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 - 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 - Linnemann, Konrad A1 - Komann, Steffen A1 - Wille, Frank A1 - Reiche, I. A1 - Ramsay, J. T1 - New Sco-III regulations to ship large objects 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. The latest 2018 edition of the International Atomic Energy Agency (IAEA) Regulations for the Safe Transport of Radioactive Material, No. SSR-6 [1], incorporates regulations for the shipment of large objects as a new category of surface contaminated object, SCO-III, based on the IAEA “performance package” concept. This paper provides background and practical guidance on these regulations. Additionally, the experiences of BAM with the appoval of two steam converters of the NPP Lingen are presented as the first approval process for SCO-III objects in Germany. The primary additions to SSR-6 include SCO-III classification and requirements, approval and administrative requirements for the new classification, and the addition of SCO-III to the proper shipping name for UN 2913. Advisory material drafted for the new requirements will be included in the next revision of SSG-26, Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material, expected to be published soon. Note that at this time the proposed provisions for large objects do not include components such as reactor vessels, due to the more limited experience and greater radioactivity levels. The SCO-III concept lays the groundwork and may be extended to cover other large objects that are classified as low specific activity (LSA) material in the future. T2 - Kerntechnik 2022 CY - Leipzig, Germany DA - 21.06.2022 KW - Transport packages KW - Radioactive materials KW - SCO-III KW - Large objects PY - 2022 SP - 1 EP - 7 AN - OPUS4-55423 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 - 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 - CONF A1 - Komann, Steffen 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 - 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 - Neumeyer, Tino A1 - Komann, Steffen A1 - Wille, Frank T1 - Tightness Verification Procedures of Welded Lids for Encapsulations for Damaged Spent Nuclear Fuel in the Design Approval Process of Dual-Purpose Cask (DPC) N2 - Damaged spent nuclear fuel (DSNF) can be loaded in Germany in dual-purpose casks (DPC) for transport and interim storage. Therefore, special encapsulations are needed to guarantee a safe handling and a reliable tight closure. These encapsulations must be durable and leak-tight for a possible long interim storage term and the transport thereafter. Because components inside the package containment like these special encapsulations are not accessible within periodical inspections of the DPC. Due to the general design of DPCs for standard fuel assemblies, specific requirements needed to be considered for the design of the special encapsulations for DSNF to ensure the loading in existing package designs. Especially the content of the encapsulations with damaged spent nuclear fuel must be avoided to enter in the inner space of the DPC, due to the possible higher amount of corrosive and radioactive material as well as greater quantities of water, which can be trapped in defect fuel rods. The main difference of the encapsulations for damaged spent nuclear fuel is that they have no replaceable sealing as tightness barrier. They have a specific welded lid. Therefor the tightness needs to be shown in the design approval process for all transport conditions especially for accident conditions, where high internal impact loads can occur. As German competent authority, BAM is responsible for the safety assessment of the mechanical and thermal design, the release of radioactive material and the quality assurance aspects of manufacturing and operation. Concerning the mechanical package design BAM carried out a comprehensive safety assessment for the specific encapsulations during the package design approval process of the commonly used DPC. For the proof of tightness of the welded lid no representative standards are available. For minor loads the German KTA standards can verify the tightness of a weld seam but only for low stresses. For higher impact loads physical tests are required. These tests were performed with representative drop tests and highly sensitive leak tests. The present paper gives an overview of the applied assessments by BAM and points out the main findings of the mechanical design for the welded sealing. Both verification procedures are described, on one hand with German standards for lower loads and on the other hand with physical tests for higher impact loads. The drop tests based on complex numerical pre-calculations by the applicant which result in optimized dimensions and properties of the test specimens and representative drop heights for the drop tests. The horizontal drop orientation with a plane-parallel impact on the unyielding basement was most challenging and led to specific adaptations during preparation and conduction of the drop tests performed by BAM. With leak tests and the corresponding test evaluation, it could be finally approved that the specific welded sealing of an encapsulations for DSNF is tight for all transport conditions. T2 - ASME PVP 2024 CY - Bellevue, WA, USA DA - 28.07.2024 KW - Tightness verification KW - Weld seams KW - Drop tests KW - Encapsulations for damaged spent nuclear fuel KW - Helium leak tests PY - 2024 SN - 978-0-7918-8851-3 SP - 1 EP - 10 CY - New York AN - OPUS4-61351 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -