, France AN - OPUS4-57246 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Völzke, Holger 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 SP - 1 EP - 10 AN - OPUS4-60323 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Völzke, Holger 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 -