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 - 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 - Gröke, Carsten A1 - Neumann, Martin A1 - Komann, Steffen A1 - Wille, Frank T1 - Anforderungen an die Bauartprüfung und Maßnahmen zur Qualitätssicherung für nicht zulassungspflichtige Versandstücke N2 - Überblick über die Regelwerksanforderungen an die Bauarten von nicht zulassungspflichtigen Versandstücken, sowie die zu erfüllenden Maßnahmen zur Qualitätssicherung für die Auslegung, die Herstellung und den Betrieb. T2 - KONTEC 2019 CY - Dresden, Germany DA - 27.03.2019 KW - Radioaktiv KW - Beförderung KW - IP-2 KW - IP-3 KW - Typ A KW - Anerkennung KW - Bauartprüfung PY - 2019 SP - 136 EP - 142 AN - OPUS4-56343 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 - Rolle, Annette A1 - Komann, Steffen A1 - Wille, Frank ED - Baraldi, P. ED - Di Maio, F. ED - Zio, E. T1 - Ageing aspect in the safety evaluation of special form radioactive material N2 - In accordance with the IAEA transport regulations Special Form Radioactive Material (SFRM) is either an indispersible solid radioactive material or a sealed capsule containing radioactive material. The design of special form radioactive material has to resist a severe transport accident without undue loss or dispersal of radioactive material. The safety assessment for design approval shall include besides the required test program (impact, percussion, bending and heat test) also the evaluation of the management system for design, manufacture, testing, documentation, use, maintenance and inspection. The specified quality assurance measures have to assure, that every specimen of the approved design is produced in the same verified quality and every specimen must be able to survive the severe mechanical and thermal tests at any time of its working life. Due to the long-term use of SFRM the consideration of ageing is an important aspect in the approval procedure by BAM, the competent authority for SFRM approval in Germany. Different fields of application imply a wide range of environmental conditions, from clean room atmosphere to highly aggressive industrial conditions. So, besides of radioactive content, corrosion is a main factor for possible design degradation. This paper will describe major influencing factors to be taken into account to assess the ageing behavior of a SFRM design and will emphasize that there is a need for a regulatory specification of a SFRM-working life as basis for the aging evaluations. T2 - 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL2020 PSAM15) CY - Online meeting DA - 01.11.2020 KW - Special form radioactive material KW - Transport KW - Ageing KW - Material KW - Safety assessment KW - Management system PY - 2020 UR - https://www.rpsonline.com.sg/proceedings/esrel2020/html/3659.xml SN - 987-981-14-8593-0 SP - Paper 3659,1 EP - 5 PB - Research Publishing CY - Singapore AN - OPUS4-50969 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -