Dokument-ID Dokumenttyp Autoren/innen Persönliche Herausgeber/innen Haupttitel Abstract Auflage Verlagsort Verlag Herausgeber (Institution) Erscheinungsjahr Titel des übergeordneten Werkes Jahrgang/Band ISBN Veranstaltung Veranstaltungsort Beginndatum der Veranstaltung Enddatum der Veranstaltung Ausgabe/Heft Erste Seite Letzte Seite URN DOI Lizenz Datum der Freischaltung OPUS4-56343 Beitrag zu einem Tagungsband Gröke, Carsten; Neumann, Martin; Komann, Steffen; Wille, Frank Anforderungen an die Bauartprüfung und Maßnahmen zur Qualitätssicherung für nicht zulassungspflichtige Versandstücke 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. 2019 Kontec 2019 Tagungsband - Proceedings KONTEC 2019 Dresden, Germany 27.03.2019 136 142 2022-11-23 OPUS4-53180 Beitrag zu einem Tagungsband Schubert, Sven; Reichardt, Adrian; Wille, Frank; Komann, Steffen; Neumann, Martin Introduction of an ageing management approach for packages for the transport of radioactive materials 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. Mount Laurel, NJ Institute of Nuclear Materials Management (INMM) European Safeguards Research and Development Association (ESARDA) 2021 Proceedings of the INMM & ESARDA Joint Annual Meeting INMM & ESARDA Joint Annual Meeting Online meeting 30.08.2021 01.09.2021 1 6 2021-09-03 OPUS4-55374 Vortrag Gleim, Tobias; Komann, Steffen Fracture Mechanical Analyses of a Welding Seam of a Thick-Walled Transport Package 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. 2022 Pressure Vessels & Piping Conference® 2022 Las Vegas, NV, USA 17.07.2022 22.07.2022 2022-07-26 OPUS4-55375 Beitrag zu einem Tagungsband Gleim, Tobias; Neumann, Martin; Linnemann, Konrad; Komann, Steffen; Wille, Frank Fracture Mechanical Analyses of a Welding Seam of a Thick-Walled Transport Package 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. 2022 Proceedings of the ASME 2022 Pressure Vessels & Piping Conference Pressure Vessels & Piping Conference® 2022 Las Vegas, NV, USA 17.07.2022 22.07.2022 1 9 2022-07-27 OPUS4-55423 Beitrag zu einem Tagungsband Linnemann, Konrad; Komann, Steffen; Wille, Frank; Reiche, I.; Ramsay, J. New Sco-III regulations to ship large objects as surface contaminated objects 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. 2022 Proceedings der Kerntechnik 2022 Kerntechnik 2022 Leipzig, Germany 21.06.2022 22.06.2022 1 7 2022-08-03 OPUS4-44168 Zeitschriftenartikel Komann, Steffen; Wille, Frank Zur Prüfung verpflichtet - Die BAM-Gefahrgutregel 016 (BAM-GGR 016) 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. München Springer Fachmedien München GmbH 2018 Gefahrgut 2018 1-2 10 12 2018-02-15 OPUS4-44170 Beitrag zu einem Tagungsband Komann, Steffen; Gröke, Carsten; Wille, Frank Assessment of quality management for transport packages not requiring authority design approval 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. WM Symposia Inc. 2018 Waste Management Conference 2018 Waste Management Conference 2018 Phoenix, USA 18.3.2018 22.3.2018 Article 18194, 1 7 2018-02-15 OPUS4-45258 Beitrag zu einem Tagungsband Müller, Lars; Schönfelder, Thorsten; Komann, Steffen; Ballheimer, Viktor; Wille, Frank Müller, Lars Assessment experience on packages loaded with damaged spent nuclear fuel for transport after storage 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. 2018 11th International Conference on the Transport, Storage and Disposal of Radioactive Materials 11th International Conference on the Transport, Storage and Disposal of Radioactive Materials London, UK 15.05.2018 17.05.2018 Paper 18524, 1 8 2018-06-22 OPUS4-49054 Beitrag zu einem Tagungsband Komann, Steffen; Ballheimer, Viktor; Quercetti, Thomas; Scheidemann, Robert; Wille, Frank Design assessment by bam of a new package design for the transport of snf from a german research reactor For disposal of the German research reactor of the Technical University Munich FRM II a new transport and storage cask design was under approval assessment by the German authorities. The Bundesanstalt für Materialforschung und -prüfung (BAM) assessed the mechanical and thermal package safety and performed drop tests. The activity release approaches and subjects of quality assurance and surveillance for manufacturing and operation of the package were assessed by BAM as well. 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 a wood-filled impact limiter is installed to reduce impact loads to the cask under drop test conditions. In the cavity of the cask a basket for five spent fuel assemblies is arranged. For the safety case 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 test 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-Analysis (FEA) used for the safety analysis of the package design. The finite-element models incorporated in the package design safety report include the cask body, the lid system, the inventory and the impact limiters with the fastening system. In this context special attention was paid to the modeling of the encapsulated wood-filled impact limiters. Additional calculations using the verified numerical models were done by the applicant and assessed by BAM to investigate e.g. the brittle fracture of the cask body made of ductile cask iron within the package design approval procedure. This paper describes the package design assessment from the view of the competent authority BAM including the applied assessment strategy, the conducted drop tests and the additional calculations by using numerical and analytical methods. 2019 Proceedings of the 19th International Symposium on the Packaging and Transportation of Radioactive Materials PATRAM 2019 19th International Symposium on the Packaging and Transportation of Radioactive Materials PATRAM 2019 New Orleans, LA, USA 04.08.2019 09.08.2019 Paper 1176, 1 2019-09-23 OPUS4-49044 Vortrag Komann, Steffen Design assessment by BAM of a new package design for the transport of SNF from a German research reactor For disposal of the German research reactor of the Technical University Munich FRM II a new transport and storage cask design was under approval assessment by the German authorities. The Bundesanstalt für Materialforschung und -prüfung (BAM) assessed the mechanical and thermal package safety and performed drop tests. The activity release approaches and subjects of quality assurance and surveillance for manufacturing and operation of the package were assessed by BAM as well. 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 a wood-filled impact limiter is installed to reduce impact loads to the cask under drop test conditions. In the cavity of the cask a basket for five spent fuel assemblies is arranged. For the safety case 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 test 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-Analysis (FEA) used for the safety analysis of the package design. The finite-element models incorporated in the package design safety report include the cask body, the lid system, the inventory and the impact limiters with the fastening system. In this context special attention was paid to the modeling of the encapsulated wood-filled impact limiters. Additional calculations using the verified numerical models were done by the applicant and assessed by BAM to investigate e.g. the brittle fracture of the cask body made of ductile cask iron within the package design approval procedure. This paper describes the package design assessment from the view of the competent authority BAM including the applied assessment strategy, the conducted drop tests and the additional calculations by using numerical and analytical methods. 2019 PATRAM 2019 New Orleans, USA 04.08.2019 09.08.2019 2019-09-23