TY - CONF A1 - Komann, Steffen A1 - Neumann, Martin A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Weber, Mike A1 - Qiao, Linan A1 - Droste, Bernhard T1 - Mechanical Assessment within Type B Packages Approval, The Application of Static and Dynamic Calculation Approaches T2 - 16th International Symposium on the Packaging and Transport of Radioactive Materials CY - London, England DA - 2010-10-03 PY - 2010 AN - OPUS4-22252 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Komann, Steffen A1 - Neumann, Martin A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Weber, Mike A1 - Qiao, Linan A1 - Droste, Bernhard T1 - Mechanical assessment within type B packages approval: application of static and dynamic calculation approaches N2 - This paper demonstrates exemplarily how numerical and experimental approaches can be combined reasonably in mechanical assessment of package integrity according to the IAEA regulations. The paper also concentrates on the question about how static mechanical approaches can be applied, and what their problems are in relation to dynamic calculation approaches. Under defined impact tests, which represent accident transport conditions, the package has to withstand impact loading, e.g. resulting from a 9 m free drop onto an unyielding target in sequence with a 1 m puncture drop test. Owing to the local character of the interaction between the puncture bar and the cask body, it is possible to develop a dynamic numerical model for the 1 m puncture drop which allows an appropriate simulation of the interaction area. Results from existing experimental drop tests with prototype or small scale cask models can be used for verification and validation of applied analysis codes and models. The link between analysis and experimental drop testing is described exemplarily by considering a regulatory 1 m puncture bar drop test onto the cask body of a recently approved German high level waste transport package. For the 9 m drop test of the package, it is difficult to develop a dynamic numerical model of the package due to the complexity of the interaction between cask body, impact limiters and unyielding target. Dynamic calculations require an extensive verification with experimental results. The simulation of a 9 m drop of a package with impact limiters is thereby often more complex than the simulation of a 1 m puncture drop onto the cask body. A different approximation method can be applied for the consideration of dynamic effects on the impact loading of the package. In a first step, maximum impact force and rigid body deceleration of the cask body during the impact process can be calculated with simplified numerical tools. This rigid body deceleration can subsequently be applied on a verified static numerical model. Dynamic effects, which cannot be covered by the static numerical analysis, have therefore to be considered by using an additional dynamic factor. The paper describes this approach exemplarily for a 9 m horizontal drop of a typical spent fuel cask design. KW - Package assessment KW - Package KW - Packaging KW - Structural analysis PY - 2011 DO - https://doi.org/10.1179/1746510911Y.0000000012 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 22 IS - 4 SP - 179 EP - 183 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-26448 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen T1 - Assessment of quality assurance measures for radioactive material transport packages not requiring competent authority design approval T2 - PATRAM Conference 2013 CY - San Francisco, CA, USA DA - 2013-08-18 PY - 2013 AN - OPUS4-29753 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen T1 - Verpackungen für den Transport radioaktiver Stoffe T2 - 14. Gefahrgut-Technik-Tage CY - Berlin, Germany DA - 2013-11-11 PY - 2013 AN - OPUS4-29754 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen T1 - Assessment of ductile cast iron fracture mechanics analyses within licensing of german transport packages T2 - Patram Conference 2013 CY - San Francisco, CA, USA DA - 2013-08-18 PY - 2013 AN - OPUS4-29755 LA - deu 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 design approval 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 is 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 safety against brittle fracture. The German guideline BAM-GGR 007 defines requirements for fracture mechanics of packagings made of DCI. Due to complex cask body structure and the dynamic loading a fracture mechanical assessment by analytical approaches is not always possible. Experience of recent design approval procedures show that the application of numerical calculations are applicable to determine the stresses and stress intensity factors 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 the flaw. The size of the artificial flaw is based on the ultrasonic inspection acceptance criteria applied for cask body manufacture. The applicant (GNS) 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 proof of the results. The paper describes the authority assessment approach for DCI fracture mechanics analysis. The validation procedure incl. the development of own tools is explained. BAM developed a postprocessor to determine the fracture mechanical loads. A horizontal 1 m puncture bar drop test is used to give a detailed description of the assessment procedure. T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 PY - 2013 SP - 1 EP - 10 PB - Omnipress AN - OPUS4-30175 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Schubert, Sven A1 - Reichardt, Adrian A1 - Wille, Frank A1 - Komann, Steffen A1 - Neumann, Martin T1 - Introduction of an ageing management approach for packages for the transport of radioactive materials N2 - With integration of the new para 613A into SSR-6 [1] the consideration of ageing mechanisms is now obligatory for the design of packages and their approval. In addition, para 809(f) of SSR-6 [1] requires consideration of the effects of storage on ageing mechanisms, safety analyses and operation and maintenance instructions. German competent authorities Bundesanstalt für Materialforschung und -prüfung (BAM) and Bundesamt für die Sicherheit der nuklearen Entsorgung (BASE) are considering the aspect of ageing in approval procedures. Ageing assessment is mainly focused on dual purpose casks (DPC) package designs which are long-term stored in interim storage facilities. For these package designs, the evaluation of ageing management is now mandatory for the maintenance of the package design approvals with a validity period of 5 years and beyond. The ageing management includes amongst others a gap analysis, the assessment of ageing effects and operational experiences during operation and interim storage. BAM works on the compilation of a guideline for implementation of paras 613A, 809(f) and 809(k) for packages requiring competent authority approval at the application procedure in Germany. The paper describes essential items of ageing mechanisms and will give a foresight to the ageing management evaluation by BAM. T2 - INMM & ESARDA Joint Annual Meeting CY - Online meeting DA - 30.08.2021 KW - Dual purpose casks KW - Ageing KW - SSR-6 KW - Interim storage KW - Ageing mechanisms KW - Transport of radaioactive materials PY - 2021 UR - https://www.abstractsonline.com/pp8/#!/10383/presentation/964 SP - 1 EP - 6 PB - Institute of Nuclear Materials Management (INMM) CY - Mount Laurel, NJ AN - OPUS4-53180 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Reichardt, Adrian A1 - Komann, Steffen A1 - Schubert, Sven A1 - Müller, Lars A1 - Wille, Frank A1 - Neumann, Martin A1 - Gröke, Carsten T1 - Experience regarding ageing management and maintenance of transport packages in Germany N2 - The presentation gives an overview about the current experiences of BAM regarding ageing management of transport packages in Germany. The concept of the BAM-GGR 023 for ageing management of transport packagings will be described. The fundamentals of BAM-GGR 023 will be shown, including the following key aspects: -the organization of an Ageing Management System (AMS), -the issue of an Ageing Management Plan (AMP), -the ageing evaluation as a part of the AMP, -the Ageing Surveillance Program (ASP) as a part of the AMP, -the Ageing Management Documentation (AMD). T2 - First Consultancy Meeting on Drafting the Safety Guide on Ageing Management and Maintenance of Radioactive Material Transport Packages CY - Vienna, Austria DA - 20.11.2023 KW - Ageing Management KW - Dual Purpose Cask KW - BAM-GGR 023 PY - 2023 AN - OPUS4-58863 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Schubert, Sven A1 - Komann, Steffen A1 - Neumann, Martin A1 - Reichardt, Adrian A1 - Müller, Lars A1 - Wille, Frank T1 - Ageing management measures of transport packages for radioactive materials N2 - The consideration of ageing mechanisms is with integration into the IAEA-requirements now obligatory for the design of transport packages. For packages intended to be used for shipment after storage the consideration of the effects of ageing mechanisms during storage in safety analyses and the implementation of corresponding instructions for operation and maintenance are required. Therefore, these packages have been maintained during storage in a manner that all requirements specified in the IAEA-requirements and in the applicable certificates of approval have been fulfilled. BAM is the competent authority in Germany for evaluating the thermal and mechanical design of packages for the transport of radioactive materials. In this context, BAM also evaluates the aging mechanisms and aging effects occurring during the long-term storage period of such transport packages. BAM has developed a guideline for the implementation of ageing assessment and of the measures for ageing management into the approval procedure based on IAEA-requirements. The goal of the presented paper is to introduce this guideline and the general approach for ageing management requirements and to describe the application to a specific package design. The type and amount of the ageing management not only depend on the use of the packaging, but also on the accessibility of the components that are considered in the ageing assessment. The different operating phases of a packaging are considered for the ageing assessment, such as loading, interim storage or transport. The implementation of measures for ageing management is divided into three levels – systemic measures, package design related measures and documentation. The systemic measures are attributed to the general management system and define the whole activities for organization of ageing management like structure, responsibilities, documentation, reports and evaluation. The package design related measures are defined in an ageing management plan. These measures shall ensure that the anticipated changes of the package design under consideration of ageing effects still complies with the design approval specification. Therefore, an ageing surveillance program and, if necessary, a gap analysis program shall be developed. The ageing management documentation ensures the continuous documentation of the compliance of a specific package to the approved package design, comprising mainly records resulting from operation and surveillance. T2 - RAMTRANS 2024 CY - London, United Kingdom DA - 15.05.2024 KW - Radioactive Materials KW - Ageing Management KW - Transport Packages KW - Dual Purpose Casks PY - 2024 VL - 2024 SP - 1 EP - 8 PB - Nuclear Institute CY - London AN - OPUS4-60186 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 - 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 - Komann, Steffen A1 - Ballheimer, Viktor A1 - Quercetti, Thomas A1 - Scheidemann, Robert A1 - Wille, Frank T1 - Mechanical and thermal assessment by BAM of a new package design for the transport of SNF from a german research reactor N2 - For disposal of the research reactor of the Technical University Munich FRM II a new transport and storage cask design was under approval assessment by the German authorities on the basis of International Atomic Energy Agency (IAEA) requirements. The cask body is made of ductile cast iron and closed by two bolted lid systems with metal seals. The material of the lids is stainless steel. On each end of the cask the wood-filled impact limiters are installed to reduce impact loads to the cask under drop test conditions. In the cavity of the cask a basket for five spent fuel elements is arranged. This design has been assessed by the Bundesanstalt für Materialforschung und -prüfung (BAM) in view to the mechanical and thermal safety analyses, the activity release approaches, and subjects of quality assurance and surveillance for manufacturing and operation of the package. For the mechanical safety analyses of the package a combination of experimental testing and analytical/numerical calculations were applied. In total, four drop tests were carried out at the BAM large drop test facility. Two tests were carried out as a full IAEA drop test sequence consisting of a 9m drop test onto an unyielding target and a 1m puncture bar drop test. The other two drop tests were performed as single 9m drop tests and completed by additional analyses for considering the effects of an IAEA drop test sequence. The main objectives of the drop tests were the investigation of the integrity of the package and its safety against release of radioactive material as well as the test of the fastening system of the impact limiters. Furthermore, the acceleration and strain signals measured during the tests were used for the verification of finite-element (FE) models applied in the safety analysis of the package design. T2 - Pressure Vessels & Piping Conference 2020 CY - Online meeting DA - 03.08.2020 KW - Mechanik KW - radioaktives Material KW - Transportbehälter KW - Antragsverfahren KW - Zulassungen KW - Typ-B Versandstück KW - Thermik PY - 2020 VL - 2020 SP - 1 EP - 7 PB - ASME CY - New York AN - OPUS4-51103 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Neumann, Martin A1 - Wille, Frank T1 - Transportbehälter für Klasse 7 - Ein Überblick N2 - Die Präsentation gibt einen Überblick über die Tranportbehälter der Klasse 7. Behandelt werden neben der Versandstückklassifikation auch die Anforderungen an diese Verpackungen gemäß der gefahrgutrechtlichen Vorschriften. T2 - 30. Münchner Gefahrguttage CY - Online meeting DA - 20.07.2020 KW - Gefahrgutrecht KW - Transportbehälter KW - Radioaktives Material KW - Zulassungspflichtige Versandstücke PY - 2020 AN - OPUS4-51104 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 - Quercetti, Thomas A1 - Scheidemann, Robert A1 - Komann, Steffen A1 - Ballheimer, Viktor A1 - Wille, Frank T1 - Package testing of a dual purpose cask for SNF from German research reactors N2 - A new dual purpose cask design was developed for the safe transport and interim storage of spent fuel elements of German research reactors. In the framework of the safety assessment within the licensing procedure the Bundesanstalt für Materialforschung und –prüfung (BAM) as competent authority performed a series of drop tests according to the IAEA-Regulations [1]. The package consists of a cylindrical thick-walled ductile cast iron cask body closed by a double lid system with metallic seals. A lid and bottom sided impact limiter consisting of a wood/steel construction limit the mechanical impact loading. The full-scale test specimen was equipped with a basket and assembled with dummy-fuel elements. The package and test specimen, respectively have a total mass of approximately 24 metric tons. The mechanical test program included three 9m free drop tests, in horizontal, vertical and oblique cask orientation onto the lid system. Additionally, a 1m-puncture drop test followed the horizontal drop test to consider an IAEA-test sequence. The horizontal and vertical drop tests were performed at a temperature of minus 40°C. During the oblique drop test the upper impact limiter was heated to +80°C. The tests were conducted onto an unyielding target, fulfilling the requirements of the IAEA regulations [1]. The test specimen was considerably instrumented with strain gauges and accelerometers. Transient strains at selected locations of the inner and outer container walls, of the primary and secondary lid, as well as of the corresponding lid bolts were measured during the drop tests. Furthermore, decelerations in different locations at the cask body and the lids were measured. The complex geometrical deformation of the impact limiters due to the impact were determined by optical 3d- measurements using the projected fringe method in combination with multi-image photogrammetry. Before and after the drop tests the leakage rate of the lid system was determined by helium leakage testing. The experimental results contribute to the evaluation of the package response to mechanical tests, demonstrating safety under normal and accident conditions of transport. Especially to the verification of the dynamic finiteelement model of the package used in the package design safety report [2]. The paper describes the performance of the drop tests, selected test results focusing on the lid screws and the cask body and the deformation of the impact limiters as well as impact kinematics, respectively. T2 - IHLRWM2019 CY - Knoxville, TN, USA DA - 14.04.2019 KW - Drop test KW - Package testing KW - Dual purpose cask PY - 2019 SN - 978-0-89448-761-3 VL - 2019 SP - paper 27283, 1 EP - 7 PB - ANS AN - OPUS4-50619 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Quercetti, Thomas A1 - Scheidemann, Robert A1 - Komann, Steffen A1 - Ballheimer, Viktor A1 - Wille, Frank T1 - Drop Testing of a New Package Design for the Transport of SNF from German Research Reactors N2 - A new dual purpose cask design was developed for the safe transport and interim storage of spent fuel elements of German research reactors. In the framework of the safety assessment within the package approval procedure the Bundesanstalt für Materialforschung und –prüfung (BAM) as competent authority performed a series of drop tests according with the IAEA Transport Regulations. The package consists of a cylindrical thick-walled ductile cast iron cask body closed by a bolted lid system with metallic seals. A lid and bottom sided impact limiter consisting of a wood/steel construction limit the mechanical impact loading. The full-scale test specimen was equipped with a basket and assembled with dummy-fuel elements. The package and test specimen, respectively have a total mass of approximately 24 metric tons. The mechanical drop test program included three 9m free drop tests, in horizontal, vertical and oblique cask orientation onto the lid system. Additionally, a 1m-puncture bar drop test followed the horizontal drop test to consider an IAEA-drop test sequence. The horizontal and vertical drop tests were performed at a temperature of minus 40°C. During the oblique drop test the upper impact limiter was heated up to +80°C. The tests were conducted onto an unyielding target, fulfilling the IAEA requirements. The test specimen was considerably instrumented with strain gauges and accelerometers. Transient strains at selected locations of the inner and outer container walls, of the primary and secondary lid, as well as of the corresponding lid bolts were measured during the drop tests. Furthermore, decelerations in different locations at the cask body and the lids were measured. The complex geometrical deformation of the impact limiters due to the impact were determined by optical 3D- measurements. Before and after the drop tests the leakage rate of the lid system was determined by helium leakage testing. The experimental results contribute to the evaluation of the package response to mechanical tests, demonstrating safety under normal and accident conditions of transport and especially to the verification of the dynamic finite-element model of the package used in the package design safety report. T2 - PATRAM 2019 CY - New Orleans, LA, USA DA - 04.08.2019 KW - SNF KW - Drop testing KW - New package design PY - 2019 SP - Paper 19-A-1142,1 EP - 10 AN - OPUS4-50622 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen T1 - Assessment of Quality Assurance Measures for Rdioactive Material Transport Packages not Requiring Competent Authority Design Approval T2 - ISSPA Annual General Meeting 2014 Workshop, Source Security in Transport and Source Control CY - Washington, USA DA - 2014-02-26 PY - 2014 AN - OPUS4-30801 LA - deu 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 - Komann, Steffen T1 - Qualitätssicherung pfüfpflichtiger Versandstücke T2 - 4. RAM-Behältersicherheitstage CY - Berlin, Germany DA - 2014-03-26 PY - 2014 AN - OPUS4-30679 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen T1 - Qualitätssicherung prüfpflichtiger Versandstücke T2 - 4. RAM-Behältersicherheitstage CY - Berlin, Germany DA - 2014-03-26 PY - 2014 AN - OPUS4-30710 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Müller, Lars A1 - Schilling, O. A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Komann, Steffen T1 - Assessment strategy of numerical analyses of RAM package components - 14619 N2 - In Germany current package design safety cases include more and more advanced numerical methods, e. g. finite element analysis (FEA), often in combination with local concepts of strength evaluation of the structure. This approach requires extensive modeling and verification procedures. As a consequence the efforts of authority assessment of design safety analysis increase as well. Only the check of pre- and post-data of numerical calculations is often not sufficient for the safety assessment. On the other hand own analyses of the mechanical problem by performing an Independent numerical modeling and analyzing is not always realizable. Therefore it is necessary to look for optimized procedures of the assessment, without loss of safety. This paper shows possibilities for the assessment strategy of numerical analyses with focus on simple analytical approaches as comparative calculations. Such approaches can be helpful to support evaluation of numerical calculations in the whole assessment procedure. Three examples are considered to Show which possibilities and limits exist to support the assessment of numerical analyses using analytical comparative calculations. Two examples of bolt and lid analysis show the influence of component and boundary stiffness on the results. Thickness to length/width ratios are partially exceeded and only fixed or free boundary conditions can be analyzed analytical. Nevertheless These analytical approaches can help to evaluate the numerical results for the assessment. The example of a trunnion demonstrates the limits of analytical approaches. The trunnion shows a complex deformation behavior and local stresses. A single basic theory isn’t matching and a construct of several approaches is not useable for calculations of local stresses. Therefore numerical calculations during assessment are necessary. Analytical approaches are not always purposeful but often effective to reduce the effort of assessment for numerical analysis of complex and safety relevant components of RAM packages. T2 - WM2014 Conference CY - Phoenix, Arizona, USA DA - 02.03.2014 PY - 2014 SN - 978-0-9836186-3-8 SP - Paper 14619, 1 EP - 14 AN - OPUS4-31047 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Schilling, O. A1 - Müller, Lars A1 - Ballheimer, Viktor A1 - Komann, Steffen A1 - Wille, Frank T1 - Analytical approaches within the assessment strategy of numerical analyses of components of RAM packages N2 - Load attachment points, consisting of the lifting component and corresponding bolt connections, and lid systems, consisting of a lid, gaskets and bolt connections, are usually analyzed numerically using the Finite Element Method. Reasons for applying the FEM are complex geometry, specific load distribution and the violation of application criteria for analytical approaches. For example the elementary beam theory is not suitable for the calculation of compact trunnions and the Kirchhoff plate theory is not suitable for the calculation of lids with a large thickness to diameter ratio because the Bernoulli-hypothesis is violated. In addition for structural integrity assessments often local stress and for the evaluation of lid systems tightness local contact opening are needed, which can be obtained only through accurately discretized numerical models. Independent comparative calculations are essential for complex calculations. Simplified but appropriate analytical approaches are an efficient way of examination. Is it possible to provide conservative analytical estimates of such kind of analysis problems? Are analytical calculations an appropriate approach in today's authority assessment business? Using the example of bolted trunnion and lid systems of a heavy package the possibilities and limits of analytical comparative calculations are shown. T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 PY - 2013 SP - Paper 217, 1 EP - 9 PB - Omnipress AN - OPUS4-31048 LA - eng 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 - Komann, Steffen A1 - Neumann, Martin A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Weber, Mike A1 - Qiao, Linan A1 - Droste, Bernhard T1 - Mechanical assessment within type B packages approval - the application of static and dynamic calculation approaches 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 - OPUS4-23384 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Günther, U. A1 - Baden, M. A1 - Komann, Steffen A1 - Nitz, Thilo T1 - Performance and restrictions of non-destructive testing (NDT) within the quality surveillance during manufacturing of type B-packages 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 KW - Qualitätsüberwachung PY - 2010 SP - 1 EP - 8 AN - OPUS4-23675 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - GEN A1 - Pasternak, H. A1 - Komann, Steffen ED - Gizejowski, ED - Kozlowski, ED - Sleczka, ED - Ziólko, T1 - A new design formula for determining the load-bearing capacity of locally supported shells T2 - XI. Miedzynarodowa Konferencja Naukowo-Techniczna "Konstrukcje Metalowe" CY - Rzeszów, Poland DA - 2006-06-21 KW - Cylindrical shell KW - Stability KW - Steel KW - Buckling KW - Longitudinal stiffeners PY - 2006 SN - 0-415-40120-8 SP - 160 EP - 161 PB - Taylor & Francis CY - London, UK AN - OPUS4-19152 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - GEN A1 - Pasternak, H. A1 - Komann, Steffen ED - Camotin, D. T1 - A new approach for determining the load-bearing capacity of locally supported shells N2 - In practice usually thin-walled cylindrical silos are supported locally at the bottom rim of the shell. In this area a stress concentration may occur, which is responisble for buckling of the shell. In the currently applied codes, there is no information, how to analyze the load-bearing capacity of theselocally supported shells. In this paper, a new design formula is given for determining the load-bearing capacity. T2 - International Colloquium on Stability and Ductility of Steel Structures CY - Lisbon, Portugal DA - 2006-09-06 KW - Cylindrical shells KW - Stability KW - Steel KW - Buckling KW - Longiudinal stiffeners PY - 2006 SN - 972-846961-6 VL - 2 SP - 595 EP - 602 PB - IST Press CY - Lisabon AN - OPUS4-19153 LA - eng 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 - CONF A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank T1 - Transport of Large Nuclear Power Plant Components - Experiences in Mechanical Design Assessment - T2 - 16.th International Symposium on the Packaging and Transport of Radioactive Materials CY - London, England DA - 2010-10-03 PY - 2010 AN - OPUS4-22216 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Qiao, Linan A1 - Zencker, Uwe A1 - Musolff, André A1 - Komann, Steffen T1 - Dynamic finite element analyses of a spent fuel transport and storage cask with impact limiters by 9 meter drop tests N2 - The 9 meter drop onto an unyielding target is one of the important mechanical tests within the safety assessment of transport casks for radioactive material. In general, the cask is equipped with impact limiters to reduce the dynamic load on the cask body by absorbing a major part of the kinetic energy. The impact limiters are often made of wood or aluminium. In this study an elastic-plastic material model with volume change was used to describe the stress-strain behaviour of wood found in crush tests. For aluminium, an elastic-incremental plastic material model with Cowper-Symonds parameters for strain rate depending material hardening was used to model the adiabatic stress-strain relations measured at specimens at constant ambient temperature. Hereafter simulations with a sophisticated finite element model were carried out and compared with different drop tests. Four drop tests of a half-scale cask model equipped with wood and aluminium impact limiters with different drop positions were selected to investigate the impact limiter behaviour during a 9 meter drop test. All drop tests were simulated with the same FE mesh but under different boundary and initial conditions. T2 - SIMULIA Customer conference CY - Barcelona, Spain DA - 17.05.2011 KW - Cask KW - Drop test KW - Impact limiter KW - Wood KW - Aluminium KW - Dynamics PY - 2011 SP - 932 EP - 944 AN - OPUS4-23706 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Wille, Frank A1 - Droste, Bernhard T1 - Gap analysis by periodical reviews of transport package design safety reports of German 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 - 18th International Symposium on the Packaging and Transportation of Radioactive Materials PATRAM2016 CY - Kobe, Japan DA - 18.09.2016 KW - Radioactiv material KW - Transportation KW - Periodoc inspection PY - 2016 AN - OPUS4-37748 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Gröke, Carsten T1 - Anforderungen an Verpackungen für prüfpflichtige Versandstücke N2 - Ziel dieses Beitrages soll es sein, zum einen das Vorgehen bei der Erlangung einer Anerkennung zur qualitätsgesicherten Herstellung einer Verpackung für prüfpflichtige Versandstücke der zuständigen Behörde im Sinne des ADR Unterabschnittes 1.7.3 darzustellen und zum anderen Probleme bei der Umsetzung der rechtsverbindlichen Vorschriften des Gefahrgutrechts aufzuzeigen. T2 - 5. RAM-Behältersicherheitstage CY - Berlin, Germany DA - 16.03.2016 KW - Qualitätssicherung KW - Radioaktive Stoffe KW - Gefahrgut KW - Transport PY - 2016 AN - OPUS4-37749 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 by periodical reviews of transport package design safety reports of German 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 - 18th International Symposium on the Packaging and Transportation of Radioactive Materials PATRAM2016 CY - Kobe, Japan DA - 18.09.2016 KW - Radioactive material KW - Gap analysis KW - Periodical review KW - Approved packages PY - 2016 SP - Paper 5004, 1 EP - 9 AN - OPUS4-37641 LA - eng 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 - Komann, Steffen T1 - BAM-GGR 016, Neue BAM-Gefahrgutregel zur Qulitätssicherung von Verpackungen für nicht zulassungspflichtige Versandstücke T2 - Informationsaustausch zur verkehrsrechtlichen Aufsicht bei der Beförderung radioaktiver Stoffe CY - Bremen, Deutschland DA - 2015-01-27 PY - 2015 AN - OPUS4-32782 LA - deu 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 Integritätsbewertung 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. Für diese Versandstücke ist eine mechanische Integritätsbewertung, nach den Vorschriften der Internationalen Atomenergie-Organisation (IAEO), notwendig, die in der Regel unter dem Einsatz einer komplexen numerischen Berechnung erfolgt. ---------------------------------------------------------- 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. For these packages a safety evaluation considering the International Atomic Energy Agency (IAEA) transport regulations is necessary, preferably by a complex numerical analysis. KW - Radioaktive Stoffe KW - Transport KW - Mechanik KW - Numerische Analyse PY - 2016 SN - 0005-6650 VL - 91 IS - 5 SP - 188 EP - 192 PB - Springer VDI Verlag CY - Düsseldorf AN - OPUS4-36016 LA - deu 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 - Long time storage KW - Aging management KW - Approval certificate PY - 2017 AN - OPUS4-39916 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 - Komann, Steffen A1 - Linnemann, Konrad A1 - Wille, Frank A1 - Reiche, Ingo A1 - Ramsay, J. T1 - New SCO-iii regulations to ship large comonents as surface contaminated objects N2 - The decommissioning or refurbishment of nuclear facilities necessitates either the storage or disposal of large radioactive components such as steam generators, pressurizers, reactor pressure vessels and heads, and coolant pumps, to list the major contributors. These components or objects are large in size and mass, measuring up to approximately 6 meters in diameter, up to 20 meters in length, and weighing over 400 000 kg. In many situations, the components are transported off-site to a storage, disposal or recycling/treatment facility. Previously, many large objects had to be transported under special arrangement. T2 - PATRAM22 CY - Juan les Pins, France DA - 11.06.2023 KW - Gefahrgut KW - Radioaktive Stoffe KW - Rückbau kerntechnischer Anlagen KW - Transport PY - 2023 SP - 1 EP - 8 AN - OPUS4-57750 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Linnemann, Konrad A1 - Wille, Frank A1 - Reiche, Ingo A1 - Ramsay, Jeff T1 - New SCO-III regulations to ship large components as surface contaminated objects N2 - The decommissioning or refurbishment of nuclear facilities necessitates either the storage or disposal of large radioactive components such as steam generators, pressurizers, reactor pressure vessels and heads, and coolant pumps, to list the major contributors. These components or objects are large in size and mass, measuring up to approximately 6 meters in diameter, up to 20 meters in length, and weighing over 400 000 kg. In many situations, the components are transported off-site to a storage, disposal or recycling/treatment facility. Previously, many large objects had to be transported under special arrangement T2 - PATRAM22 Konferenz CY - Juan-les-Pins, France DA - 11.06.2023 KW - Radioaktive Stoffe KW - Rückbau kerntechnischer Anlagen KW - Transport KW - Gefahrgut PY - 2023 AN - OPUS4-57751 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Schubert, Sven A1 - Reichardt, Adrian A1 - Müller, Lars A1 - Neumann, Martin A1 - Wille, Frank T1 - Ageing management measures for transport packages for radioactive materials N2 - The consideration of ageing mechanisms is with integration into the IAEA-requirements now obligatory for the design of transport packages. For packages intended to be used for shipment after storage the consideration of the effects of ageing mechanisms during storage in safety analyses and the implementation of corresponding instructions for operation and maintenance are required. Therefore, these packages have been maintained during storage in a manner that all requirements specified in the IAEA-requirements and in the applicable certificates of approval have been fulfilled. BAM is the competent authority in Germany for evaluating the thermal and mechanical design of packages for the transport of radioactive materials. In this context, BAM also evaluates the aging mechanisms and aging effects occurring during the long-term storage period of such transport packages. BAM has developed a guideline for the implementation of ageing assessment and of the measures for ageing management into the approval procedure based on IAEA-requirements. The goal of the presented paper is to introduce this guideline and the general approach for ageing management requirements and to describe the application to a specific package design. The type and amount of the ageing management not only depend on the use of the packaging, but also on the accessibility of the components that are considered in the ageing assessment. The different operating phases of a packaging are considered for the ageing assessment, such as loading, interim storage or transport. The implementation of measures for ageing management is divided into three levels – systemic measures, package design related measures and documentation. The systemic measures are attributed to the general management system and define the whole activities for organization of ageing management like structure, responsibilities, documentation, reports and evaluation. The package design related measures are defined in an ageing management plan (AMP). Therefore, an ageing surveillance program (ASP) and, if necessary, a gap analysis program has to be developed. The ageing management documentation (AMD) ensures the continuous documentation of the compliance of a specific package to the approved package design, comprising mainly records resulting from operation and surveillance. T2 - PVP2023-Pressure, Vessels & Piping Conference CY - Atlanta, GA, USA DA - 16.07.2023 KW - Radioaktive Stoffe KW - Transport KW - Gefahrgut PY - 2023 AN - OPUS4-58003 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Schubert, Sven A1 - Reichardt, Adrian A1 - Müller, Lars A1 - Neumann, Martin A1 - Komann, Steffen A1 - Wille, Frank T1 - Introduction of the German ageing management guide for packages for transport of radioactive materials N2 - The consideration of ageing mechanisms is with integration of the new para 613A into IAEA SSR-6 (Rev. 1) now obligatory for the design of transport packages. In addition, para 809(f) requires for packages intended to be used for shipment after storage the consideration of the effects of ageing mechanisms during storage in safety analyses and the implementation of corresponding instructions for operation and maintenance. Para 503(e) requires that all packaging components and radioactive contents have been maintained during storage in a manner that all requirements specified in IAEA SSR-6 (Rev.1) and in the applicable certificates of approval have been fulfilled. The evaluation of ageing mechanisms and their effects including monitoring are part of BAM’s authority assessment tasks related to the mechanical and thermal package design and quality assurance aspects. BAM has compiled a guideline for the implementation of ageing assessment and of the measures for ageing management of the approval procedure based on requirements of IAEA SSR-6 (Rev.1). The guideline is applicable only for packages requiring a competent authority approval. The paper aims to describe the structure of the guideline and the general approach for ageing management requirements. The type and amount of measures for ageing management depend mainly on the use of the package and on the ageing effects for the component, which result from relevant ageing mechanisms during package operation time. The implementation of measures for ageing management is divided into three levels – systemic measures, package design related measures and documentation. The systemic measures are attributed to the general management system and define the whole activities for organization of ageing management like structure, responsibilities, documentation, reports and evaluation. The package design related measures are defined in an ageing management plan (AMP). These measures shall ensure that the anticipated changes of the package design under consideration of ageing effects still complies with the design approval specification. Therefore, an ageing surveillance program (ASP) and, if necessary, a gap analysis program shall be developed. The ageing management documentation (AMD) ensures the continuous documentation of the compliance of a specific package to the approved package design, comprising mainly records resulting from operation and surveillance. T2 - PATRAM22 - 20th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Juan-les-Pins, France DA - 11.06.2023 KW - Guide KW - Ageing KW - Mechanism KW - Package KW - Management PY - 2023 SP - 1 EP - 10 PB - World Nuclear Transport Institute (WNTI) CY - London AN - OPUS4-57770 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Rolle, Annette A1 - Komann, Steffen A1 - Wille, Frank T1 - Ageing aspect in design approval of special form radioactive material N2 - In accordance with the IAEA transport regulations, the design of special form radioactive material (SFRM) shall resist a severe transport accident without undue loss or dispersal of radioactive material. The safety assessment for design approval includes besides the program for physical tests (impact, percussion, bending and heat test) also the evaluation of the management system for design, manufacture, testing, documentation, use, maintenance, and inspection. SFRM source design plus management system shall ensure, that every specimen of the approved design is able to survive the severe mechanical and thermal tests at any time of its SFRM-working life. Due to the long-term use of SFRM designs in most cases, the assessment of the source ageing behavior is an important aspect in the approval procedure. Different fields of application imply a wide range of environmental conditions, from clean room atmosphere to highly aggressive industrial conditions. Besides of radioactive content, corrosion is a main factor for possible SFRM design degradation. Although the IAEA Advisory Material SSG-26 already implies an indication of the need for considering ageing mechanisms, suitable amendments in the regulatory requirements of SSR-6 should be introduced to make the approval procedure more transparent and help to reduce rounds of questions by the authority. A supplementary requirement for considering of ageing mechanisms could be a helpful contribution to an international harmonization of the approval procedure. This paper will describe major influencing factors to be considered to assess the ageing behavior of a SFRM design and will identify the need for a regulatory specification of a SFRM-working life as basis for the assessment of the SFRM design regarding time-dependent weakening. A proposal for an explicit requirement for consideration of ageing mechanisms in safety assessment of SFRM, which should be considered in the ongoing SSR-6 revision cycle, will be explained. T2 - PATRAM22 Conference CY - Juan-Les-Pins, France DA - 11.06.2023 KW - Radioaktive Stoffe KW - Beförderung KW - Zulassung KW - Strahlenschutz PY - 2023 AN - OPUS4-57781 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Rolle, Annette A1 - Komann, Steffen A1 - Wille, Frank T1 - Ageing aspect in design approval of special form radioactive material N2 - In accordance with the IAEA transport regulations, the design of special form radioactive material (SFRM) shall resist a severe transport accident without undue loss or dispersal of radioactive material. The safety assessment for design approval includes besides the program for physical tests (impact, percussion, bending and heat test) also the evaluation of the management system for design, manufacture, testing, documentation, use, maintenance, and inspection. SFRM source design plus management system shall ensure, that every specimen of the approved design is able to survive the severe mechanical and thermal tests at any time of its SFRM-working life. Due to the long-term use of SFRM designs in most cases, the assessment of the source ageing behavior is an important aspect in the approval procedure. Different fields of application imply a wide range of environmental conditions, from clean room atmosphere to highly aggressive industrial conditions. Besides of radioactive content, corrosion is a main factor for possible SFRM design degradation. Although the IAEA Advisory Material SSG-26 already implies an indication of the need for considering ageing mechanisms, suitable amendments in the regulatory requirements of SSR-6 should be introduced to make the approval procedure more transparent and help to reduce rounds of questions by the authority. A supplementary requirement for considering of ageing mechanisms could be a helpful contribution to an international harmonization of the approval procedure. This paper will describe major influencing factors to be considered to assess the ageing behavior of a SFRM design and will identify the need for a regulatory specification of a SFRM-working life as basis for the assessment of the SFRM design regarding time-dependent weakening. A proposal for an explicit requirement for consideration of ageing mechanisms in safety assessment of SFRM, which should be considered in the ongoing SSR-6 revision cycle, will be explained. T2 - PATRAM22 Conference CY - Juan-Les-Pins, Antibes, France DA - 11.06.2023 KW - Ttransport KW - Radioactive material KW - Sealed sources KW - Ageing PY - 2023 SP - 1 EP - 6 AN - OPUS4-57786 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Komann, Steffen A1 - Wille, Frank A1 - Schubert, Sven T1 - Neue Regel für das Alterungsmanagement N2 - Die Regelwerksanforderungen für Verpackungen für den Transport radioaktiver Stoffe basieren auf den Empfehlungen der IAEA und werden über die verkehrsträgerspezifischen internationalen Regelwerke in nationales Recht umgesetzt. Um die Anforderungen des Regelwerkes zu erläutern, entwickelt die Bundesanstalt für Materialforschung und -prüfung (BAM) Gefahrgutregeln (BAM-GGR), welche die Entwickler, Hersteller und Betreiber von Behältern über die von der BAM angewandten Verfahren bei der Umsetzung gesetzlicher Bestimmungen informieren und unterstützen sollen. KW - Gefahrgut KW - Radioaktive Stoffe KW - Alterung KW - Transport PY - 2023 VL - 2023 IS - 9 SP - 16 EP - 18 PB - Heinrich Vogel CY - München AN - OPUS4-58577 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Müller, Lars A1 - Scheidemann, Robert A1 - Schönfelder, Thorsten A1 - Komann, Steffen A1 - Wille, Frank T1 - Drop tests assessment of internal shock absorbers for packages loaded with encapsulations for damaged spent nuclear fuel N2 - Damaged spent nuclear fuel (DSNF) can be loaded in German dual-purpose casks (DPC) for transport and interim storage. Encapsulations are needed to guarantee a safe handling and a tight closure, separated from the package enclosure. These encapsulations shall be durable and leak-tight for a long storage period, because they are usually not accessible within periodical inspections of the DPC. Due to the general design of DPCs for standard fuel assemblies, specific requirements have to be considered for the design of encapsulations for DSNF to ensure the loading in existing package designs. Especially the primary lid system of a DPC is designed for maximum loads due to the internal impact of the content during drop test conditions. The main difference of encapsulations for damaged spent nuclear fuel is that they have usually a much higher stiffness than standard fuel assemblies. Therefore the design of an internal shock absorber, e.g. at the head of an encapsulation is required to reduce mechanical loads to the primary lid system during impacts. BAM as part of the German competent authority system is responsible for the safety assessment of the mechanical and thermal package design, the release of radioactive material and the quality assurance of package manufacturing and operation. Concerning the mechanical design of the encapsulation BAM was involved in the comprehensive assessment procedure during the package design approval process. An internal shock absorber was developed by the package designer with numerical analyses and experimental drop tests. Experimental drop tests are needed to cover limiting parameters regarding, e.g. temperature and wall thickness of the shock absorbing element to enable a detailed specification of the whole load-deformation behavior of the encapsulation shock absorber. The paper gives an overview of the assessment work by BAM and points out the main findings which are relevant for an acceptable design of internal shock absorbers. The physical drop tests were planned on the basis of pre-investigations of the applicant concerning shape, dimension and material properties. In advance of the final drop tests the possible internal impact behavior had to be analyzed and the setup of the test facility had to be validated. The planning, performance and evaluation of the final drop tests were witnessed and assessed by BAM. In conclusion it could be approved that the German encapsulation system for damaged spent nuclear fuel with shock absorbing components can be handled similar to standard fuel assemblies in existing package designs. T2 - Pressure Vessels & Piping Conference 2020 CY - Online meeting DA - 03.08.2020 KW - Encapsulations for damaged spent nuclear fuel KW - Drop tests KW - Internal shock absorber KW - Design assessment of RAM packages PY - 2020 SP - 1 EP - 9 AN - OPUS4-51546 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Komann, Steffen T1 - Aktuelle Entwicklungen und Prüfungen der BAM N2 - Es werden im Rahmen des Vortrages die aktuellen Prüftätigkeiten und Arbeitsschwerpunkte der BAM vorgestellt. Des Weiteren wird darüber berichtet was es Neues in der aktuellen Normen- und Richtlinienentwicklung gibt. T2 - Informationsveranstaltung zum Transport radioaktiver Stoffe CY - Online meeting DA - 14.04.2021 KW - Gefahrgutregeln der BAM KW - Fallprüfungen KW - Radioaktive Materialien KW - Zulassung PY - 2021 AN - OPUS4-52493 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias A1 - Komann, Steffen A1 - Neumann, Martin A1 - Linnemann, Konrad A1 - Wille, Frank T1 - Fracture Mechanical Analyses of a Welding Seam of a Thick-Walled Transport Package N2 - Transport packages shall satisfy various safety criteria regarding mechanical, thermal and radiation phenomena. Typical requirements focusing mechanical aspects are usually drop tests in accordance with IAEA regulations [1]. The drop tests are usually carried out experimentally and, as an additional measure, finite element analyses (FEA) are performed. A specific part of the investigations presented is the evaluation of the welding seam connecting cask shell and cask bottom. Experimental results and FEA are presented and compared. The evaluation of the welding seam performed includes a variety of aspects. In addition to the experimental and analytical stresses determined, different standards are used to investigate a possible crack initiation. Several destructive and non-destructive tests are performed for quality assurance in the manufacturing process as well as for different input parameters. The necessary monitoring and non-destructive measurement methods to define the boundary conditions of the standards are introduced. Taking into account all required parameters, the welding seam is examined and evaluated using the failure assessment diagrams (FAD) of the respective standards. It can be shown under the given boundary conditions that considering the experimental data, the welding seam is in the context of crack initiation below the enveloping curve in the acceptable region. More critical drop tests to be conducted are proposed and need to be investigated in future work. T2 - Pressure Vessels & Piping Conference® 2022 CY - Las Vegas, NV, USA DA - 17.07.2022 KW - Drop test KW - Fracture initiation KW - Transport package PY - 2022 AN - OPUS4-55374 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias A1 - Neumann, Martin A1 - Linnemann, Konrad A1 - Komann, Steffen A1 - Wille, Frank T1 - Fracture Mechanical Analyses of a Welding Seam of a Thick-Walled Transport Package N2 - Transport packages shall satisfy various safety criteria regarding mechanical, thermal and radiation phenomena. Typical requirements focusing mechanical aspects are usually drop tests in accordance with IAEA regulations. The drop tests are usually carried out experimentally and, as an additional measure, finite element analyses (FEA) are performed. A specific part of the investigations presented is the evaluation of the welding seam connecting cask shell and cask bottom. Experimental results and FEA are presented and compared. The evaluation of the welding seam performed includes a variety of aspects. In addition to the experimental and analytical stresses determined, different standards are used to investigate a possible crack initiation. Several destructive and non-destructive tests are performed for quality assurance in the manufacturing process as well as for different input parameters. The necessary monitoring and non-destructive measurement methods to define the boundary conditions of the standards are introduced. Taking into account all required parameters, the welding seam is examined and evaluated using the failure assessment diagrams (FAD) of the respective standards. It can be shown under the given boundary conditions that considering the experimental data, the welding seam is in the context of crack initiation below the enveloping curve in the acceptable region. More critical drop tests to be conducted are proposed and need to be investigated in future work. T2 - Pressure Vessels & Piping Conference® 2022 CY - Las Vegas, NV, USA DA - 17.07.2022 KW - Transport package KW - Drop test KW - Fracture initiation PY - 2022 SP - 1 EP - 9 AN - OPUS4-55375 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -