TY - CONF A1 - Neumann, Martin A1 - Wille, Frank A1 - Droste, Bernhard T1 - Introduction of BAM safety assessment experience feedback list N2 - BAM as competent authority regarding assessment of mechanical and thermal design, activity release and quality assurance aspects of SNF and HLW transport packages developed a thesaurus of experience feedback topics from recent approval procedures. The list is structured according to the European PDSR guide. It involves issues, which from BAM point of view, needed clarification during last package design assessment procedures. The list contains issues from operation (e.g. deactivation of handling lugs not intended for package handling), maintenance (e.g. leak tightness of impact limiting devices) to technical assessment (e.g. formation of hydrogen by radiolysis and its impact on pressure and ignitability, consideration of ageing mechanisms, thermo-mechanical assessment, impact of gaps between content and flask, spent fuel behaviour) as well as to general and specific safety analysis report requirements (e.g. report structure and required data). T2 - PATRAM 2016 CY - Kobe, Japan DA - 18.09.2016 KW - Experience feedback list KW - Radioactive material KW - Transport PY - 2016 UR - http://www.patram2016.org/ SP - Paper 4002, 1 AN - OPUS4-37850 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin A1 - Nehrig, Marko A1 - Bletzer, Claus A1 - Musolff, André A1 - Erenberg, Marina A1 - Wille, Frank T1 - Behaviour of Wood Filled Impact Limiters during Fire Test N2 - Packages for the transport of radioactive material are often equipped with impact limiters consisting of wood, encapsulated by steel sheets. These impact limiters shall ensure the transport cask meets the mechanical and thermal IAEA regulatory test requirements. After damage caused by the mechanical tests the package has to withstand a severe fire scenario. According to the regulations during and following the thermal test, the specimen shall not be artificially cooled and any combustion of materials of the package shall be permitted to proceed naturally. Corresponding to results of the French institute IRSN combustion and smouldering of wood inside the impact limiter occurred during and after the fire test. An additional energy supply from a pre-damaged impact limiter to the cask could be the consequence for the safety assessment of the containment. BAM started a first test phase to examine the issue of combustion for such kind of package components. The goal was to understand the phenomena under the consideration of relevant regulatory boundary conditions. Several metal buckets were filled with wood and equipped with thermocouples. The test specimens have been prepared with different damage arrangements to take the influence of the mechanical tests into account. This paper shows the experimental setup and the conduction of the tests. The first test shows that pre-damaged metal encapsulations can lead to smouldering of the wood and with this to a supplement energy release after the end of the 30 minute fire. BAM is in the preparation process for a second test phase. A thermal test will take place with a wood filled test specimen weighing about 2Mg. T2 - The 18th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 2016) CY - Kobe, Japan DA - 18.09.2016 KW - Large scale testing KW - Fire test KW - Impact limiter KW - Shock absorber KW - Thermal test KW - Typ-B KW - Wood KW - Smouldering PY - 2016 SP - paper no. 1011 AN - OPUS4-37854 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Komann, Steffen A1 - Wille, Frank T1 - Zur Prüfung verpflichtet - Die BAM-Gefahrgutregel 016 (BAM-GGR 016) N2 - Im Gegensatz zu den zulassungspflichtigen Versandstücken für die gemäß der gefahrgutrechtlichen Regelwerksanforderungen eine behördlich ausgestellte Zulassung erforderlich ist, ist für die prüfpflichtigen Versandstücke lediglich eine behördliche Anerkennung und Überwachung des Managementsystems für die Auslegung, Herstellung, Prüfung, Dokumentation, den Gebrauch, die Wartung und Inspektion erforderlich. In der Bundesrepublik Deutschland ist gemäß den Festlegungen in der Gefahrgutverordnung Straße, Eisenbahn und Binnenschifffahrt die Bundesanstalt für Materialforschung und -prüfung (BAM) zuständig für die Anerkennung und Überwachung von Managementsystemen. KW - Gefahrgut KW - Radioaktive Stoffe KW - Normen KW - Regeln PY - 2018 SN - 0944-6117 VL - 2018 IS - 1-2 SP - 10 EP - 12 PB - Springer Fachmedien München GmbH CY - München AN - OPUS4-44168 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Gröke, Carsten A1 - Wille, Frank T1 - Assessment of quality management for transport packages not requiring authority design approval N2 - The majority of transports of radioactive materials are carried out in packages which don’t need a package design approval of a competent authority. Low active radioactive materials are transported in such kind of packages e.g. in the medical and pharmaceutical industry and in the nuclear industry as well. In Germany the decision to phase out nuclear energy leads to a strong demand for packages to transport low and middle active radioactive waste due to the dismantling and decommissioning of nuclear power plants. According to IAEA regulations the “non-competent authority approved package types” are the excepted packages and the industrial packages of Type IP-1, IP-2 and IP-3 and of Type A. For the packages of Type IP-2, IP-3 and Type A an assessment by the German competent authority is required for the quality management for the design, manufacture, testing, documentation, use, maintenance and inspection. In general a compliance audit of the manufacturer of the packaging is required during this assessment procedure. T2 - Waste Management Conference 2018 CY - Phoenix, USA DA - 18.3.2018 KW - Radioactive material KW - Quality management system KW - Transport packages PY - 2018 SP - Article 18194, 1 EP - 7 AN - OPUS4-44170 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin A1 - Erenberg, Marina A1 - Nehrig, Marko A1 - Bletzer, Claus A1 - Musolff, André A1 - Schönfelder, Thorsten A1 - Wille, Frank T1 - Behavior of wood filled impact limiters during the IAEA thermal test N2 - Packages for the transport of radioactive material are often equipped with impact limiters consisting of wood, encapsulated by steel sheets. These impact limiters shall ensure that transport casks meet the IAEA safety requirements. After damage caused by the mechanical tests the package has to withstand a severe fire scenario. It is required that the mechanical tests have to produce maximum damage, taking into account the thermal test. Furthermore, any damage, which would give rise to increased radiation or loss of containment or affect the confinement system after the thermal test, should be considered. Concerning the thermal test, the IAEA safety requirements state that during and following the fire test, the specimen shall not be artificially cooled and any combustion of materials of the package shall be permitted to proceed naturally. Different works from the French Institute for Radiological Protection and Nuclear Safety (IRSN) and BAM show that additional energy supply from a pre-damaged impact limiter to the cask could occur caused by smoldering of the wood. This effect should be considered within the safety assessment of the package. A heat wave from the fire could overlap with the additional energy from the impact limiter in the sealing system. In 2015 BAM conducted small scale fire tests with wood filled metal drums showing continuing combustion processes during the cooling down phase. As not much is known about smoldering processes in wood filled impact limiters, it is highly complex to define pre-damage of impact limiters, which are conservative, regarding the most damaging energy flow from the impact limiter to the containment system in dependence of time. More research has to be done to develop models to examine the effects of smoldering impact limiters on the containment of packages for the assessment. The process of smoldering is described with regard to the requirements in the thermal safety assessment. Parameters influencing the smoldering process are identified. BAM operates test facilities to examine the issue of mechanical damage, combustion and heat transfer of packages for transport of radioactive material. A thermal test will take place with a wood filled test specimen with a diameter of about 2.3 meters. The aim is to understand the phenomena of smoldering under the consideration of relevant regulatory boundary conditions. T2 - ASME 2017 Pressure Vessels and Piping Conference (PVP2017) CY - Waikoloa, Hawaii, USA DA - 16.07.2017 KW - Fire test KW - Wood KW - IAEA KW - Smoldering KW - Smouldering KW - Shock absorber KW - Thermal test KW - Combustion KW - Impact limiter PY - 2017 SN - 978-0-7918-5802-8 VL - 7 SP - Article UNSP V007T07A038, 1 EP - 9 AN - OPUS4-43172 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Linnemann, Konrad A1 - Ballheimer, Viktor A1 - Rolle, Annette A1 - Wille, Frank T1 - Considerations on spent fuel behavior for transport after extended storage N2 - Packages for the transport of spent nuclear fuel shall meet the International Atomic Energy Agency regulations to ensure safety under different transport conditions. The physical state of spent fuel and the fuel rod cladding as well as the geometric configuration of fuel assemblies are important inputs for the evaluation of package capabilities under these conditions. In this paper, the mechanical behavior of high burn-up spent fuel assemblies (> approx. 50 GWd/tHM, value averaged over the fuel assembly) under transport conditions is analyzed with regard to the assumptions which are used in the Containment and criticality safety analysis. In view of the complexity of the interactions between the fuel rods as well as between the fuel assemblies, basket, and cask containment, the exact mechanical analysis of such phenomena is nearly impossible. Additionally, the gaps in information concerning the material properties of cladding and pellet behavior, especially for the high burn-up fuel, make the analysis more complicated. Considerations and knowledge gaps for the transport after extended interim storage are issues of growing interest. In this context, practical approaches are discussed based on the experience of BAM within the safety assessment of packages approved for transport of spent nuclear fuel. KW - Transport packages for radioactive material KW - Spent nuclear fuel PY - 2018 VL - 83 IS - 6 SP - 488 EP - 494 PB - Hanser AN - OPUS4-47359 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Erenberg, Marina A1 - Bletzer, Claus A1 - Feldkamp, Martin A1 - Musolff, André A1 - Nehrig, Marko A1 - Wille, Frank T1 - Experimental investigations of the burning behaviour of transport package impact limiters and of fire spread impact onto the cask N2 - Accident safe packages for the transport of spent nuclear fuel and high-level waste shall fulfil international IAEA safety requirements. Compliance is shown by consecutive mechanical and thermal testing. Additional numerical analysis are usually part of the safety evaluation. For damage protection some package designs are equipped with wood filled impact limiters encapsulated by steel sheets. The safety of these packages is established in compliance with IAEA regulations. Cumulative mechanical and fire tests are conducted to achieve safety standards and to prevent loss of containment. Mechanical reliability is proven by drop tests. Drop testing might cause significant damage of the impact limiter steel sheets and might enable sufficient oxygen supply to the impact limiter during the fire test to ignite the wood filling. The boundary conditions of the fire test are precisely described in the IAEA regulatory. During the test the impact limiter will be subjected to a 30 minute enduring fire phase. Subsequent to the fire phase any burning of the specimen has to extinguish naturally and no artificial cooling is allowed. At BAM a large-scale fire test with a real size impact limiter and a wood volume of about 3m3 was conducted to investigate the burning behaviour of wood filled impact limiters in steel sheet encapsulation. The impact limiter was equipped with extensive temperature monitoring equipment. Until today burning of such impact limiters is not sufficiently considered in transport package design and more investigation is necessary to explore the consequences of the impacting fire. The objective of the large scale test was to find out whether a self-sustaining smouldering or even a flaming fire inside the impact limiter was initiated and what impact on the cask is resulting. The amount of energy, transferred from the impact limiter into the cask is of particular importance for the safety of heavy weight packages. With the intention of heat flux quantification a new approach was made and a test bench was designed. T2 - ASME 2018 Pressure Vessels and Piping Conference CY - Prague, Czech Republic DA - 15.07.2018 KW - Shock absorber KW - Impact limiter KW - Wood KW - Thermal testing KW - Fire KW - Smoldering KW - IAEA KW - Fire test PY - 2018 SN - 978-0-7918-5170-8 VL - PVP2018 SP - 84714-1 EP - 84714-10 AN - OPUS4-46984 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Spent fuel behavior under Transport Conditions N2 - German packages for the transport of spent nuclear fuel are assessed with respect to specific transport conditions which are defined in the safety regulations of the International Atomic Energy Agency. In general, gastight fuel rods constitute the first barrier of the containment system. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important inputs for the evaluation of the package safety under transport conditions. The objective of this paper is to discuss the methodologies accepted by German authority BAM for the evaluation of spent fuel behavior within the package design approval procedure. Specific test conditions will be analyzed with regard to assumptions to be used in the activity release and criticality safety analysis. In particular the different failure modes of the fuel rods, which can cause release of gas, volatiles, fuel particles or fragments, have to be properly considered in these assumptions. The package as a mechanical system is characterized by a complex set of interactions, e.g. between the fuel rods within the assembly as well as between the fuel assemblies, the basket, and the cask containment. This complexity together with the limited knowledge about the material properties and the variation of the fuel assemblies regarding cladding material, burn-up and the operation history makes an exact mechanical analysis of the fuel rods nearly impossible. The application of sophisticated numerical models requires extensive experimental data for model verification, which are in general not available. The gaps in information concerning the material properties of cladding and pellets, especially for the high burn-up fuel, make the analysis more complicated additionally, and require a conservative approach. In this context some practical approaches based on experiences by BAM within safety assessment of packages for transport of spent fuel will be discussed. Ongoing research activities to investigate SNF mechanical behavior in view of gas and fissile material release under transport loads are presented. T2 - IRSN Conference on Safe Transport of Radioactive Material CY - Fontenay aux Roses, France DA - 13.11.2018 KW - ENSA KW - SNL KW - Real transport KW - IAEA KW - Drop test KW - Multi-Modal Transportation Test KW - Radioactive KW - ACT KW - NCT KW - Hot cell KW - Activity release KW - Fuel rod KW - Fuel rods KW - Burn-up KW - Cladding alloy KW - Criticality safety analysis KW - Containment analysis KW - Fuel assemblies KW - Spent fuel PY - 2018 AN - OPUS4-46880 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Rolle, Annette A1 - Ballheimer, Viktor A1 - Neumeyer, Tino A1 - Wille, Frank T1 - Application of leakage rates measured on scaled cask or component models to the package containment safety assessment N2 - The containment systems of transport casks for spent fuel and high radioactive waste usually include bolted lids with metallic or elastomeric seals. The mechanical and thermal loadings associated with the routine, normal and accident conditions of transport can have a significant effect on the leak tightness of such containment system. Scaled cask models are often used for providing the required mechanical and thermal tests series. Leak tests have been conducted on those models. It is also common practice to use scaled component tests to investigate the influence of deformations or displacements of the lids and the seals on the standard leakage rate as well as to study the temperature and time depending alteration of the seals. In this paper questions of the transferability of scaled test results to the full size design of the containment system will be discussed. T2 - ASME 2018 Pressure Vessels and Piping Conference CY - Prag, Czech Republic DA - 15.08 2018 KW - Containment KW - Spent fuel KW - Leakage rate PY - 2018 VL - 7 SP - PVP2018-84089, 1 EP - 8 AN - OPUS4-46533 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Zencker, Uwe T1 - Numerical approach to determine the correct puncture bar length for the IAEA puncture bar drop test N2 - Among other mechanical tests the 1 meter drop onto a steel puncture bar shall be considered for accident safe packages for the transport of radioactive material. According with the IAEA regulations “the bar shall be of solid mild steel of circular section, 15.0 ± 0.5 cm in diameter and 20 cm long, unless a longer bar would cause greater damage…”. The most damaging puncture bar length can be estimated by iterative processes in numerical simulations. On the one hand, a sufficient puncture bar length has to guarantee that shock absorbers or other attachments do not prevent or reduce the local load application to the package, on the other hand, a longer and thus less stiff bar causes a smaller maximum contact force. The contrary influence of increasing puncture bar length and increasing effective drop height shall be taken into account if a shock absorber is directly placed in the target area. The paper presents a numerical approach to identify the bar length that causes maximum damage to the package. Using the example of two typical package masses the sensitivity of contact forces and puncture bar deformations to the initial length are calculated and assessed with regard to the international IAEA package safety requirements. T2 - ASME Pressure Vessels and Piping Conference 2018 CY - Prague, Czech Republic DA - 15.07.2018 KW - Length of puncture bar KW - Mechanical assessment KW - Numerical simulation KW - Puncture bar test KW - Transport of radioactive materials PY - 2018 SP - PVP2018-84614, 1 EP - 7 AN - OPUS4-46538 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Rolle, Annette A1 - Ballheimer, Viktor A1 - Neumeyer, Tino A1 - Wille, Frank T1 - Application of leakage rates measured on scaled cask or component models to the package containment safety assessment N2 - The mechanical and thermal loadings associated with the routine, normal and accident conditions of transport can have a significant effect on the leak tightness of the containment system of transport casks for spent fuel and high radioactive waste. The containment systems of such transport casks usually include bolted lids with metallic or elastomeric seals. Scaled cask models are often used for providing the required mechanical and thermal tests series. Leak tests have been conducted on those models. It is common practice to use scaled component tests to investigate the influence of deformations or displacements of the lids and the seals on the standard leakage rate as well as to study the temperature and time depending alteration of the seals. In this paper questions and open points of the transferability of scaled test results to the fullscale design of the containment system will be discussed. T2 - 11th International Conference on the Transport, Storage and Disposal of Radioactive Materials CY - London, UK DA - 15. Mai 2018 KW - Transport packages KW - Radioactve KW - Seals KW - Leakage rate PY - 2018 AN - OPUS4-46734 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Müller, Lars A1 - Schönfelder, Thorsten A1 - Komann, Steffen A1 - Ballheimer, Viktor A1 - Wille, Frank ED - Müller, Lars T1 - Assessment experience on packages loaded with damaged spent nuclear fuel for transport after storage N2 - In 2017 the first German package approval certificate was issued for a dual purpose cask (DPC) design with encapsulated damaged spent nuclear fuel. At the Bundesanstalt für Materialforschung und -prüfung (BAM) a comprehensive assessment procedure was carried out with respect to the mechanical and thermal design, the containment design and quality assurance for manufacturing and operation. Main objective of this procedure was to verify the Package Design Safety Report (PDSR) fulfils the requirements according to the IAEA regulations SSR-6. Until now only standard spent nuclear fuel assemblies were designated for interim storage and transports. Due to nuclear phase out in Germany all other kinds of SNF in particular damaged fuel has to be packed. Therefore specific requirements have to be considered in accordance with international experiences written in IAEA technical reports. In Germany damaged spent nuclear fuel (DSNF) needs a tight encapsulation with special encapsulations and clearly defined properties. Due to the limited amount of DSNF these encapsulations are designed for storage and transport in existing packages. From the assessment experience it has been seen, corresponding PDSR need an extensively expansion to cover the design of these encapsulations and their influences on the package. Then such well-defined encapsulations can be handled like standard fuel assemblies. The main difference to standard package components is, encapsulations with permanent closure achieve their specified condition not after manufacturing but only during operation after loading and closing. Thus specific handling instruction and test procedures are necessary especially for welding, where BAM is able to survey the quality of this first part of operation. T2 - 11th International Conference on the Transport, Storage and Disposal of Radioactive Materials CY - London, UK DA - 15.05.2018 KW - Assessment KW - Dual purpose cask KW - Spent nuclear fuel PY - 2018 SP - Paper 18524, 1 EP - 8 AN - OPUS4-45258 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Revision of IAEA Guidance Material for Package Load Attachment Points N2 - For transport operations according IAEA regulations SSR-6, the package shall be securely stowed and the package shall be capable to withstand specific effects, e.g. accelerations, during routine transport conditions. The supporting IAEA Advisory Material SSG-26 provides information how to do that. General information and detailed load factors for use in safety demonstrations are provided. Up to now package designers in different countries use other load factors for the design of attachment points than those specified in the IAEA guidance material. In particular the acceleration factors vary between different countries and lead to difficulties during the validation of foreign approval certificates. The background of the existing guidance including specified load values was not clear enough. The application of specified values and the boundary conditions, i.e. transport conditions, seemed to be improvable. The IAEA started a discussion process to review the existing guidance text. An international working group was constituted in 2013. Representatives came from different stakeholders, e.g. transport operators, competent authorities and modal organizations. The discussions concluded especially on the transport conditions which has to be considered for stowage design, including on the one hand the relevance of the load factors used for strength and fatigue analysis and, on the other hand the criteria which have to be considered for the attachment points. In addition, complementary questions relative to the operational aspects were discussed. Revised load values for different transport modes were proposed. The values are derived from standards and include safety factors now. As a result of those discussions, the international working group wrote new guidance material for stowage in transport, addressing all the previously quoted topics. T2 - WM2018 Conference CY - Phoenix, Arizona, USA DA - 18.03.2018 KW - Acceleration values KW - IAEA regulations KW - Routine conditions of transport PY - 2018 AN - OPUS4-44866 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Bauartprüfung von Transportbehältern für verglaste Abfälle aus der Wiederaufarbeitung N2 - Aufgrund zwischenstaatlicher Verträge werden in den nächsten Jahrzehnten zahlreiche Transporte von hoch radioaktiven Abfällen notwendig. Die BAM führt die Bauartprüfung von dafür vorgesehenen Castor-Behältern durch. In diesem Zusammenhang sind 17 Fallprüfungen durchgeführt worden. Ergänzt werden diese durch komplexe numerische Analysen, die die BAM prüft und eigene Modellierungen vornimmt. Der Beitrag stellt darüber hinaus das mehrjährige Validierungsverfahren zur Akzeptanz der bruchmechanischen Berechnungswerkzeuge vor. T2 - VdTÜV Forum Kerntechnik CY - Berlin, Germany DA - 16.04.2018 KW - Transport KW - Bauartprüfung KW - Fallprüfungen KW - Radioaktive Stoffe PY - 2018 AN - OPUS4-44853 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Linnemann, Konrad A1 - Nehrig, Marko A1 - Musolff, André A1 - Wille, Frank T1 - Experience by BAM in the transport package safety assessment of German waste containers N2 - Transport and storage containers for low and midlevel radioactive waste are getting more and more of relevance due to the nuclear phase out decision in 2011. For higher activities Type B(U) approved waste containers will be needed for the shut down and dismantling phases of NPPs. It is expected that large quantities are required in the near future. German waste containers are generally approved for transport and interim storage and are also intended for the final disposal in the Konrad repository. BAM is involved in the authority licensing of transport package designs in Germany. In this context, BAM is responsible for the assessment of safe containment, mechanical, thermal and quality management issues. BAM also operates test facilities and performs drop as well as thermal tests during package licensing procedures. This paper summarizes our experience in the transport package design assessment of Type B(U) waste containers. A general overview of the approval process, the requirements and approaches BAM applies are described. Some examples are used to illustrate different aspects and technical issues we are addressing during the package assessment. In the first part of the paper the specific design aspects of German waste containers are described. Here, a general overview is given e.g. about closure systems, impact limiter designs, and the handling concepts. Furthermore, the wide range of radioactive content and their physical behavior including the impact on the packages assessment are described. The second part is focused on questions about the licensing and assessment process of German waste containers from the BAM authority point of view. The general approaches for the strategy of demonstration are outlined on the basis of the test conditions according to IAEA Regulations SSR-6. Furthermore, particular issues of the mechanical and thermal assessment with respect to the specific test conditions are discussed. For accident conditions of transport, aspects to be mentioned are e.g. the assessment of the lid bolts, the axial gap applied between content and lid, and the thermal behavior of the wood filled impact limiter after the fire test. However, issues of the assessment for routine and normal conditions of transport are addressed in this paper, too. The intention of this paper is introduce recent approval procedure experience in Germany, describing technical evaluation issues and so reduce rounds of questions during applications. T2 - 2018 WM Symposia CY - Phoenix, AZ, USA DA - 18.03.2018 KW - Transport packages KW - Safety assessment KW - Lowlevel and midlevel radioactive waste PY - 2018 SP - Paper 18469, 1 EP - 10 AN - OPUS4-44861 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Linnemann, Konrad A1 - Ballheimer, Viktor A1 - Sterthaus, Jens A1 - Rolle, Annette A1 - Wille, Frank A1 - Nasyrow, R. A1 - Papaioannou, D. A1 - Rondinella, V. A1 - Vlassopoulos, E. A1 - Pautz, A. T1 - Finite element modeling of spent fuel rod segments under bending loads N2 - Transport packages for spent nuclear fuel have to be assessed with respect to specific transport conditions which are defined in the regulations of the International Atomic Energy Agency. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important Inputs for the evaluation of the package capabilities under These conditions. Cracks or failures in the fuel rod cladding can cause the release of gas, volatiles or fuel particles into the cavity. The amount of substances in the cavity has to be considered in the assessment of the activity release and criticality safety. The mechanical analysis of the compound system formed by the fuel rod cladding and the spent fuel pellets is very difficult due to the limited knowledge of the material properties and the insufficient understanding of the interaction between pellets and cladding and between adjacent pellets. The variation of fuel assembly properties regarding cladding material, burn-up and the history of usage makes reliable predictions of the fuel rod behavior even harder. For a better understanding about the behavior of spent fuel rods, JRC and BAM have started a joint research project. In this context, JRC has developed a test device which allows quasi-static 3-point-bending test on fuel rod segments in the hot cell. The loads are applied with respect to the boundary conditions of the activity release assessment. This paper deals with the numerical calculation of a single fuel rod segment under bending load. The aim is to identify the governing mechanical parameters by the variation of constitutive assumptions, contact conditions, inner constraints, etc. This knowledge helps for the interpretation of the experimental results. Furthermore, the improved understanding about the behavior of the cladding-pellets system will be beneficial for the assessment of spent fuel transport conditions. T2 - SMIRT24 - 24th Conference on Structural Mechanics in Reactor Technology CY - Busan, Korea DA - 20.08.2017 KW - Finite element methods KW - Spent fuel assessment KW - Transport packages PY - 2017 SP - 1 EP - 8 AN - OPUS4-45316 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CHAP A1 - Nehrig, Marko A1 - Wille, Frank ED - Saegusa, Toshiari ED - Sert, Gilles ED - Völzke, Holger ED - Wille, Frank T1 - Internal Pressure build-up of Waste Packages with Wet Contents Under Fire Impact N2 - Intermediate level waste often contains residual water or other liquids. The liquids can be decomposed due to radiolysis and can also vaporize especially under fire impact. These pressure generating processes are described. One example for the pressure determination inside a closed cask is discussed using two different approaches: analytical and CFD KW - Wet Content KW - Pressure build-up KW - Type-B Package PY - 2018 SN - 978-981-3234-03-1 SP - 195 EP - 209 PB - World Scientific Publishing CY - Singapore ET - Erste AN - OPUS4-45301 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Nehrig, Marko A1 - Erenberg, Marina A1 - Feldkamp, Martin A1 - Bletzer, Claus A1 - Musolff, André A1 - Wille, Frank T1 - Large Scale IAEA Thermal Test with Wood filled Impact Limiters N2 - Packages for the transport of radioactive material are often equipped with impact limiters consisting of wood. Mostly this wood is encapsulated by steel sheets. The impact limiters are needed to ensure that the transport casks meet the IAEA safety requirements. According to the IAEA safety requirements a package has to withstand consecutively severe mechanical tests followed by a thermal test. The mechanical tests have to produce maximum damage concerning the thermal test. Following this, the impact limiters may have serious pre-damage when the thermal tests begins. The IAEA safety requirements state that during and following the fire test, the specimen shall not be artificially cooled and any combustion of materials of the package shall be permitted to proceed naturally. Small scale fire tests with wood filled metal drums by BAM and works of the French Institute for Radiological Protection and Nuclear Safety (IRSN) showed that pre-damaged steel encapsulated wooden structures could start smoldering initiated by the thermal test. These processes supply additional energy to the cask which should be considered within the safety assessment of the package. As not much is known about smoldering processes in encapsulated wooden structures with a reduced oxygen supply the need for a test was identified. To investigate the influence of a smoldering impact limiter concerning the amount of energy supplied to the cask in dependence of the time BAM conducted a large scale impact limiter thermal test. For that, a pre-damaged impact limiter with a diameter of 2,3 m was mounted on a water tank simulating a cask. A complex system of a regulated pump, a heater, a cooler, a slide valve, a flow meter and numerous thermocouples were installed and connected to a control unit to ensure all needed operating conditions. After a pre-heating compared to typical SNF decay-heat, the 30 min lasting fire phase of the thermal test was started. After that, the expected and initiated smoldering began. The results of the large scale test are presented in this poster. Systematic small scale tests will follow to identify the influence of different parameters, e.g. moisture content and scale effects. The tests took place at BAM Test Site for Technical Safety (TTS) with its various possibilities for mechanical and thermal tests. The results of these tests will have direct influence in the safety assessment of transport cask for the transport of radioactive material T2 - 2018 WM Symposia CY - Phoenix, AZ, USA DA - 18.03.2018 KW - Type-B Package KW - Thermal Test PY - 2018 SP - Paper 18257, 1 EP - 11 AN - OPUS4-45311 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Recent developments in standards and IAEA guidance material for package load attachment points N2 - For transport package design and operation according to IAEA regulations, the package shall be securely stowed and its retention system shall be capable to withstand load conditions of routine transport. The supporting IAEA Advisory Material SSG-26 provides information how to do that. Up to now package designers in different countries use other load factors for the design of attachment points than those specified in the IAEA guidance material. In particular the acceleration values vary between different countries and lead to difficulties during the validation of foreign approval certificates. Therefore the IAEA started a discussion process to review the existing guidance text. An international working group was constituted in 2013. Representatives came from different stakeholders, e.g. transport operators, competent authorities and modal organizations. The discussions concluded especially on the transport conditions which has to be considered for stowage design, including on the one hand the relevance of the load factors used for strength and fatigue analysis and on the other hand the criteria which have to be considered for the attachment points. Another standard with relevance to the load attachment of packages is ISO 10276. This standard deals with trunnion design, manufacturing and operational aspects. The regular standard revision phase started in 2017. An expert group discussed new state-of-the-art technology and proposed revised text for the ISO standard for international discussion. The paper describes relevant tie-down aspects, gives background argumentation, and tries to support harmonized application of the revised IAEA guidance material and future ISO 10276 standard. T2 - 11th International Conference on the Transport, Storage and Disposal of Radioactive Materials CY - London, UK DA - 15.05.2018 KW - Acceleration KW - Package design KW - ISO KW - Ladungssicherung PY - 2018 AN - OPUS4-45293 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Moutarde, M. A1 - Apel, Andreas A1 - Ballheimer, Viktor A1 - Sterthaus, Jens T1 - Recent developments in standards and IAEA guidance material for package load attachment points N2 - For transport package design and operation according to IAEA regulations, the package shall be securely stowed and its retention system shall be capable to withstand load conditions of routine transport. The supporting IAEA Advisory Material SSG-26 provides information how to do that. Up to now package designers in different countries use other load factors for the design of attachment points than those specified in the IAEA guidance material. In particular the acceleration values vary between different countries and lead to difficulties during the validation of foreign approval certificates. Therefore the IAEA started a discussion process to review the existing guidance text. An international working group was constituted in 2013. Representatives came from different stakeholders, e.g. transport operators, competent authorities and modal organizations. The discussions concluded especially on the transport conditions which has to be considered for stowage design, including on the one hand the relevance of the load factors used for strength and fatigue analysis and on the other hand the criteria which have to be considered for the attachment points. Another standard with relevance to the load attachment of packages is ISO 10276. This standard deals with trunnion design, manufacturing and operational aspects. The regular standard revision phase started in 2017. An expert group discussed new state-of-the-art technology and proposed revised text for the ISO standard for international discussion. The paper describes relevant tie-down aspects, gives background argumentation, and tries to support harmonized application of the revised IAEA guidance material and future ISO 10276 standard. T2 - 11th International Conference on the Transport, Storage and Disposal of Radioactive Materials CY - London, UK DA - 15.05.2018 KW - Verzögerungswerte KW - Lastanschlag KW - Trunnion design KW - ISO KW - Retention PY - 2018 SP - 18511, 1 EP - 8 AN - OPUS4-45294 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Debruyne, M. A1 - Eckert, B. A1 - Wille, Frank A1 - Gauthier, F. A1 - Le Bars, I. A1 - Cordier, N. A1 - Jouve, A.-C. T1 - Assessment of safety demonstrations relative to packages containing UF6 N2 - Specific attention should be paid on safety demonstrations transmitted by applicants in the case of approval request for the package designs containing enriched UF6. Concerning the shipment of enriched UF6, the package designs consist in general of a filled 30B cylinder surrounded by an overpack. The description of the content, considering the UF6 origin, i.e. natural or reprocessed, shall be clearly justified especially when the UF6 isotopic composition exceeds the limits specified in ASTM standards. Concerning the containment of the UF6, the applicant shall demonstrate in all conditions of Transport the leak-tightness of the valve and plug of the cylinders filled with enriched UF6. In this regard, when justifications are based on numerical calculations, the absence of contact between These components of the cylinder and the internal surfaces of the overpack after the regulatory drop tests shall be shown. In particular, absence of contact between the valve and any other component of the packaging shall be confirmed to respect the current IAEA regulations [3]. If complementary calculations show a contact between the plug and the internal surfaces of the overpack, additional tests are required to confirm that the strength resulting from this contact will not affect the plug leak-tightness. It can be noticed that the future revision of the IAEA regulations will include additional provision in case of contact of the plug with any other component of the packaging. In addition, the applicant shall demonstrate that the melting temperature of the valve, including the tinned joint, will not be exceeded during the regulatory fire test. Furthermore, the representativeness of the ballast used to simulate the behaviour of the UF6 loaded within the cylinder shall be justified if drop tests are performed. Finally, specific provisions relative to the use of plugs and the maintenance of cylinders should be included in the safety analysis reports. T2 - 11th International Conference on the Transport, Storage and Disposal of Radioactive Materials CY - London, UK DA - 15.05.2018 KW - Uranhexaflourid KW - Package safety KW - Valve KW - Drop testing KW - Radioactive material PY - 2018 SP - 18523, 1 EP - 8 AN - OPUS4-45295 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CHAP A1 - Ballheimer, Viktor A1 - Quercetti, Thomas A1 - Wille, Frank ED - Saegusa, Toshiari ED - Sert, Gilles ED - Völzke, Holger ED - Wille, Frank T1 - Internal cask content collision during drop tests N2 - The interaction between the package lid system and internal Content during mechanical drop testing is a decisive matter in evaluating Impact loads and the safety of the package. In the case of movable contents ist impact onto the inner side of the package lid can cause additional load peaks on the lid and the lid bolts. Some aspects of this issue were discussed on the basis of experimental results from instrumented drop tests with transport casks and on the basis of analytical approaches. KW - Transport casks KW - Drop tests KW - Internal collision PY - 2018 SN - 978-981-3234-03-1 SP - Chapter 7, 103 EP - 120 PB - World Scientific Publishing CY - Singapore ET - 1 AN - OPUS4-47539 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - KI-unterstützte Begutachtung der Behältersicherheit N2 - Die BAM führt sicherheitswissenschaftliche Begutachtungen von Gefahrgutbehältern durch, die mechanische, thermische, abschirmende und transportbezogene Analysen sowie Materialbewertungen und experimentelle Prüfungen umfassen. Diese Begutachtungen basieren auf umfassenden Sicherheitsberichten und Betriebsdokumentationen, die zahlreiche verknüpfte Dokumente und Daten enthalten. Änderungen an Komponenten oder regulatorischen Vorgaben können große Auswirkungen auf den Prüfumfang haben. Das Projekt sieht vor, die Datenanalyse und -bereitstellung mittels KI-Ansätzen zu verbessern. Es sollen Methoden zur automatischen Analyse und Verknüpfung komplexer Dokumentationen entwickelt werden. Ziel ist es, die Effizienz und Sicherheit der Datenanalyse und späteren Sicherheitsbewertung zu steigern, menschliche Fehler zu minimieren und die Qualität der Sicherheitsbewertungen zu verbessern. Die Qualität der Datenbereitstellung aus Zeichnungen, Tabellen, Texten und Berechnungen ist entscheidend für die Verarbeitung digitaler Formate der Behälterdokumentation. Ein neuro-symbolischer KI-Ansatz kombiniert neuronale Netze und symbolische KI, um die komplexen Inhalte zu analysieren und zu verknüpfen, was für die spätere Sicherheitsbewertung der Gefahrgutbehälter entscheidend ist. In vortrainierten Large Language Models (LLMs) fehlen für diesen Anwendungsfall spezifische fachliche Daten zur Behältersicherheit was zu fehlerhaften Ergebnissen führen kann. Für die Datenbereitstellung und Analyse im Zuge einer Sicherheitsbewertung gilt es dieses auszuschließen. Hierzu soll die RAG-Architektur (Retrieval-Augmented Generation) im Zusammenspiel mit LLMs angewendet werden. Dieser neue Ansatz der natürlichen Sprachverarbeitung und generativen KI soll die Stärken vortrainierter Sprachmodelle mit Fachwissen und Informationen aus Datenbanken und Dokumentensammlungen kombinieren. Die generierten Ergebnisse sollen dabei zusammen mit den verwendeten Quellen bereitgestellt werden, um eine fundierte und nachvollziehbare Basis für die Sicherheitsbewertung bilden. T2 - Informationsaustausch zur verkehrsrechtlichen Aufsicht bei der Beförderung radioaktiver Stoffe CY - Bonn, Germany DA - 04.02.2025 KW - KI KW - RAG KW - LLM PY - 2025 AN - OPUS4-62538 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias A1 - Feldkamp, Martin A1 - Quercetti, Thomas A1 - Wille, Frank T1 - Experimental and Numerical Analyses for the Evaluation of Heat Fluxes of a Fire Reference Test N2 - Packages for the transport of high-level radioactive material are designed to endure severe accidents. To obtain approval, these transport packages must adhere to the specification-based criteria of the IAEA SSR-6. To ensure compliance with these requirements, specific mechanical and thermal tests need to be appointed with respect to the package type. Typically, IAEA SSR-6 prescribes mechanical tests followed by a thermal fire test as part of a cumulative test sequence. To approve the fire test, BAM uses a reference package representing the original package in its outer geometry for characterizing the actual fire and its impact on the package. This serves a dual purpose: it allows the precise adjustment of experimental parameters for the package design to be approved, and it provides input parameters for thermomechanical simulations, cf. With the help of this methodology for characterizing boundary conditions of the package, temperature evolutions within the reference package can be studied in a Finite-Element Analyses (FEA). This helps to compare the experiments with the numerical simulations on the case of the reference package, but also serves at the same time for preliminary simulations with the package design to be approved. The thermal test consists of a 30-minute fully engulfing 800°C pool fire or an equally severe fire, e.g. a propane gas fire. The fire reference test is performed prior to the regulatory fire test with the package design to be approved. The fire reference package in the described case is a closed sheet 316L steel cylinder with a wall thickness of 10 mm, a length of 182 mm, and a diameter of 102 mm. The package was instrumented with thermocouples and filled with heat resistant insulating material. T2 - ASME PVP 2024 CY - Bellevue, WA, USA DA - 28.07.2024 KW - Reference Test KW - Experimental KW - Numerical KW - Heat Flux PY - 2024 SN - 978-0-7918-8851-3 SP - 1 EP - 8 AN - OPUS4-60848 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Naster, Maximilian A1 - Gleim, Tobias A1 - Wille, Frank T1 - Experimental and Numerical Analyses for the Evaluation of Hydrogen as an Energy Source for Thermal Testing of Transport Packages of Radioactive Material N2 - In this paper we present a new hydrogen-based test rig for an ongoing feasibility study of using hydrogen as an energy source for the thermal testing of transport packages containing radioactive materials. The test rig will be capable of combusting hydrogen for a wide range of different burner geometries, mass flows and if necessary hydrogen blends. As this type of fire test according to the IAEA boundary conditions does not yet exist, a large number of preliminary investigations, safety assessments and calculations must be carried out in order to develop a viable concept for hydrogen fires. In the first step of the feasibility study, the temperature, structure, and radiation of various hydrogen flames are surveyed. In future works, the results will make it possible to design burner frames that are suitable for fire reference tests in order to make comparisons with pool and propane fires used in assessment procedures today. In parallel comparative numerical simulations are conducted to model the thermal behaviour of hydrogen flames using the software package Ansys®. On the one hand, the numerical simulations support the experiments by providing an overview of numerous parameters and the measuring range; on the other hand, they will help with the design of the burner frame in future work. This paper gives an overview in the design and capabilities of the test rig. Furthermore, the results of the parameter studies show that burner geometry and mass flow provide a significant design margin for the thermal shape of the hydrogen flames. In addition, the results of the initial numerical calculations will be used to determine the necessary sensors, the positions, and their operating range. Only the optimal interaction allows a controlled system that permits user-defined hydrogen fires. T2 - PVP2024, Pressure Vessels & Piping Conference CY - Bellevue, Washington USA DA - 29.07.2024 KW - Hydrogen KW - Fire KW - Computational Fluid Dynamics (CFD) KW - IAEA fire test KW - Burner PY - 2024 SN - 978-0-7918-8851-3 SP - 1 EP - 9 AN - OPUS4-60852 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Müller, Lars A1 - Scheidemann, Robert A1 - Neumeyer, Tino A1 - Komann, Steffen A1 - Wille, Frank T1 - Tightness Verification Procedures of Welded Lids for Encapsulations for Damaged Spent Nuclear Fuel in the Design Approval Process of Dual-Purpose Cask (DPC) N2 - Damaged spent nuclear fuel (DSNF) can be loaded in Germany in dual-purpose casks (DPC) for transport and interim storage. Therefore, special encapsulations are needed to guarantee a safe handling and a reliable tight closure. These encapsulations must be durable and leak-tight for a possible long interim storage term and the transport thereafter. Because components inside the package containment like these special encapsulations are not accessible within periodical inspections of the DPC. Due to the general design of DPCs for standard fuel assemblies, specific requirements needed to be considered for the design of the special encapsulations for DSNF to ensure the loading in existing package designs. Especially the content of the encapsulations with damaged spent nuclear fuel must be avoided to enter in the inner space of the DPC, due to the possible higher amount of corrosive and radioactive material as well as greater quantities of water, which can be trapped in defect fuel rods. The main difference of the encapsulations for damaged spent nuclear fuel is that they have no replaceable sealing as tightness barrier. They have a specific welded lid. Therefor the tightness needs to be shown in the design approval process for all transport conditions especially for accident conditions, where high internal impact loads can occur. As German competent authority, BAM is responsible for the safety assessment of the mechanical and thermal design, the release of radioactive material and the quality assurance aspects of manufacturing and operation. Concerning the mechanical package design BAM carried out a comprehensive safety assessment for the specific encapsulations during the package design approval process of the commonly used DPC. For the proof of tightness of the welded lid no representative standards are available. For minor loads the German KTA standards can verify the tightness of a weld seam but only for low stresses. For higher impact loads physical tests are required. These tests were performed with representative drop tests and highly sensitive leak tests. The present paper gives an overview of the applied assessments by BAM and points out the main findings of the mechanical design for the welded sealing. Both verification procedures are described, on one hand with German standards for lower loads and on the other hand with physical tests for higher impact loads. The drop tests based on complex numerical pre-calculations by the applicant which result in optimized dimensions and properties of the test specimens and representative drop heights for the drop tests. The horizontal drop orientation with a plane-parallel impact on the unyielding basement was most challenging and led to specific adaptations during preparation and conduction of the drop tests performed by BAM. With leak tests and the corresponding test evaluation, it could be finally approved that the specific welded sealing of an encapsulations for DSNF is tight for all transport conditions. T2 - ASME PVP 2024 CY - Bellevue, WA, USA DA - 28.07.2024 KW - Tightness verification KW - Weld seams KW - Drop tests KW - Encapsulations for damaged spent nuclear fuel KW - Helium leak tests PY - 2024 SN - 978-0-7918-8851-3 SP - 1 EP - 10 CY - New York AN - OPUS4-61351 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Wille, Frank A1 - Linnemann, Konrad A1 - Ramsay, Jeff A1 - Reiche, Ingo T1 - Transport of large nuclear power plant components – new SCO-III regulations and mechanical design assessment N2 - The decommissioning or refurbishment of nuclear facilities necessitates either the storage or disposal of large radioactive components such as steam generators, pressurizers, reactor pressure vessels and heads, and coolant pumps, to list the major contributors. These components or objects are large in size and mass, measuring up to approximately 6 meters in diameter, up to 20 meters in length, and weighing over 400 000 kg. In many situations, the components are transported off-site to a storage, disposal or recycling/treatment facility. Previously, many large objects had to be transported under special arrangement. The latest 2018 edition of the International Atomic Energy Agency (IAEA) Regulations for the Safe Transport of Radioactive Material, No. SSR-6, incorporates regulations for the shipment of contaminated, but not activated large objects as a new category of surface contaminated object, SCO-III, based on the IAEA “performance package” concept. This paper provides background and practical guidance on these regulations. Additionally, first experiences with SCO-III in the approval process are presented. For SCO-III components a mechanical assessment regarding the tests specified in the transport regulations SSR-6 for demonstrating ability to withstand normal conditions of transport is required - a free drop of the package onto an unyielding target and a penetration test have to be investigated. Large components are unique objects; therefore, it is not possible to choose experimental testing as assessment method. The application of a complex numerical analysis for mechanical proof is necessary. The assessment of the loads takes place based on local stress distributions. The large nuclear components have typically been transported in an unpackaged manner, so that the external shell of the component provides the containment wall. The paper will represent the methods, which are used for the evaluation of the mechanical integrity of the object for transport approval. T2 - PVP2024, Pressure Vessels & Piping Conference CY - Bellevue, Washington USA DA - 29.07.2024 KW - Transport KW - Radioaktive Stoffe KW - Rückbau kerntechnischer Anlagen KW - Gefahrgut PY - 2024 SN - 978-0-7918-8851-3 SP - 1 EP - 4 AN - OPUS4-60829 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Baeßler, Matthias A1 - Herrmann, Ralf A1 - Hille, Falk A1 - Wille, Frank T1 - Dynamisches Verhalten und Überwachung des Großen Fallturms der BAM in Horstwalde N2 - Falltürme sind Bauwerke, die als Versuchsanlagen sehr speziellen, impulsartigen Belastungen ausgesetzt sind. Am Fallturm auf dem Testgelände Technische Sicherheit (TTS) der BAM wurde bei Routineinspektionen eine Abnahme der strukturellen Integrität in Form von Vorspannungsverlusten in den Bolzen der Stahlverbindungen festgestellt. Um ein Verständnis für die zugrundeliegenden Trag- und Schädigungsmechanismen zu erlangen, wurde ein umfassendes Structural Health Monitoring (SHM) System geplant und am Bauwerk installiert unter Nutzung von digitalen Bauwerksmodellen. Für die Auslegung des Überwachungssystems, insbesondere aber zur Unterstützung der Untersuchung des Schädigungsprozesses, wurden Finite-Elemente-Modelle erstellt. Um sicherzustellen, dass die FE-Modelle das reale Tragwerksverhalten mit ausreichender Genauigkeit abbilden, mussten sie jedoch in Bezug auf die gemessenen Antworten des Tragwerks kalibriert werden. Der vorliegende Beitrag beschreibt experimentelle und numerische Untersuchungen zur Identifizierung des strukturellen Systems des Stahlrohrgitterturms in Vorbereitung einer Überwachungskampagne. Die Auswertung von gemessenen Schwingungen unter ambienter Anregung ermöglichte die Identifizierung der Eigenfrequenzen mehrerer globaler Schwingungsmoden des Fallturms. Zur Modellvalidierung wurde zunächst eine Sensitivitätsanalyse durchgeführt, um die Parameter mit dem größten Einfluss zu ermitteln. Anschließend wurde ein evolutionärer Algorithmus (EA) zur Optimierung nach dem Prinzip der Minimierung der Differenzen zwischen gemessenen und simulierten charakteristischen Antworten eingesetzt. Das aktualisierte Modell wurde schließlich an der dynamischen Reaktion der Turmstruktur infolge einer realen Falltest-induzierten Stoßbelastung validiert. Die Ergebnisse zeigten eine gute Übereinstimmung zwischen numerischen und experimentellen Ergebnissen. N2 - Drop test towers are structures subjected to very specific dynamic loadings. At the drop test tower at the Test Site Technical Safety (TTS) of BAM routine inspections revealed a decline of structural integrity in form of loss of pretension in bolts of structural steel connections. To get an understanding about the underlying damage mechanisms as well as to assure the structural safety in operation, a multifaceted Structural Health Monitoring (SHM) system was planned and installed at the structure with the help of digital models. For planning the monitoring system but especially for supporting the investigation of the damaging process finite element models are deployed. Though, to ensure the capability of the FE model to reproduce the real structural behavior in sufficient accuracy, it needs to be calibrated in respect to measured responses of the structure. The present contribution describes experimental and numerical investigations to identify the structural system of the steel tube lattice tower in preparation of the monitoring campaign. An ambient vibration test allowed the identification of the natural frequencies of several global vibrational modes of the drop test tower. For model updating first a sensitivity analysis is applied to identify the most influencing parameter. In succession a genetic algorithm was applied for optimization in the sense of minimization of the difference between measured and simulated characteristic responses. The updated model is finally validated on the dynamic response of the tower structure to a real drop test induced impact loading. The results showed a good agreement between numerical and experimental results though the robustness of the simulation should be improved by more sophisticated numerical modelling. T2 - 7. VDI-Fachtagung Baudynamik CY - Würzburg, Germany DA - 27.04.2022 KW - Großer Fallturm Horstwalde KW - Structural Health Monitoring KW - Belastungsversuch KW - Modalanalyse KW - Schwingungsdynamik KW - Finite Elemente Simulation PY - 2022 SN - 978-3-18-092379-6 SN - 978-3-18-102379-2 DO - https://doi.org/10.51202/9783181023792-265 SN - 0083-5560 VL - 2379 SP - 265 EP - 284 PB - VDI-Verlag CY - Düsseldorf AN - OPUS4-55472 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Droste, Bernhard A1 - Komann, Steffen A1 - Wille, Frank A1 - Rolle, Annette A1 - Probst, Ulrich A1 - Schubert, Sven T1 - Considerations of aging mechanisms influence on transport safety and reliability of dual purpose casks for spent nuclear fuel or HLW N2 - When storage of spent nuclear fuel (SNF) or high-level waste (HLW) is done in dual purpose casks (DPC), the effects of aging on safety relevant DPC functions and properties have to be managed in a way that a safe transport after the storage period of several decades is capable, and can be justified and certified permanently throughout that period. The effects of aging mechanisms (like e.g. radiation, different corrosion mechanisms, stress relaxation, creep, structural changes and degradation) on the transport package design safety assessment features have to be evaluated. The consideration of these issues in the DPC transport safety case will be addressed. Special attention is given to all cask components which cannot be directly inspected or changed without opening the cask cavity, what are the inner parts of the closure system and the cask internals, like baskets or spent fuel assemblies. The design criteria of that transport safety case have to consider the operational impacts during storage. Aging is not subject of technical aspects only, but also of 'intellectual' aspects, like changing standards, scientific/ technical knowledge development and personal as well as institutional alterations. Those aspects are to be considered in the management system of the license holders and in appropriate design approval update processes. The paper addresses issues which are subject of an actual IAEA TECDOC draft 'Preparation of a safety case for a dual purpose cask containing spent nuclear fuel'. T2 - PSAM 12 - Probabilistic safety assessment and management CY - Honolulu, Hawaii, USA DA - 22.06.2014 KW - Transport and storage casks for spent nuclear fuel or high level waste KW - Aging mechanisms KW - Corrosion KW - Safety assessment KW - Metal seals KW - Closure system KW - Spent fuel/high-level waste KW - Dual purpose casks KW - Metal seals reliability KW - Cesium corrosion PY - 2014 UR - http://psam12.org/proceedings/paper/paper_180_1.pdf SP - 1 EP - 10 AN - OPUS4-32518 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Rolle, Annette A1 - Droste, Bernhard A1 - Schubert, Sven A1 - Probst, Ulrich A1 - Wille, Frank T1 - Verification of activity release compliance with regulatory limits within spent fuel transport casks T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Radioactive material KW - Safety assessment KW - Leakage mechanism KW - Leakage rate KW - Sealing behavior PY - 2010 SP - 1 EP - 8(?) AN - OPUS4-32193 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Qualitätsfall bei der Fertigung von Tragzapfen für Castor-Behälter T2 - Informationsaustausch zur verkehrsrechtlichen Aufsicht bei der Beförderung radioaktiver Stoffe CY - Handelskammer, Bremen, Deutschland DA - 2015-01-27 PY - 2015 AN - OPUS4-32785 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Auffälligkeiten an Hebeösen von UF6-Behältern T2 - Informationsaustausch zur verkehrsrechtlichen Aufsicht bei der Beförderung radioaktiver Stoffe CY - Handelskammer, Bremen, Deutschland DA - 2015-01-27 PY - 2015 AN - OPUS4-32786 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Wille, Frank A1 - Droste, Bernhard T1 - Gap analysis examples from periodical reviews of transport package design safety reports of SNF/HLW dual purpose casks N2 - Storage of spent nuclear fuel and high-level waste in dual purpose casks (DPC) is related with the challenge of maintaining safety for transportation over several decades of storage. Beside consideration of aging mechanisms by appropriate design, material selection and operational controls to assure technical reliability by aging management measures, an essential issue is the continuous control and update of the DPC safety case. Not only the technical objects are subject of aging but also the safety demonstration basis is subject of "aging" due to possible changes of regulations, standards and scientific/technical knowledge. The basic document, defining the transport safety conditions, is the package design safety report (PDSR) for the transport version of the DPC. To ensure a safe transport in future to a destination which is not known yet (because of not yet existing repository sites) periodical reviews of the PDSR, in connection with periodic renewals of package design approval certificates, have to be carried out. The main reviewing tool is a gap analysis. A gap analysis for a PDSR is the assessment of the state of technical knowledge, standards and regulations regarding safety functions of structures, systems and components. T2 - PSAM 12 - Probabilistic safety assessment and management CY - Honolulu, Hawaii, USA DA - 22.06.2014 KW - Dual Purpose Casks KW - Aging KW - Transportation KW - Periodical Review PY - 2014 UR - http://psam12.org/proceedings/paper/paper_259_1.pdf SP - Paper 259, 1 EP - 9 AN - OPUS4-32574 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Linnemann, Konrad A1 - Ballheimer, Viktor A1 - Sterthaus, Jens A1 - Wille, Frank T1 - Safety assessment of Lid systems of transport packages for radioactive materials by use of finite element analyses N2 - The regulatory compliance of the containment system is of essential importance for the design assessment of transport packages for radioactive materials. The requirements of the IAEA transport regulations SSR-6 for accident conditions implies high load on the containment system of Type B(U) packages. The integrity of the containment system has to be ensured under the mechanical and thermal tests. The containment system of German transport packages for spent nuclear fuel (SNF) and high level waste (HLW) usually includes bolted lids with metal gaskets. BAM Federal Institute for Materials Research and Testing as the German competent authority for the mechanical and thermal design assessment of approved transport packages has developed the guideline BAMGGR 012 for the analysis of bolted lid and trunnion systems. According to this guideline the finite element (FE) method is recommended for the calculations. FE analyses provide more accurate and detailed information about loading and deformation of such kind of structures. The results allow the strength assessment of the lid and bolts as well as the evaluation of relative displacements between the lid and the cask body in the area of the gasket groove. This paper discusses aspects concerning FE simulation of lid systems for SNF and HLW transport packages. The work is based on the experiences of BAM within safety assessment procedures. The issues considered are the assessment methods used in the BAM-GGR 012 for bolted lid systems along with the nominal stress concept which is applied for bolts according to that guideline. Additionally, modeling strategies, analysis techniques and the interpretation of the results are illustrated by the example of a generalized bolted lid systems under selected accident conditions of transport. T2 - PVP2014 - ASME 2014 Pressure vessels & piping division conference CY - Anaheim, CA, USA DA - 20.07.2014 PY - 2014 SN - 978-0-7918-4600-1 SP - 28304, 1 EP - 12 AN - OPUS4-32383 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Sterthaus, Jens A1 - Ballheimer, Viktor A1 - Bletzer, Claus Wilhelm A1 - Linnemann, Konrad A1 - Nehrig, Marko A1 - Wille, Frank T1 - Numerical approach for containment assessment of transport packages under regulatory thermal test conditions N2 - The requirements of the IAEA safety standards for Type B(U) packages include the thermal test as part of test sequences that represents accident conditions of transport. In comparison to mechanical tests, e.g., 9 m drop onto an unyielding target with short impact durations in a range of approximately 10 ms to 30 ms, the extended period of 30 min is defined in regulations for exposure of a package to a fire environment. Obviously, the required containment capability of the package has to be ensured not only after completing the test sequence but also over the course of the fire test scenario. Especially, deformations in the sealing area induced by the non-uniform thermal dilation of the package can affect the capability of the containment system. Consequently, thermo-mechanical analyses are required for the assessment. In this paper some aspects of finite element analysis (FEA) of transport packages with bolted closure systems under thermal loading are discussed. A generic FE model of a cask is applied to investigate the stress histories in the bolts, lid, and cask body as well as the deformations in the sealing area and the compression conditions of the gasket. Based on the parameter variations carried out, some recommendations in regard to modeling technique and results interpretation for such kind of analyses are finally given. T2 - PVP2014 - ASME 2014 Pressure vessels & piping division conference CY - Anaheim, CA, USA DA - 20.07.2014 KW - Transport packages KW - Containment KW - Thermo-mechanical analyses KW - FEA KW - Gasket PY - 2014 SN - 978-0-7918-4600-1 DO - https://doi.org/10.1115/PVP2014-28199 VL - 3 SP - 28199, 1 EP - 14 AN - OPUS4-32349 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Applicabele Standards on Retention System Design in Germany T2 - IAEA Technical Meeting on Tie-Down System Design CY - Vienna, Austria DA - 2014-03-24 PY - 2014 AN - OPUS4-30707 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Konzept Brennelement-Transporte nach Zwischenlagerung T2 - 4. RAM Behältersicherheitstage, BAM CY - Berlin, Germany DA - 2014-03-26 PY - 2014 AN - OPUS4-30708 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - German Approach and Experience Feedback of Transport ability of SNF Packages after interim storage T2 - PATRAM 2013 17th International Symposium on Packaging and Transportation of Radioactive Materials CY - San Francisco, CA, USA DA - 2013-08-18 PY - 2013 AN - OPUS4-29771 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Müller, Lars A1 - Komann, Steffen A1 - Wille, Frank A1 - Droste, Bernhard T1 - Sicherheitstechnische Begutachtung der strukturmechanischen Auslegung von Castor-Behältern N2 - Die Sicherheitsanforderungen, die an Behälter für den Transport radioaktiver Stoffe auf öffentlichen Verkehrswegen gestellt werden, orientieren sich an dem Gefährdungspotenzial des radioaktiven Inhalts. So müssen für den Transport abgebrannter Brennelemente oder hochradioaktiver Abfalle aus nuklearen Anlagen Behälter eingesetzt werden, die auch schweren Unfällen standhalten. Zu ihnen gehören z. B. die CASTOR®-Behälter (Cask for Storage and Transport of Radioactive Material). T2 - BTU Stahlbau-Symposium 2014 CY - Cottbus, Germany DA - 23.05.2014 PY - 2014 SN - 1611-5023 N1 - Serientitel: Schriftenreihe Stahlbau – Series title: Schriftenreihe Stahlbau IS - 8 SP - 33 EP - 40 AN - OPUS4-31148 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Situation and Gap Analysis of SNF Transportation after Interim Storage in Germany T2 - INMM Spent Fuel Management Seminar XXIX 2014 CY - Arlington, VA, USA DA - 2014-01-13 PY - 2014 AN - OPUS4-30271 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wolff, Dietmar A1 - Völzke, Holger A1 - Wille, Frank A1 - Droste, Bernhard T1 - Extended storage after long-term storage N2 - Existing spent nuclear fuel (SF) and high active waste (HAW) management policies and practices worldwide are the result of past presumptions that sufficient reprocessing and/or disposal capacity would be available in the near term. Consequently, in the past many countries have developed specific solutions for different periods of time due to their individual national nuclear policies. In Germany the concept of dry interim storage in dual purpose metal casks before disposal is being pursued for SF and HAW management and transport and storage licenses have been issued accordingly. The current operation licenses for existing storage facilities have been granted for a storage period of up to 40 years. This concept has demonstrated its suitability for over 20 years so far. Relevant safety requirements haven been assessed for the short-term as well as for the long-term for site-specific operational and accidental storage conditions. But in the meantime significant delays in the national repository siting procedure occured which will make extended storage periods necessary in the future. This paper describes the current situation in Germany with regard to dry cask storage and focuses on current perspectives considering regulatory, technical, and scientific aspects for storage license renewal. Since there is one case of a storage license limitation to only 20 years due to administrative reasons, first experience is currently gathered in case of an expiring storage license. Subsequent license options have been pursued intensively including the extension of the initial storage license as well as shipping all casks to another storage facility considering an extended storage period at that storage facility as well. All safety relevant aspects have to be reviewed on basis of the current state-of-the-art which might be different from the initial safety demonstrations. That includes new safety assessment standards as well as improved knowledge base. Major issues are e.g. improved accident scenario analyses, assessment methods, and consideration of aging effects from previous operation periods. Inspection programs with respect, e.g. to bolts and seals to verify leak-tightness and lid screw pre-stress have been initiated to demonstrate proper cask conditions for extended storage as well as transportation to another storage facility. T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 KW - Extended storage KW - Long-term storage KW - Dual purpose cask PY - 2013 SP - Paper 184, 1 EP - 12 PB - Omnipress AN - OPUS4-30242 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Nehrig, Marko A1 - Feldkamp, Martin A1 - Bletzer, Claus Wilhelm A1 - Wille, Frank T1 - Thermal test driven pressure build-up inside type-B packages containing wet radioactive waste N2 - In recent years several German approval procedures for ductile cast iron transport containers containing wet intermediate level waste were conducted. BAM, as one of the German competent authorities, was involved in the complex design assessment work with this specific issue. Thermal analysis is one part of the authority assessment work done by BAM in Germany. The radioactive contents of package designs which were not dried, only drained, consist of saturated ion exchange resin and a small amount of free water. Compared to the safety assessment of packages with dry content, attention must be paid to some more specific points. The most interesting point, however, is the pressure build-up inside the package due to vaporization. This could be caused by radiolysis of the liquid and must be taken into account for the storage period. The inner pressure of the package leads to mechanical loads to the package body, the lid and the lid bolts. Thus, the pressure is the driving force on the gasket system regarding the activity release and a possible loss of tightness. The paper deals primarily with the pressure build-up inside the package caused by the transport regulatory thermal test (30 min at 800 °C) as part of the cumulative test scenario under accident conditions of transport. The pressure build-up is estimated by calculation in a very conservative way regarding conduction and heat radiation. Furthermore the paper discusses a conservative approach for the estimation of the resulting pressure depending on the percentage of water inside the cask. To get trustworthy results without an exact specification of the content, experimental fire tests should be conducted. However, this paper shows the difficulties of assessing casks containing wet content. From the authority assessment point of view, drying of the content could be an effective way to avoid the above described pressure build-up and the associated difficulties for the safety assessment. T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 KW - Type-B KW - Wet content KW - Pressure build-up KW - Radioactive waste KW - Thermal test PY - 2013 SP - Paper 108, 1 EP - 10 PB - Omnipress AN - OPUS4-30197 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank T1 - Dicht und sicher abgeschirmt N2 - BAM-GGR Transportverpackungen für radioaktive Stoffe müssen den BAM-Gefahrgutregeln entsprechen. Ein kurzer Einblick in die Richtlinien. PY - 2014 SN - 0944-6117 VL - 3 SP - 18 EP - 20 PB - Vogel CY - München AN - OPUS4-30487 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Linnemann, Konrad A1 - Ballheimer, Viktor A1 - Rolle, Annette T1 - Assessment approaches of mechanical behavior of SNF under transport package test conditions N2 - German packages for the transport of spent nuclear fuel are assessed with respect to specific transport conditions which are defined in the safety regulations of the International Atomic Energy Agency. In general, gastight fuel rods constitute the first barrier of the containment system. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important inputs for the evaluation of the package safety under transport conditions. The objective of this paper is to discuss the methodologies accepted by German authority BAM for the evaluation of spent fuel behavior within the package design approval procedure. Specific test conditions will be analyzed with regard to assumptions to be used in the activity release and criticality safety analysis. In particular the different failure modes of the fuel rods, which can cause release of gas, volatiles, fuel particles or fragments, have to be properly considered in these assumptions. The package as a mechanical system is characterized by a complex set of interactions, e.g. between the fuel rods within the assembly as well as between the fuel assemblies, the basket, and the cask containment. This complexity together with the limited knowledge about the material properties and the variation of the fuel assemblies regarding cladding material, burn-up and the operation history makes an exact mechanical analysis of the fuel rods nearly impossible. The application of sophisticated numerical models requires extensive experimental data for model verification, which are in general not available. The gaps in information concerning the material properties of cladding and pellets, especially for the high burn-up fuel, make the analysis more complicated additionally, and require a conservative approach. In this context some practical approaches based on experiences by BAM within safety assessment of packages for transport of spent fuel will be discussed. Ongoing research activities to investigate SNF mechanical behavior in view of gas and fissile material release under transport loads are presented. T2 - International High-Level Radioactive Waste Management Conference, IHLRWM 2017 CY - Charlotte, NC, USA DA - 09.04.2017 KW - Brennelementverhalten KW - Spent fuel behavior KW - Hüllrohr KW - Kritikalität KW - Aktivitätsfreisetzung KW - Mechanisches Verhalten KW - Cladding KW - Criticality PY - 2017 SP - 472 EP - 475 AN - OPUS4-39898 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Assessment approaches of mechanical behavior of SNF under transport package test conditions N2 - German packages for the transport of spent nuclear fuel are assessed with respect to specific transport conditions which are defined in the safety regulations of the International Atomic Energy Agency. In general, gastight fuel rods constitute the first barrier of the containment system. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important inputs for the evaluation of the package safety under transport conditions. The objective of this paper is to discuss the methodologies accepted by German authority BAM for the evaluation of spent fuel behavior within the package design approval procedure. Specific test conditions will be analyzed with regard to assumptions to be used in the activity release and criticality safety analysis. In particular the different failure modes of the fuel rods, which can cause release of gas, volatiles, fuel particles or fragments, have to be properly considered in these assumptions. The package as a mechanical system is characterized by a complex set of interactions, e.g. between the fuel rods within the assembly as well as between the fuel assemblies, the basket, and the cask containment. This complexity together with the limited knowledge about the material properties and the variation of the fuel assemblies regarding cladding material, burn-up and the operation history makes an exact mechanical analysis of the fuel rods nearly impossible. The application of sophisticated numerical models requires extensive experimental data for model verification, which are in general not available. The gaps in information concerning the material properties of cladding and pellets, especially for the high burn-up fuel, make the analysis more complicated additionally, and require a conservative approach. In this context some practical approaches based on experiences by BAM within safety assessment of packages for transport of spent fuel will be discussed. Ongoing research activities to investigate SNF mechanical behavior in view of gas and fissile material release under transport loads are presented. T2 - International High-Level Radioactive Waste Management Conference, IHLRWM 2017 CY - Charlotte, NC, USA DA - 09.04.2017 KW - Spent fuel KW - Cladding failure KW - Criticality analysis KW - Encapsulation PY - 2017 AN - OPUS4-39908 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Wille, Frank A1 - Schönfelder, Thorsten T1 - German concept to ensure transportability of SNF packages after interim storage N2 - In Germany the concept of dry interim storage of spent fuel in dual purpose metal casks is implemented, currently for periods of up to 40 years. The casks being used have an approved transport package design in accordance with the international IAEA transport regulations as well. The storage of spent nuclear fuel and high-level waste in dual purpose casks (DPC) is related with the challenge of maintaining safety for transportation over several decades of storage. Besides consideration of aging mechanisms by appropriate design, material selection and operational controls to assure technical reliability by aging management measures, an essential issue is the continuous control and update of the DPC safety case. Not only the technical objects are subject of aging but also the safety demonstration basis is subject of “aging” due to possible changes of regulations, standards and scientific/technical knowledge. To ensure a safe transport in future to a destination which is not known yet (because of not yet existing repository sites) periodical reviews of the Package Design Safety Report (PDSR), in connection with periodic renewals of package design approval certificates, have to be carried out. T2 - International High-Level Radioactive Waste Management Conference, IHLRWM 2017 CY - Charlotte, NC, USA DA - 09.04.2017 KW - Dual purpose casks KW - Interim storage KW - Transportability PY - 2017 SP - 476 EP - 481 AN - OPUS4-39919 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin A1 - Erenberg, Marina A1 - Nehrig, Marko A1 - Bletzer, Claus A1 - Musolff, André A1 - Wille, Frank A1 - Schönfelder, Thorsten T1 - Aspects of assessment of packages with wood filled impact limiters during fire tests N2 - Packages for the transport of radioactive material are often equipped with impact limiters consisting of wood, encapsulated by steel sheets. These impact limiters shall ensure that the transport casks meet the mechanical and thermal IAEA regulatory test requirements. According to the accident conditions of transport it is mandatory to expose the specimens to a cumulative effect by mechanical and thermal impacts. The mechanical tests consist of a free drop from 9 m onto a flat unyielding target and a 1 m drop onto a puncture bar. After damage caused by mechanical test sequences the package has to withstand a severe fire scenario. Corresponding to the IAEA advisory material it is required that the impact attitudes for the 9 m drop test and for the puncture test have to be such as to produce maximum damage, taking into account the thermal test. Moreover, any damage, which would give rise to increased radiation or loss of containment or affect the confinement system after the thermal test, should be considered. During and following the thermal test, the specimen shall not be artificially cooled and any combustion of materials of the package shall be permitted to proceed naturally. Different works from the French Institute for Radiological Protection and Nuclear Safety (IRSN) and BAM show that additional energy supply from a pre-damaged impact limiter to the cask could occur. This effect should be considered within the safety assessment of the containment. Thermal effects at the closure system of the cask, which might result in an elevated activity release, have to be excluded. BAM conducted small scale tests with wood filled metal buckets showing continuing combustion processes during the cooling down phase. These test results are presented. As not much is known about smouldering processes in wood filled impact limiters, it is highly complex to define pre-damage of impact limiters, which are conservative, regarding the maximum damaging energy flow from the impact limiter to the containment system. More research has to be done to develop models to examine the effects of smouldering impact limiters on the containment of packages for the transport of radioactive material. Aspects of assessment and its difficulties are shown. BAM as a competent authority for the approval of transport casks for radioactive material in Germany operates the test facilities to examine the issue of mechanical damage, combustion and heat transfer for such kind of package systems. For this purpose the knowledge from real drop tests with casks of a mass partly over 100 tons was transferred to a test application. A thermal test will take place with a wood filled test specimen with a diameter of about 2.3 meters. The aim is to understand the phenomena of smouldering under the consideration of relevant regulatory boundary conditions. The process of smouldering is described with regard to the requirements in the thermal assessment of safety of packages for the transport of radioactive material. Requirements concerning the pre-damage of packages for the maximum damage of impact limiters are discussed. Parameters influencing the smouldering process are identified. T2 - WM 2017 Conference CY - Phoenix, AZ, USA DA - 05.03.2017 KW - Fire test KW - Wood KW - Combustion KW - Smouldering KW - Smoldering KW - Impact limiter KW - Shock absorber KW - Thermal test KW - IAEA PY - 2017 SN - 978-0-9828171-6-2 SP - 1 AN - OPUS4-40157 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Linnemann, Konrad A1 - Ballheimer, Viktor A1 - Sterthaus, Jens A1 - Rolle, Annette A1 - Wille, Frank A1 - Nasyrow, R. A1 - Papaioannou, D. A1 - Vlassopoulos, E. A1 - Rondinella, V. T1 - Analysis of parameters affecting the bending behavior of spent fuel rods N2 - Transport packages for spent nuclear fuel have to be assessed with respect to specific transport conditions which are defined in the regulations of the International Atomic Energy Agency. The physical state of the spent fuel and the fuel rod cladding as well as the geometric configuration of the fuel assemblies are important inputs for the evaluation of the package capabilities under these conditions. Cracks or failures in the fuel rod cladding can cause the release of gas, volatiles or fuel particles into the cavity. The amount of substances in the cavity has to be considered in the assessment of the activity release and criticality safety. The mechanical analysis of the compound system formed by the fuel rod cladding and the spent fuel pellets is very difficult due to the limited knowledge of the material properties and the insufficient understanding of the interaction between pellets and cladding and between adjacent pellets. The variation of fuel assembly properties regarding cladding material, burn-up and the history of usage makes reliable predictions of the fuel rod behavior even harder. For a better understanding about the behavior of spent fuel rods, JRC-ITU and BAM have started a joint research project. In this context, JRC-ITU has developed a test device which allows quasi-static 3-point-bending test on fuel rod segments in the hot cell. The loads are applied with respect to the boundary conditions of the activity release assessment. This paper deals with the numerical calculation of a single fuel rod segment under bending load. The aim is to identify the governing mechanical parameters by the variation of constitutive assumptions, contact conditions, inner constraints, etc. This knowledge helps for the interpretation of the experimental results. Furthermore, the improved understanding about the behavior of the cladding-pellets system will be beneficial for the assessment of spent fuel transport conditions. T2 - PATRAM 2016 - 18th International symposium on the packaging and transportation of radioactive materials CY - Kobe, Japan DA - 18.09.2016 KW - Transport packages KW - Finite element methods KW - Spent fuel assessment PY - 2016 SP - Paper 2012, 1 AN - OPUS4-40000 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Nasyrow, R. A1 - Papaioannou, D. A1 - Rondinella, V. A1 - Vlassopoulos, E. A1 - Linnemann, Konrad A1 - Ballheimer, Viktor A1 - Sterthaus, Jens A1 - Rolle, Annette A1 - Wille, Frank A1 - Caruso, St. T1 - Bending test device for mechanical integrity studies of spent nuclear fuel rods N2 - This paper presents data obtained from experiments performed using a bending test set-up developed at the Joint Research Centre (JRC) – Karlsruhe, for spent fuel segment testing. Adjustable sample holders, loading modes and other experimental conditions can be im- plemented in the experiments to study the effects of different deformation ranges up to cladding failure. The experimental set-up has been adapted to hot cell remote controlling and has a modular configuration, which allows manual and motor-driven loading option. The device has been calibrated on hydrogenated, unirradiated cladding tube segments filled with alumina pellets. The final application of present set-up is to test non-defueled spent fuel rod segments, pressurized to the original spent fuel rod pressure level. The range of applicability of this device, the scope of the experimental program and the first results from actual bending tests will be discussed. T2 - PATRAM 2016 - 18th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Kobe, Japan DA - 18.09.2016 KW - Hot cell testing KW - Spent fuel assessment KW - Transport packages PY - 2016 SP - Paper 6023, 1 AN - OPUS4-40002 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Ballheimer, Viktor A1 - Quercetti, Thomas A1 - Sterthaus, Jens T1 - Package design assessment aspects of gaps between content and lid N2 - Type B(U) packages for the transport of radioactive material have to withstand accident conditions of transport defined in the regulations of the IAEA International Atomic Energy Agency in form of different mechanical (drop) tests with a subsequent thermal test. According to the regulatory requirements the orientation of the package in drop tests shall be such to cause the most damaged state in the components performing the safety functions. For the package lid system a 9-m drop onto the unyielding target with lid side downwards is often the most damaging orientation. The impact loads acting on the lid in this orientation result mainly from interaction between lid and internal content. In case of a movable content its impact onto the inner side of the lid can cause additional load peaks on the lid and the lid bolts. The intensity of the internal collision depends on the position of content relating to lid at the time of package first contact with target. Due to physical limitations an axial gap, which could be set in “pre-drop” configuration of package or which could spontaneously appear during the drop test, usually does not cover the maximum size possible in specific package design. In this context, the combination of drop tests with post-test analysis can be helpful to better estimate the effect of internal impact. The paper summarized some aspects of this issue based on the BAM experience in the design assessment of Type B(U) transport packages. Additionally the paper shall support applicants in German approval procedures to reduce rounds of questions and ensure delivery of reliable safety case documents to the authorities. International discussions of this issue at the IAEA and a joint proposal by France and Germany to improve Advisory Material text will be introduced. T2 - 18th International Symposium on the Packaging and Transportation of Radioactive Materials PATRAM 2016 CY - Kobe, Japan DA - 18.09.2016 KW - Lid behavior KW - Transport package KW - Secondary impact KW - Inventaraufprall KW - Fallprüfung PY - 2016 SP - 1 EP - 10 AN - OPUS4-37548 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Package design assessment aspects of gaps between content and lid N2 - Type B(U) packages for the transport of radioactive material have to withstand accident conditions of transport defined in the regulations of the IAEA International Atomic Energy Agency in form of different mechanical (drop) tests with a subsequent thermal test. According to the regulatory requirements the orientation of the package in drop tests shall be such to cause the most damaged state in the components performing the safety functions. For the package lid system a 9-m drop onto the unyielding target with lid side downwards is often the most damaging orientation. The impact loads acting on the lid in this orientation result mainly from interaction between lid and internal content. In case of a movable content its impact onto the inner side of the lid can cause additional load peaks on the lid and the lid bolts. The intensity of the internal collision depends on the position of content relating to lid at the time of package first contact with target. Due to physical limitations an axial gap, which could be set in “pre-drop” configuration of package or which could spontaneously appear during the drop test, usually does not cover the maximum size possible in specific package design. In this context, the combination of drop tests with post-test analysis can be helpful to better estimate the effect of internal impact. The paper summarized some aspects of this issue based on the BAM experience in the design assessment of Type B(U) transport packages. Additionally the paper shall support applicants in German approval procedures to reduce rounds of questions and ensure delivery of reliable safety case documents to the authorities. International discussions of this issue at the IAEA and a joint proposal by France and Germany to improve Advisory Material text will be introduced. T2 - 18th International Symposium on the Packaging and Transportation of Radioactive Materials PATRAM 2016 CY - Kobe, Japan DA - 18.09.2016 KW - Impact behavior KW - Fallprüfung KW - Inventarverhalten KW - Secondary impact KW - Spalte PY - 2016 AN - OPUS4-37549 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Komann, Steffen A1 - Wille, Frank A1 - Droste, Bernhard T1 - Transport von Großkomponenten aus der Stilllegung kerntechnischer Anlagen in Deutschland-Erfahrungen bei der mechanischen Intergritätsbewertung N2 - Im Zuge des Rückbaus kerntechnischer Anlagen in Deutschland müssen u. a. Großkomponenten (Dampferzeuger, Reaktordruckbehälter) über öffentliche Verkehrswege in Zwischenlager transportiert werden, in denen sie zerlegt bzw. zwischengelagert werden. Da es sich hierbei um aktivierte und/oder oberflächenkontaminierte Objekte handelt, ist eine Begutachtung unter Berücksichtigung der Gefahrgutbeförderungsvorschriften notwendig. In den meisten Fällen handelt es sich um oberflächenkontaminierte Gegenstände, welche als Industrieversandstücke des Typs IP-2 zu befördern sind. KW - Zwischenlagerung KW - Entsorgung KW - Kerntechnik KW - Radioaktive Stoffe PY - 2016 SN - 0005-6650 VL - Mai 2016 IS - Band 91 SP - 188 EP - 192 PB - Springer VDI Verlag CY - Düsseldorf AN - OPUS4-36852 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Droste, Bernhard A1 - Komann, Steffen A1 - Wille, Frank A1 - Rolle, Annette A1 - Probst, Ulrich A1 - Schubert, Sven T1 - Consideration of aging mechanism influence on transport safety of dual purpose casks for spent nuclear fuel of HLW N2 - When storage of spent nuclear fuel or high level waste is carried out in dual purpose casks (DPC), the effects of aging on safety relevant DPC functions and properties have to be managed in a way that a safe transport after the storage period of several decades is capable and can be justified and certified permanently throughout that period. The effects of aging mechanisms (e.g. radiation, different corrosion mechanisms, stress relaxation, creep, structural changes and degradation) on the transport package design safety assessment features have to be evaluated. Consideration of these issues in the DPC transport safety case will be addressed. Special attention is given to all cask components that cannot be directly inspected or changed without opening the cask cavity, like the inner parts of the closure system and the cask internals, like baskets or spent fuel assemblies. The design criteria of that transport safety case have to consider the operational impacts during storage. Aging is not the subject of technical aspects only but also of ‘intellectual’ aspects, like changing standards, scientific/technical knowledge development and personal as well as institutional alterations. Those aspects are to be considered in the management system of license holders and in appropriate design approval update processes. The paper addresses issues that are subject of an actual International Atomic Energy Agency TECDOC draft 'Preparation of a safety case for a dual purpose cask containing spent nuclear fuel'. KW - Transport KW - Storage KW - Spent fuel KW - High-level waste KW - Aging KW - Metal seals KW - Transport and storage casks KW - Spent nuclear fuel KW - Aging mechanisms KW - Corrosion KW - Safety assessment PY - 2014 DO - https://doi.org/10.1179/1746510914Y.0000000070 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 25 IS - 3-4 SP - 105 EP - 112 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-33428 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Nehrig, Marko A1 - Wille, Frank A1 - Feldkamp, Martin A1 - Rolle, Annette T1 - Aspects of gas generation caused by residual water inside ILW packages N2 - Packages for intermediate level waste (ILW) often contain residual water besides the actual waste. The water either exists as obvious free water or it may be bound physically or chemically, e.g. as pore water. Water driven gas generation could occur by vaporisation and by radiolysis. Steam as the result of vaporisation causes an increasing pressure inside a package and can affect corrosion. Vaporisation and condensation processes itself change the thermal behaviour of the content especially during strongly unsteady thermal situations like accident fire situations. Radiolysis changes the chemical composition of the content which could cause an unexpected interaction, e.g. hydrogen embrittlement. Besides the pressure build-up the radiolysis of water generates hydrogen and oxygen, which can be highly flammable respectively explosive. The gas generation caused by vaporisation and radiolysis must be taken into account during the design and the safety assessment of a package. Pressure build-up, a changed thermal behaviour and content chemistry, and especially the risk of accumulation of combustible gases exceeding the limiting concentration for inflammability has to be considered in the safety assessment. Approaches to ensure the transportability of stored packages due to radiolysis will be discussed. T2 - RAMTRANS 2015 - 10th International conference on radioactive materials transport and storage CY - Oxford, UK DA - 19.05.2015 KW - Typ-B KW - ILW package KW - Gas generation KW - Radiolysis KW - Wet content PY - 2015 SP - 1 EP - 7 AN - OPUS4-33429 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Considereation of Gaps between Content and Lid within Package Design Assessment T2 - Ramtrans 2015, Conference on Radioactive Materials Transport and Storage CY - Oxforn, UK DA - 2015-05-19 PY - 2015 AN - OPUS4-33448 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kuschke, Christian A1 - Wille, Frank A1 - Ballheimer, Viktor A1 - Sterhaus, Jens A1 - Apel, Andreas T1 - Standards and guidelines for design of trunnions of RAM transport packages N2 - Trunnion systems of packages for the transport of radioactive materials have to guarantee the safe handling of package during crane operations (lifting, tilting) and to secure package tie down to the transport vehicle, if the trunnions are used as attachment points during transport. The design of trunnions is based on IAEA Regulations SSR-6, the supporting Advisory Material SSG-26 and further appropriate technical standards and/or guidelines. For package approval procedures in Germany the guideline BAM-GGR 012 has to be applied. If a package is handled inside a german nuclear power plant, the standard KTA 3905 has to be fulfilled additionally. In this paper the requirements of KTA 3905 concerning the trunnion systems as load attaching points (LAP) are discussed in connection with the recommendations in the guideline BAM-GGR 012. This guideline is prepared at BAM Federal Institute for Materials Research and Testing for analysis and assessment of bolted lid and trunnion systems of Type B(U) transport packages. The Quality assurance questions concerning trunnion systems are discussed as well. T2 - WM 2017 conference CY - Phoenix, USA DA - 05.03.2017 KW - RAM transport package KW - Trunnions PY - 2017 SP - Paper 17365, 1 EP - 12 AN - OPUS4-40815 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Neumann, Martin A1 - Eisenacher, Germar A1 - Schönfelder, Thorsten A1 - Wille, Frank T1 - Development of a material model for the crush of spruce wood N2 - Typical transport packages used in Germany are equipped with wooden impact limiting devices. In this paper we give an overview of the latest status regarding the development of a finite element material model for the crush of spruce wood. Although the crush of wood – mainly in longitudinal direction – is a phenomenon governed by macroscopic fracture and failure of wood fibres we smear fracture and failure mechanisms over the continuous voume. In first step we altered an existing LS-DYNA material model for foams, which considers an ellipse shaped yield surface written in terms of the first two stress invariants. The evolution of the yield surface in the existing model depends on the volumetric strain only. For the use with spruce wood, we modified the existing material model to consider the deviatoric strain for the evolution of the yield surface as well. This is in accordance with the results of crush tests with spruce wood specimens, where the crushing deformation was rather deviatoric for uniaxial stress states and rather volumetric for multiaxial stress states We rate the basic idea of this approach to be reasonable, though other problems exist regarding the shape of the yield surface and the assumption of isotropic material properties. Therefore we developed a new transversal isotropic material model with two main directions, which considers different yield curves according to the multiaxiality of the stress state via a multi-surface yield criterion and a non-associated flow rule. The results show the ability to reproduce the basic strength characteristics of spruce wood. Nevertheless, problems with regularization etc. show that additional investigations are necessary. T2 - ASME PVP 2017 CY - Waikaloa, HY, USA DA - 16.07.2017 KW - Chrush KW - Spruce KW - Wood KW - FEM PY - 2017 SN - 978-0-7918-5802-8 VL - 7 SP - Article UNSP V007T07A037, 1 EP - 9 AN - OPUS4-41613 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Design assessment of spent fuel and HLW transport casks T2 - 49. INMM-Konferenz CY - Nashville, TN, USA DA - 2008-07-13 PY - 2008 AN - OPUS4-17759 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Reflektion der verkehrsrechtlichen Bauartprüfung des CASTOR HAW 28M aus Sicht der zuständigen Behörde T2 - Jahrestagung Kerntechnik 2010 CY - Berlin, Germany DA - 2010-05-04 PY - 2010 AN - OPUS4-21321 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Droste, Bernhard T1 - Mechanical safety analysis for high burn-up spent fuel assemblies under accident transport conditions KW - Fuel rods KW - High burn-up KW - Accident conditions KW - Mechanical analysis KW - Failure KW - Accident transport conditions PY - 2010 DO - https://doi.org/10.1179/174650910X12913756439836 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 21 IS - 4 SP - 212 EP - 217 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-23051 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Wille, Frank A1 - Droste, Bernhard A1 - Müller, Karsten A1 - Zencker, Uwe T1 - Mechanical design assessment approaches of actual spent fuel and HLW transport package designs KW - Package design KW - Spent fuel KW - High level waste KW - Design computations KW - Testing KW - Design testing KW - Transport packages KW - Numerical analysis KW - Mechanical analysis PY - 2010 DO - https://doi.org/10.1179/174650910X12913738452796 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 21 IS - 4 SP - 203 EP - 207 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-23052 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Kernbrennstoffkreislauf - Ver- und Entsorgung deutscher Kernkraftwerke T2 - Blockseminar "Strucktur der Materie - Kernphysik", TFH Wildau CY - Wildau, Germany DA - 2008-06-17 PY - 2008 AN - OPUS4-17545 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - History of Implementation of Crush-Test to IAEA Regulations T2 - ASME Pressure Vessels and Piping Conference 2009 CY - Prague, Czech Republic DA - 2009-07-26 PY - 2009 AN - OPUS4-20956 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Droste, Bernhard A1 - Müller, Karsten A1 - Zencker, Uwe T1 - Recent experiences in mechanical design assessment of spent fuel and HLW casks by competent authority in Germany - 10093 T2 - WM2010 Conference CY - Phoenix, Arizona, USA DA - 2010-03-07 KW - Design testing KW - Transport packages KW - Numerical analysis PY - 2010 SP - 1 EP - 10 AN - OPUS4-21106 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Recent Experiences in Mechanical Design Assessment of Spent Fuel and HLW Casks by Competent Authority in Germany T2 - Waste Management Conference 2010 CY - Phoenix, AZ, USA DA - 2010-03-07 PY - 2010 AN - OPUS4-21117 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank T1 - Transport von Großkomponenten aus der Stilllegung kerntechnischer Anlagen - Erfahrungen bei der mechanischen Bewertung T2 - Jahrestagung Kerntechnik 2010 CY - Berlin, Deutschland DA - 2010-05-04 KW - Kerntechnische Anlagen KW - Rückbau KW - Mechanische Bewertung KW - Großkomponenten KW - Numerische Analyse PY - 2010 IS - Sektion 7 / 706 SP - 1 EP - 5 CY - Berlin AN - OPUS4-21239 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Droste, Bernhard A1 - Zencker, Uwe A1 - Komann, Steffen T1 - Reflektion der verkehrsrechtlichen Bauartprüfung des CASTOR HAW 28M aus Sicht der zuständigen Behörde BAM T2 - Jahrestagung Kerntechnik 2010 CY - Berlin, Deutschland DA - 2010-05-04 KW - Bauartprüfung KW - Fallprüfungen KW - Transportbehälter KW - FEM PY - 2010 SP - 1 EP - 5 CY - Berlin AN - OPUS4-21241 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Wille, Frank A1 - Ballheimer, Viktor A1 - Musolff, André T1 - Numerische Analyse der 1-m-Fallprüfung auf einen Stahldorn T2 - Jahrestagung Kerntechnik 2010 CY - Berlin, Deutschland DA - 2010-05-04 KW - IAEO KW - 1-m-Fallprüfung KW - Stahldorn KW - Numerische Analyse KW - Verifizierung PY - 2010 IS - Sektion 5 / 505 SP - 1 EP - 5 CY - Berlin AN - OPUS4-21292 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Nehrig, Marko A1 - Wille, Frank A1 - Quercetti, Thomas A1 - Masslowski, Jörg-Peter A1 - Droste, Bernhard A1 - Pope, R. T1 - Historical view and experiences with the crush test for light weight packages T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Crush test KW - IAEA regulations KW - Type-B PY - 2010 SP - 1 EP - 8(?) AN - OPUS4-23303 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Aktuelle Aspekte der Qualitätssicherung bei Herstellung und Betrieb zulassungspflichtiger und nicht zulassungspflichtiger Verpackungen für Versandstücke zur Beförderung radioaktiver Stoffe T2 - BMVBS Informationsaustausch zur verkehrsrechtlichen Aufsicht CY - Bonn, Germany DA - 2011-03-01 PY - 2011 AN - OPUS4-23457 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Qiao, Linan A1 - Zencker, Uwe A1 - Wille, Frank A1 - Musolff, André T1 - Numerical simulation of 9 meter drop of a transport and storage cask with aluminium impact limiter N2 - For the purpose of numerical simulation of 9 meter drop of a transport and storage cask with aluminium impact limiter, an elastic-incremental plastic material model with strain rate hardening acc. to Cowper-Symonds is used for the development of isothermal as well as adiabatic stress-strain relations of aluminium from the compression test at constant ambient temperature. After that, two different simulation strategies are compared. At first, the drop test is calculated fully coupled, i.e. with isothermal stress-strain relations and possible heat generation in the material. Then the drop test is recalculated in a very simplified manner with adiabatic stress-strain relations from the compression test in an isothermal simulation. Both calculation strategies show similar results in the investigated load scenario. T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Drop test KW - Spent fuel transport cask KW - Finite element calculation KW - Dynamic simulation KW - Impact limiter KW - Aluminium PY - 2010 SP - 1-8 (Monday-T16-117) AN - OPUS4-23707 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Aspekte der Konzeption der sicherheitstechnischen Nachweisführung - Mechanische Nachweisführung innerhalb der verkehrsrechtlichen Bauartprüfung T2 - 2. RAM-Behältersicherheitstage 2010, BAM CY - Berlin, Germany DA - 2010-03-18 PY - 2010 AN - OPUS4-23148 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank T1 - Transport of large nuclear power plant components - experiences in mechanical design assessment T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Mechanische Bewertung KW - Radioaktives Material KW - Numerische Analyse PY - 2010 SP - 1 EP - 8(?) AN - OPUS4-23383 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Droste, Bernhard A1 - Müller, Karsten A1 - Zencker, Uwe T1 - Mechanical design assessment approaches of actual spent fuel and HLW transport package designs T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Design testing KW - Transport packages KW - Numerical analysis PY - 2010 SP - 1 EP - 8(?) AN - OPUS4-23590 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank A1 - Rittscher, D. T1 - Mechanical assessment of large nuclear power plant components in transport licensing T2 - 49th INMM Annual Meeting CY - Nashville, TN, USA DA - 2008-07-13 KW - Rückbau KW - Kerntechnische Anlagen KW - Großkomponenten KW - Transport KW - Mechanik PY - 2008 SP - 1 EP - 8 AN - OPUS4-18440 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Aktuelle Aspekte der sicherheitstechnischen Nachweisführung bei der gefahrgutrechtlichen Bauartprüfung T2 - 1. RAM-Behältersicherheitstage 2009 CY - Berlin-Horstwalde, Germany DA - 2009-03-26 PY - 2009 AN - OPUS4-19179 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Mechanical Safety Assessment Strategy for Ductile Cast Iron Casks demonstrated by means of IAEA 1m Puncture Drop Test T2 - RAMTRANS 2009 CY - Manchester, England DA - 2009-05-12 PY - 2009 AN - OPUS4-19361 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank T1 - Transport of large nuclear power plant components - experiences in mechanical design assessment T2 - Ramtransport 09 - 8th International Conference on radioactive materials transport 2009 CY - Manchester, Great Britain DA - 2009-05-12 KW - Rückbau KW - Kerntechnische Anlagen KW - Großkomponenten KW - Transport KW - Mechanik PY - 2009 IS - Session 5 / PAP 12 SP - 1 EP - 8 AN - OPUS4-20557 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank T1 - Transport of large nuclear power plant components: experiences in mechanical design assessment N2 - In the course of decommissioning of power plants in Germany large nuclear components (steam generator, reactor pressure vessel) must be transported over public traffic routes to interim storage facilities, where they are dismantled or stored temporarily. Since it concerns surface contaminated objects or low specific activity materials, a safety evaluation considering the IAEA transport regulations mainly for industrial packages (type IP-2) is necessary. For these types of industrial packages the requirements from normal transport conditions are to be covered for the mechanical proof. For example, a free drop of the package from a defined height, in dependence of its mass, onto an unyielding target, and a stacking test are required. Since physical drop tests are impossible generally due to the singularity of such 'packages', a calculation has to be performed, preferably by a complex numerical analysis. The assessment of the loads takes place on the basis of local stress distributions, also with consideration of radiation induced brittleness of the material and with consideration of recent scientific investigation results. Large nuclear components have typically been transported in an unpackaged manner, so that the external shell of the component provides the packaging wall. The investigation must consider the entire component including all penetration areas such as manholes or nozzles. According to the present IAEA regulations the drop position is to be examined, which causes the maximum damage to the package. In the case of a transport under special arrangement a drop only in an attitude representing the usual handling position (administratively controlled) is necessary. If dose rate values of the package are higher than maximum allowable values for a public transport, then it is necessary that additional shielding construction units are attached to the large component. KW - Radioactive material KW - Large components KW - Decommissioning KW - Nuclear power plant KW - Mechanical assessment KW - Numerical analysis PY - 2009 DO - https://doi.org/10.1179/174650909X12543085665266 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 20 IS - 4 SP - 149 EP - 153 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-20820 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Mechanical Design Assessment Approaches of Actual Spent Fuel and HLW Transport Package Designs T2 - 16th International Symposium of the Packaging and transportation of Radioactive Materials PATRAM CY - London, England DA - 2010-10-03 PY - 2010 AN - OPUS4-22132 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Droste, Bernhard T1 - Mechanical safety analysis for high burn-up spent fuel assemblies under accident transport conditions T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Accident transport conditions KW - Fuel rods KW - High burn-up KW - Failure PY - 2010 SP - 1 EP - 8(?) AN - OPUS4-23760 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Zencker, Uwe A1 - Weber, Mike A1 - Wille, Frank T1 - Finite element mesh design of a cylindrical cask under puncture drop test conditions N2 - Transport casks for radioactive materials have to withstand the 9 m drop test, 1 m puncture drop test and dynamic crush test with regard to the mechanical requirements according to the IAEA regulations. The safety assessment of the package can be carried out on the basis of experimental investigations with prototypes or models of appropriate scale, calculations, by reference to previous satisfactory safety demonstrations of a sufficiently similar nature or a combination of these methods. Computational methods are increasingly used for the assessment of mechanical test scenarios. However, it must be guaranteed that the calculation methods provide reliable results. Important quality assurance measures at BAM are given concerning the preparation, run and evaluation of a numerical analysis with reference to the appropriate guidelines. Hence, a successful application of the finite element method requires a suitable mesh. An analysis of the 1 m puncture drop test using successively refined finite element meshes was performed to find an acceptable mesh size and to study the mesh convergence using explicit dynamic finite element codes. The finite element model of the cask structure and the puncture bar is described. At the beginning a coarse mesh was created. Then this mesh was refined in two steps. In each step the size of the elements was bisected. The deformation of the mesh and the stresses were evaluated dependent on the mesh size. Finally, the results were extrapolated to an infinite fine mesh or the continuous body, respectively. The uncertainty of the numerical solution due to the discretization of the continuous problem is given. A safety factor is discussed to account for the uncertainty. The calculation results are compared with experimental data from a puncture drop test with a half-scale model of a cylindrical cask. This paper supports the convergence studies of the Task Group on Computational Modeling for Explicit Dynamics reporting to the ASME BPV Code Working Group on Design Methodology. T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - IAEA puncture drop test KW - Numerical analysis KW - Verification KW - Explicit dynamics PY - 2010 SP - 1 EP - 8 (Session T45 / Paper 219) AN - OPUS4-23843 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Völzke, Holger A1 - Wille, Frank A1 - Wieser, Günter A1 - Quercetti, Thomas T1 - Transport and Storage Cask Safety Assessment - Drop Tests and Numerical Calculations T2 - Waste Management Symposium 2006 CY - Tucson, Az. USA DA - 2006-02-26 KW - Transportbehälter KW - Lagerbehälter KW - Radioaktive Stoffe KW - Fallversuch KW - FEM KW - Sicherheit PY - 2006 SP - 12 pages CY - Tucson AN - OPUS4-12536 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Wille, Frank A1 - Ballheimer, Viktor A1 - Droste, Bernhard T1 - Suggestions for correct performance of IAEA 1 m puncture bar drop test with reduced scale packages considering similarity theory aspects T2 - PATRAM 2007, 15th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Miami, Florida, USA DA - 2007-10-21 KW - Similarity theory KW - Puncture bar drop test KW - Reduced scale packages KW - Drop height adaption PY - 2007 DO - https://doi.org/10.1179/174651007X191233 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 18 IS - 2 SP - 111 EP - 116 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-15891 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Suggestion for correct performance of IAEA 1 m puncture bar drop test with reduced-scale packages considereing similarity aspects T2 - PATRAM 2007 CY - Miami, FL, USA DA - 2007-10-21 PY - 2007 AN - OPUS4-15896 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - State-of-the-art assessment of package design safety analyses T2 - PATRAM 2007 The 15th International Symposium on the Packaging and Transportation of Radioactive Material CY - Miami, FL, USA DA - 2007-10-21 PY - 2007 AN - OPUS4-16014 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Ballheimer, Viktor A1 - Droste, Bernhard T1 - Suggestions for correct performance of IAEA 1 m puncture bar drop test with reduced-scale packages considering similarity theory aspects T2 - PATRAM 2007, 15th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Miami, Florida, USA DA - 2007-10-21 KW - Similarity KW - Puncture Bar Drop Test KW - Scale Models KW - IAEA Regulations PY - 2007 IS - CD-ROM SP - 1 EP - 8 PB - Institute of Nuclear Materials Management AN - OPUS4-16104 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Mechanical safety assessment with regards to reduced-scale model testing - accompanying material / component tests and numerical calculations T2 - US NRS/BAM Workshop CY - Rockville, MD, USA DA - 2007-10-18 PY - 2007 AN - OPUS4-16231 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Wolff, Dietmar A1 - Droste, Bernhard A1 - Völzke, Holger T1 - Requirements for transport packages after interim storage N2 - In Germany the concept of dry interim storage of spent fuel and vitrified high active waste in dual purpose metal casks is implemented (currently for periods of up to 40 years). The casks being used have an approved package design in accordance with the international transport regulations. The license for dry storage is granted on the German Atomic Energy Act with respect to the guidelines of the German Reactor Safety Commission for dry interim storage of spent fuel. Currently part of the assessment process of the cask design for transport on public routes is to evaluate the suitability of the used materials with respect to their properties and their compatibility for possible transport periods. For transport on public routes during or after long term interim storage periods, it has to be ensured that the transport and storage casks fulfill the specifications of the transport approval or other sufficient properties which satisfy the proofs for the compliance of the safety objectives at that time. This paper describes the state-of-the-art technology in Germany and points out arising prospective challenges and which related questions have to be answered in future. Present research and knowledge concerning the long term behavior of transport and storage cask components (such as gaskets and shielding components) have to be consulted as well as experiences from interim cask storage operations. Challenges in the safety assessment are therefore i.e. the behavior of aged metal and elastomeric gaskets under IAEA test conditions to ensure that the results of drop tests can be transferred to the compliance of the safety objectives at the time of transport after the interim storage period (aged package). Assessment methods for the material compatibility, the behavior of fuel assemblies and the aging behavior of shielding parts (e. g. influence of radiation) are further issues as well. T2 - 52nd INMM Annual meeting CY - Palm Desert, CA, USA DA - 17.07.2011 KW - Transport package design KW - Interim storage KW - Inspections KW - Log term behavior KW - Dual purpose casks PY - 2011 SP - 1 EP - 9 AN - OPUS4-24172 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Requirements on Transport Packages after Interim Storage - Current Approach and Perspectives T2 - U.S. NRC Workshop CY - Washington, D.C., USA DA - 2011-07-11 PY - 2011 AN - OPUS4-24126 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Requirements on Transport Packages after Interim Storage - Curent Spproach and Perspectives T2 - NMM 52nd Annual Meeting CY - Palm Desert, CA, USA DA - 2011-07-17 PY - 2011 AN - OPUS4-24127 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Weber, Mike A1 - Wille, Frank A1 - Ballheimer, Viktor A1 - Musolff, André T1 - From experiment to an appropriate finite element model-safety assessment for ductile cast iron casks demonstrated by means of IAEA puncture drop test N2 - In the approval procedure of transport packages for radioactive materials, the competent authority mechanical and thermal safety assessment is carried out in Germany by BAM Federal Institute for Materials Research and Testing. The combination of experimental investigations and numerical calculations in conjunction with materials and components testing is the basis of the safety assessment concept of the BAM. Among other mechanical test scenarios a 1 meter drop test onto a steel bar has to be considered for hypothetical accident conditions of Type B packages according to IAEA regulations. Within the approval procedure for the new German package design of the HLW cask CASTOR® HAW 28M, designed by GNS Gesellschaft für Nuklear-Service Germany, a puncture drop test was performed with a half-scale model of the cask at -40°C. For independent assessment and to control the safety analysis presented by applicant, BAM developed a complex finite element model for a dynamical ABAQUS/ExplicitTM analysis. This paper describes in detail the use of the finite element (FE) method for modeling the puncture drop test within an actual assessment strategy. At first investigations of the behaviour of the steel bar are carried out. Different friction coefficients and the material law of the bar are analysed by using a 'rigid-body' approximation for the cask body. In the next step a more detailed FE model with a more realistic material definition for the cask body is developed. Strain verification is possible by results of the strain gauges located at the relevant points of the cask model. The influence of the finite element meshing is described. Finally, the verified FE half-scale model is expanded to full-scale dimension. Scaling effects are analysed. The model is used for safety assessment of the package to be approved. T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - IAEA KW - Puncture drop test KW - Numerical analysis KW - Verification PY - 2010 SP - 1-8 (Tuesday/T19/26-115) AN - OPUS4-23928 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Eisenacher, Germar A1 - Scheidemann, Robert A1 - Neumann, Martin A1 - Wille, Frank A1 - Droste, Bernhard T1 - Approach for a finite element material model for wood for application in mechanical safety cases of transport packages N2 - BAM Federal Institute for Materials Research and Testing is the competent authority for mechanical safety assessment of transport packages for radioactive material in Germany. The further development of state-of-the-art technology concerning assessment methods is essential for a qualified work of involved designers and authority experts. The paper gives an example of current development done to improve understanding and modeling capabilities of wood filled impact limiter. In order to reduce the loads applied to the package containment, which result from regulatory drop tests, most packages are protected by energy dissipating impact limiter. Wood, encapsulated by steel sheets, is one of the materials typically used for energy dissipation in these impact limiter. Very often, mechanical safety cases regarding the 9 m drop test are performed computationally, where it is essential to use reliable and verified computational methods and models. In this context, the paper presents an approach for a finite element material model for wood. Thereby, the mechanical behavior of wood under compression loading is the focus of the development work. Additionally, material orientation as well as strain rate, temperature and lateral constraint may vary. A large number of experiments, particularly compression tests, was designed and performed to establish an adequate experimental database for modeling verification. The experimental results enabled the derivation of necessary requirements: The material model has to take into account strain rate and temperature dependencies as well as the anisotropic characteristics of the material, a proper yield criterion, flow rule and hardening law. Such a material model is currently not available in established commercial dynamic finite element codes. Thus it is necessary to create a user-defined material model considering the mentioned requirements. A first step was done by determining a yield surface as well as detecting flow and hardening mechanisms from experimental force-deflection curves. In a next step the LS-DYNA material model MAT_75 was altered according to conclusions of former BAM development work, regarding the modeling of post-peak softening as a function of lateral constraint. Future research will contain the further development, implementation and verification of a material model for wood. T2 - 52nd INMM Annual meeting CY - Palm Desert, CA, USA DA - 17.07.2011 KW - Impact limiter KW - Material model wood KW - Transport package PY - 2011 SP - 1 EP - 8 AN - OPUS4-24236 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Zencker, Uwe A1 - Weber, Mike A1 - Wille, Frank T1 - Finite element mesh design of a cylindrical cask under puncture drop test conditions N2 - Transport casks for radioactive materials have to withstand the 9 m drop test, 1 m puncture drop test and dynamic crush test with regard to the mechanical requirements according to the IAEA regulations. The safety assessment of the package can be carried out on the basis of experimental investigations with prototypes or models of appropriate scale, calculations, by reference to previous satisfactory safety demonstrations of a sufficiently similar nature or a combination of these methods. Computational methods are increasingly used for the assessment of mechanical test scenarios. However, it must be guaranteed that the calculation methods provide reliable results. Important quality assurance measures at the Federal Institute for Materials Research and Testing are given concerning the preparation, run and evaluation of a numerical analysis with reference to the appropriate guidelines. Hence, a successful application of the finite element (FE) method requires a suitable mesh. An analysis of the 1 m puncture drop test using successively refined FE meshes was performed to find an acceptable mesh size and to study the mesh convergence using explicit dynamic FE codes. The FE model of the cask structure and the puncture bar is described. At the beginning a coarse mesh was created. Then this mesh was refined in two steps. In each step the size of the elements was bisected. The deformation of the mesh and the stresses were evaluated dependent on the mesh size. Finally, the results were extrapolated to an infinite fine mesh or the continuous body, respectively. The uncertainty of the numerical solution due to the discretisation of the continuous problem is given. A safety factor is discussed to account for the uncertainty. KW - IAEA puncture drop test KW - Cylindrical cask KW - FE mesh refinement KW - Explicit dynamics KW - Numerical analysis PY - 2011 DO - https://doi.org/10.1179/1746510911Y.0000000008 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 22 IS - 2 SP - 112 EP - 116 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-24016 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kuschke, Christian A1 - Neumeyer, Tino A1 - Wille, Frank A1 - Droste, Bernhard T1 - Mechanical/thermal package design safety assessment and manufacturing quality assurance of spent fuel transport cask NCS 45 T2 - WM2010 - 36th Annual radioactive waste management symposium CY - Phoenix, Arizona, USA DA - 2010-03-07 KW - Package design KW - Spent fuel transport cask KW - Safety assessment PY - 2010 SN - 978-0-9828171-0-0 SP - 1 EP - 13 CY - Tempe, AZ, USA AN - OPUS4-24344 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kuschke, Christian A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Komann, Steffen T1 - Mechanical assessment criteria of spent fuel assemblies basket design N2 - Packages for the transport of radioactive material are generally equipped with specific structures (basket) to support the radioactive content in defined position. The safety function of the basket depends on the kind of transported inventory. In case of transport cask for spent fuel, the basket design has to ensure the subcriticality of the fissile material in all conditions of transport in particular. Therefore the evaluation of structural integrity and neutron absorption capability of the basket is an important part of complete safety analysis. Sufficient heat transfer to maintain fuel assembly and cask temperature within allowable limits has to be verified as well. Corrosion resistance is an additional requirement on basket materials owing to contact with water during loading and unloading operations. Computational and experimental methods or their combination along with additional material and component tests can be used to analyse the mechanical and thermal basket behaviour under transport conditions defined in IAEA regulations. By deciding between the analysis methods, the design features (including material selection concept) as well as specific safety function should be accounted. In approval procedures of transport packages for radioactive materials, the competent authority mechanical and thermal safety assessment is carried out in Germany by BAM. Some questions of safety evaluation of basket designs are discussed in this paper based on the BAM experience within approval procedures. The paper focuses primarily on the mechanical behaviour of baskets with regard to the assumptions that have to be used in the criticality safety demonstration. The state of the art methodologies for computational basket stress and deformation analysis as well as for interpretation of drop tests results are presented. T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Transport conditions KW - Spent fuel assemblies KW - Basket design PY - 2010 SP - 1 EP - 7 (Thursday T36/92) AN - OPUS4-24345 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Kiyak, Yusuf A1 - Wille, Frank A1 - Zerbst, Uwe A1 - Weber, Mike A1 - Klingbeil, Dietmar T1 - Assessment of ductile cast iron fracture mechanics analysis within licensing of German transport packages N2 - In the approval procedure of transport packages for radioactive materials, the mechanical and thermal safety assessment is carried out in Germany by competent authority BAM. In recent years BAM was involved in several licensing procedures of new spent fuel and HLW package designs, where the cask body was made of Ductile Cast Iron (DCI). According to IAEA regulations package designs have to fulfill requirements for specific conditions of transport. Type B(U) packages must withstand the defined accident conditions of transport. The temperature range from -40°C up to the operational temperature has to be considered. For the cask material DCI, it is necessary to determine the brittle fracture behavior. The German guideline BAM-GGR 007 defines requirements for the fracture mechanics of DCI. Due to complex structure of the cask body and the dynamic loading a fracture mechanical assessment in an analytical kind is not always possible. Experience of recent design approval procedures show that the application of numerical calculations are applicable to determine the fracture mechanical load in the cask body. At the first step a numerical analysis has to be done to identify the loading state at the whole cask body. Secondly an analysis of a detail of the cask body is made considering the displacement boundary conditions of the global model. An artificial flaw is considered in this detailed model to calculate the fracture mechanical loading state. The finite element mesh was strongly refined in the area of flaw. The size of the artificial flaw is characterized by the ultrasonic inspection used for the quality assurance of the package. The applicant developed additional analysis tools for calculation of stress intensity factor and/or J-Integral. The assessment approach by BAM led to the decision to develop own tools to the possibility for independent check of the results. The paper describes the authority assessment approach for the DCI fracture mechanics analysis. The validation procedure incl. the development of own tools is explained. BAM developed a post-processor called JINFEM to determine the fracture mechanical loads. The regulatory 1 m puncture bar drop test is used to give an example of the assessment procedure. T2 - 52nd INMM Annual meeting CY - Palm Desert, CA, USA DA - 17.07.2011 KW - Mechanische Bewertung KW - Radioaktives Material KW - Numerische Analyse KW - Bruchmechanik PY - 2011 SP - 1 EP - 10 AN - OPUS4-25022 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Nehrig, Marko A1 - Wille, Frank A1 - Quercetti, Thomas A1 - Masslowski, Jörg-Peter A1 - Droste, Bernhard A1 - Pope, R. T1 - Historical view and experiences with crush test for light weight packages N2 - The crush test for light weight and low density type B packages was introduced for the first time into the 1985 edition of the International Atomic Energy Agency (IAEA) transport safety regulations. In the early 1970s, the need for an additional mechanical test besides or instead of the well known 9 m drop test was deliberated. Various authors and test facilities, including BAM and Sandia National Laboratories (SNL), were able to prove that the level of safety provided by IAEA drop and puncture tests in the regulations did not protect against dynamic crush forces to smaller packages. As early as the third PATRAM symposium held in 1971 (Richland, WA, USA), Robert F. Barker asked for '... a more strenuous crushing test for protecting small, light weight packages ...' BAM developed from research activities a proposal as to which types of packages should be subject to crush tests and how the crush tests should be performed, which was presented at the 5th PATRAM symposium held in 1978 (Las Vegas, NV, USA). At the IAEA, the possible need for a crush test was first mentioned in 1977. The subject for a discussion, besides the principal need for this test, was also the development of suitable set of crush test boundary conditions. It took more than four years of discussion until a dynamic crush test similar to today's test was recommended by experts to the IAEA regulatory revision panel. Finally, after a rigorous evaluation process in which also the boundary conditions were determined, the crush test was proposed to be incorporated into the IAEA regulations. BAM and SNL participated in the crush test development and implementation process right from the beginning in the early 1970s until its implementation in the IAEA regulations in 1985. Today, BAM performs crush test procedures according to para. 727(c) of TS-R-1, which have not been changed since their first implementation. Crush tests performed in 2002 at BAM will be discussed. These approval design tests were performed on birdcage pellet transport containers under normal and accident conditions according to the IAEA regulations. KW - Package testing KW - Crush testing KW - Regulations PY - 2011 DO - https://doi.org/10.1179/1746510911Y.0000000016 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 22 IS - 3 SP - 125 EP - 129 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-25349 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Weber, Mike A1 - Wille, Frank A1 - Ballheimer, Viktor A1 - Musolff, André T1 - From experiment to appropriate finite element model-safety assessment for ductile cast iron casks demonstrated by means of IAEA puncture drop test N2 - In the approval procedure of transport packages for radioactive materials, the competent authority mechanical and thermal safety assessment is carried out in Germany by BAM Federal Institute for Materials Research and Testing. The combination of experimental investigations and numerical calculations in conjunction with materials and components testing is the basis of the safety assessment concept of the BAM. Among other mechanical test scenarios, a 1 metre drop test onto a steel bar has to be considered for the application of the hypothetical accident conditions to Type B packages according to IAEA regulations. Within the approval procedure for the new German package design of the HLW cask CASTOR® HAW 28M, designed by GNS Gesellschaft für Nuklear-Service Germany, a puncture drop test was performed with a half-scale model of the cask at -40°C. For independent assessment and to control the safety analysis presented by the applicant, BAM developed a complex finite element (FE) model for a dynamical ABAQUS/ExplicitTM analysis. This paper describes in detail the use of the FE method for modelling the puncture drop test within an actual assessment strategy. At first, investigations of the behaviour of the steel bar were carried out. Different friction coefficients and the material law of the bar were analysed by using a 'rigid-body' approximation for the cask body. In the next step, a more detailed FE model with a more realistic material definition for the cask body was developed. The validation of calculated strains was carried out by comparison with the results of the strain gauges located at the relevant points of the cask model. The influence of the FE meshing is described. Finally, the validated FE half-scale model was expanded to full-scale dimension. Scaling effects were analysed. The model was used for safety assessment of the package to be approved. KW - IAEA puncture drop test KW - Cylindrical cask KW - Explicit dynamics KW - Scaled model KW - Numerical analysis KW - Validation KW - Safety assessment PY - 2011 DO - https://doi.org/10.1179/1746510911Y.0000000010 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 22 IS - 3 SP - 148 EP - 153 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-25356 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Mechanical Behaviour of High Burn-Up SNF under Normal and Accident Transport Conditions - Present Approaches and Perspectives T2 - PSAM 11 CY - Helsinki, Finland DA - 2012-06-25 PY - 2012 AN - OPUS4-26171 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank T1 - Regelwerksentwicklung für die Auslegung von lastanschlagpunkten T2 - 3. RAM-Behältersicherheitstage 2012, BAM CY - Berlin, Germany DA - 2012-03-22 PY - 2012 AN - OPUS4-26172 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -