TY - GEN A1 - Neumann, Martin T1 - Measures for quality and ageing management for non-competent authority approved package designs for the transport of radioactive material T2 - BAM dangerous goods procedural rules N2 - The BAM-GGR 016 explains the requirements of ADR, RID, ADN and IMDG for management systems for design, manufacturing, testing, documentation, use, maintenance, and inspection of packagings for package designs for the transport of radioactive material not requiring competent authority approval for all designs of exempted packages, packages type IP-1, IP-2, IP-3 and type A. KW - Transport KW - Not requiring competent authority approval KW - Radioactive material PY - 2024 UR - https://nbn-resolving.org/urn:nbn:de:kobv:b43-597214 VL - BAM-GGR 016 SP - 1 EP - 25 PB - Bundesanstalt für Materialforschung und -prüfung (BAM) CY - Berlin ET - Revision 1 AN - OPUS4-59721 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Neumann, Martin A1 - Feldkamp, Martin A1 - Linnemann, Konrad A1 - Wille, Frank T1 - Transport of HLW canisters on sea vessels T2 - Proceedings of RAMTrans 2024 N2 - Germany had been transporting spent fuel to the reprocessing plant in Sellafield and La Hague for decades until around 2005. Resulting from the obligation to take back the vitrified high-level waste from reprocessing six CASTOR® HAW28M filled with 168 canisters with radioactive waste in vitrified form were transported from Sellafield to the interim storage facility Biblis in 2020. Rail wagons were used for the transport to the port in Barrow-in-Furness, where they were loaded into a dedicated seagoing vessel, certified as INF Class 3 according to the INF Code. This was the first time that vitrified high level waste with considerable heat load was transported under a German design approval certificate. BAM was involved in the authority assessment of the conditions for the sea transport. For the first transport BAM required among others, assessment of temperature distribution during transport, logging of temperatures of cargo bays and graphical imaging of temperatures of the bay with the cask in order to ensure compliance with temperature specifications, e.g. maximal neutron absorber and gasket temperatures. Special interest was taken in the identification of possible events exceeding the specified temperatures considering the different philosophies of IMDG code and its supplement INF code regarding temperature control of hatches. Results show compliance with assumed conditions. T2 - RAMTrans 2024 CY - London, United Kingdom DA - 14.05.2024 KW - Transport KW - Spent nuclear fuel KW - Ship KW - Sea KW - High level waste PY - 2024 SP - 1 EP - 8 AN - OPUS4-60092 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Neumann, Martin A1 - Feldkamp, Martin A1 - Linnemann, Konrad A1 - Wille, Frank T1 - Transport of HLW canisters on sea vessels N2 - Germany had been transporting spent fuel to the reprocessing plant in Sellafield and La Hague for decades until around 2005. Resulting from the obligation to take back the vitrified high-level waste from reprocessing six CASTOR® HAW28M filled with 168 canisters with radioactive waste in vitrified form were transported from Sellafield to the interim storage facility Biblis in 2020. Rail wagons were used for the transport to the port in Barrow-in-Furness, where they were loaded into a dedicated seagoing vessel, certified as INF Class 3 according to the INF Code. This was the first time that vitrified high level waste with considerable heat load was transported under a German design approval certificate. BAM was involved in the authority assessment of the conditions for the sea transport. For the first transport BAM required among others, assessment of temperature distribution during transport, logging of temperatures of cargo bays and graphical imaging of temperatures of the bay with the cask in order to ensure compliance with temperature specifications, e.g. maximal neutron absorber and gasket temperatures. Special interest was taken in the identification of possible events exceeding the specified temperatures considering the different philosophies of IMDG code and its supplement INF code regarding temperature control of hatches. Results show compliance with assumed conditions. T2 - RAMTrans 2024 CY - London, GB DA - 14.05.2024 KW - Transport KW - Spent nuclear fuel KW - Ship KW - Sea KW - High level waste PY - 2024 AN - OPUS4-60091 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 T2 - Proceedings of Conference RAMTRANS 2024 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 - CONF A1 - Quercetti, Thomas A1 - Wille, Frank A1 - Neumann, Martin A1 - Linnemann, Konrad T1 - Full-scale drop testing with a heavy-weight package for radioactive waste T2 - Proceedings of the 20th International Symposium on the Packaging and Transportation of Radioactive Materials N2 - Packages for the transport of radioactive materials shall fulfil the requirements of the IAEA regulations for the safe transport. The requirements define mechanical and thermal test conditions including criteria ensuring the package design’s ability to withstand severe accidents and provide a high level of technical safety. Different methods can be used for safety demonstration showing compliance with the regulations. The central part of a safety demonstration which is presented in this paper was a comprehensive drop test program with a full-scale model of a transport package accompanied by pre- and post-test FE analyses. Using full-scale drop test models allow the benefit that similarity and scaling issues become a significant smaller issue, additional material investigations can be limited and analyses for transferring test results to the original package design are reduced. Additionally, experience for the future serial packaging manufacturing and handling procedures can be collected in a very early state of the design approval process. The pre-test finite element analyses derived and justified the drop test program consisting of several drop sequences with different drop orientations of the specimen. The performance and the results of the drop test sequences shows the manageability and the advantage e.g., in view of the direct availability of test results for the package licensing. On the other hand, the drop test performance shows the difficulties during handling and the need for additional equipment during preparation of the specimen. The package presented was intended for the transport and storage of compacted radioactive waste from reprocessing of spent nuclear fuel assemblies - designed and applied for approval by the AGC consortium. The project ended in 2021. The package design was characterized by a cask body made of a forged thick stainless-steel shell, a bolted double lid system with metallic gaskets and wood filled shock absorbers at both ends. The total mass of the entire transport package including content was 120,000 kg, the total length was about 7000 mm and the diameter approximately 3000 mm, both measures include the shock absorbers. The paper provides an insight into the performance of a full-scale drop testing campaign within the package safety evaluation and shows some selected test results. T2 - 20th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 22) CY - Juan-Les-Pins, France DA - 11.06.2023 KW - Full-scale KW - Drop testing KW - Package KW - Radioactive materials transport PY - 2023 SP - 1 EP - 10 AN - OPUS4-57732 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 T2 - Proceedings of the 20th international symposium on the packaging and transportation 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 - Gleim, Tobias A1 - Neumann, Martin A1 - Linnemann, Konrad A1 - Wille, Frank T1 - Welding Seam Safety Evaluation in a Thick-Walled Type B Transport Package T2 - Proceedings of the 20th International Symposium on the Packaging and Transportation of Radioactive Materials N2 - The safety demonstration of heavy weight type B transport packages used for storing spent nuclear fuel (SNF) or radioactive waste is ensured by a combination of physical testing and numerical calculations. While experiments are performed in accordance with the IAEA regulations for selected drop scenarios, Finite-Element-Method (FEM) simulations are used to predict the most damaging case and to investigate additional drop positions. BAM as competent authority in Germany has performed different investigations of a welding seam for a typical large transport package made of A508 forged steel, where the bottom plate is welded to the cylindrical shell. The package has a mass of approx. 120 t. Results of physical drop tests with a full-scale model and accompanying preliminary FEM simulations are presented to determine the decisive stresses in the welding seam. A drop test only represents one set of a package and test parameters. A further parameter analysis is considered to account for allowable variations of packaging properties (e.g. resulting from the manufacturing process) and, based on IAEA requirements, the temperature dependence of the material behaviour. The results of the stress analyses from the drop test and the simulation form the basis and provide the input parameters for a fracture mechanics analysis. In addition to the IAEA specifications, further standards are taken into account for an in-depth investigation, see R6 [1], BS 7910 [2] and API 579-1/ASME FFS1 [3]. All the above-mentioned standards require a manufacturer-specific defect analysis with respect to size and position. Both result from the welding process and the following heat treatment regime. The maximum defect sizes are ensured with non-destructive test methods (such as ultrasonic or particle methods) as integral part of the manufacturing process of the welding seam. Another important parameter in the welding process is the residual stress (secondary stress). The combination of the primary and secondary stress determines the total stress in the welding seam. The most damaging case of the welding seam is determined and evaluated with help of the abovementioned standards and taking into account the IAEA requirements with respect to defect sizes, material properties, primary and residual stress, yield strength etc. T2 - PATRAM 22 - The International Symposium on the Packaging and Transportation of Radioactive Materials CY - Juan-Les-Pins, Antibes DA - 11.06.2023 KW - Welding KW - Transport Package KW - Fracture Mechanics PY - 2023 SP - 1 EP - 11 AN - OPUS4-59421 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias A1 - Neumann, Martin A1 - Linnemann, Konrad A1 - Wille, Frank T1 - Fracture Mechanical Analyses of a Welding Seam of a Thick-Walled Transport Package T2 - Proceedings of the 20th International Symposium on the Packaging and Transportation of Radioactive Materials N2 - The safety demonstration of heavy weight type B transport packages used for storing spent nuclear fuel (SNF) or radioactive waste is ensured by a combination of physical testing and numerical calculations. While experiments are performed in accordance with the IAEA regulations for selected drop scenarios, Finite-Element-Method (FEM) simulations are used to predict the most damaging case and to investigate additional drop positions. BAM as competent authority in Germany has performed different investigations of a welding seam for a typical large transport package made of A508 forged steel, where the bottom plate is welded to the cylindrical shell. The package has a mass of approx. 120 t. Results of physical drop tests with a full-scale model and accompanying preliminary FEM simulations are presented to determine the decisive stresses in the welding seam. A drop test only represents one set of a package and test parameters. A further parameter analysis is considered to account for allowable variations of packaging properties (e.g. resulting from the manufacturing process) and, based on IAEA requirements, the temperature dependence of the material behaviour. The results of the stress analyses from the drop test and the simulation form the basis and provide the input parameters for a fracture mechanics analysis. In addition to the IAEA specifications, further standards are taken into account for an in-depth investigation, see R6, BS 7910 and API 579-1/ASME FFS1. All the above-mentioned standards require a manufacturer-specific defect analysis with respect to size and position. Both result from the welding process and the following heat treatment regime. The maximum defect sizes are ensured with non-destructive test methods (such as ultrasonic or particle methods) as integral part of the manufacturing process of the welding seam. Another important parameter in the welding process is the residual stress (secondary stress). The combination of the primary and secondary stress determines the total stress in the welding seam. The most damaging case of the welding seam is determined and evaluated with help of the above-mentioned standards and taking into account the IAEA requirements with respect to defect sizes, material properties, primary and residual stress, yield strength etc. T2 - PATRAM 22 - The International Symposium on the Packaging and Transportation of Radioactive Materials CY - Juan-Les-Pins, Antibes, France DA - 11.06.2023 KW - Transport Package KW - Welding KW - Fracture Mechanics PY - 2023 SP - 1 EP - 10 AN - OPUS4-57696 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Neumann, Martin A1 - Gleim, Tobias A1 - Gradt, Thomas A1 - Wille, Frank T1 - Friction coefficients for wood-wood and wood-steel interfaces in impact limiters for transport casks N2 - Wood is widely used in impact limiters of transport casks for radioactive material. Encapsulated by an outer and inner steel structure, spruce wood is often applied in layers of alternating direction. The friction at the interfaces between these layers is of crucial importance for the impact and energy absorption e.g., at an accidental impact of a cask against a hard target. In order to get detailed information for corresponding numerical calculations, in this study the friction coefficient for the combinations wood-wood and wood-steel was measured in the temperature range between -40°C and 90°C according to the relevant stress conditions for such casks. Results show decreasing friction with increasing temperature, ranging from 0.43 at -40°C to 0.22 for 90°C for wood-steel combinations and from 0.3 at -40°C to 0.24 at 90°C to for a wood-wood combination. T2 - 20th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 22) CY - Juan-les-Pins, France DA - 11.06.2023 KW - Wood KW - Friction KW - Transport cask PY - 2023 AN - OPUS4-57702 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Neumann, Martin A1 - Quercetti, Thomas A1 - Wille, Frank A1 - Gröke, Carsten A1 - Neumeyer, Tino T1 - Design assessment, approval of management systems and ageing aspects of transport packages for radioactive material not requiring competent authority approval of design T2 - Proceedings of the 20th International Symposium on the Packaging and Transportation of Radioactive Materials N2 - Most transports of radioactive materials are carried out with packages not requiring competent authority approval of design. These encompass – in accordance with the IAEA SSR-6 regulations – packages of the classification excepted, Industrial packages Type 1, 2 and 3 and Type A packages. Currently an upsurge in number and variation of these package designs can be seen in Germany, resulting from the phase out of nuclear energy in Germany as well as e. g. increased use of radioactive material for medical purposes. A design assessment regarding the package safety is required in the international IAEA SSR-6 regulations. BAM operates facilities for the performance of all regulatory tests required such as drop towers for a wide range of package masses and dimensions, fire test, leak tightness measurements and pressure test facilities. Experiences with several package types are shown. Additionally, IAEA SSR-6 requires the establishment of a management system for design, manufacture, maintenance, and repair of the packaging as well as for the preparation, consigning, loading, carriage, unloading and receipt of the package. Relevant for Germany, BAM has published guidance material on the process of management system acceptance in the technical guide BAM-GGR 016. The requirements encompass quality management plans for the manufacturing of packages including independent manufacturing surveillance and specific instructions for operation, maintenance, and repair of packagings. Examples for management system specifics and requirements are given. Since the latest edition of the IAEA SSR-6 regulations an ageing evaluation including systematic ageing management measures are required for all kind of package types. BAM is going to update the guidance material BAM-GGR 016 to support the stakeholders with relevant information to fulfil the ageing aspect for packages not requiring competent authority approval. The paper explains how the ageing aspect may be included in the safety evaluation process and the management system measures and will give an outlook for the future guidance material. T2 - 20th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 22) CY - Juan-Les-Pins, Antibes, France DA - 11.06.2023 KW - Typ KW - Radioactive material KW - Non-approved PY - 2023 SP - 1 EP - 8 AN - OPUS4-57703 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -