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 OPUS4-33428 Zeitschriftenartikel Droste, Bernhard; Komann, Steffen; Wille, Frank; Rolle, Annette; Probst, Ulrich; Schubert, Sven Consideration of aging mechanism influence on transport safety of dual purpose casks for spent nuclear fuel of HLW 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'. Ashford Ramtrans Publ. 2014 Packaging, transport, storage & security of radioactive materials (RAMTRANS) 25 3-4 105 112 10.1179/1746510914Y.0000000070 2016-02-20 OPUS4-32518 Beitrag zu einem Tagungsband Droste, Bernhard; Komann, Steffen; Wille, Frank; Rolle, Annette; Probst, Ulrich; Schubert, Sven Considerations of aging mechanisms influence on transport safety and reliability of dual purpose casks for spent nuclear fuel or HLW When storage of spent nuclear fuel (SNF) or high-level waste (HLW) is done 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 (like 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. The consideration of these issues in the DPC transport safety case will be addressed. Special attention is given to all cask components which cannot be directly inspected or changed without opening the cask cavity, what are 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 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 the license holders and in appropriate design approval update processes. The paper addresses issues which are subject of an actual IAEA TECDOC draft 'Preparation of a safety case for a dual purpose cask containing spent nuclear fuel'. 2014 PSAM 12 - Probabilistic safety assessment and management (Proceedings) PSAM 12 - Probabilistic safety assessment and management Honolulu, Hawaii, USA 22.06.2014 27.06.2014 1 10 2016-02-20 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 OPUS4-56343 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 OPUS4-49054 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 OPUS4-51103 Beitrag zu einem Tagungsband Komann, Steffen; Ballheimer, Viktor; Quercetti, Thomas; Scheidemann, Robert; Wille, Frank Mechanical and thermal assessment by BAM of a new package design for the transport of SNF from a german research reactor 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. New York ASME 2020 Pressure Vessels Piping Conference 2020 Pressure Vessels & Piping Conference 2020 Online meeting 03.08.2020 07.08.2020 1 7 2020-08-12 OPUS4-20820 Zeitschriftenartikel Komann, Steffen; Droste, Bernhard; Wille, Frank Transport of large nuclear power plant components: experiences in mechanical design assessment 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. Since it concerns surface contaminated objects or low specific activity materials, a safety evaluation considering the IAEA transport regulations mainly for industrial packages (type IP-2) is necessary. For these types of industrial packages the requirements from normal transport conditions are to be covered for the mechanical proof. For example, a free drop of the package from a defined height, in dependence of its mass, onto an unyielding target, and a stacking test are required. Since physical drop tests are impossible generally due to the singularity of such 'packages', a calculation has to be performed, preferably by a complex numerical analysis. The assessment of the loads takes place on the basis of local stress distributions, also with consideration of radiation induced brittleness of the material and with consideration of recent scientific investigation results. Large nuclear components have typically been transported in an unpackaged manner, so that the external shell of the component provides the packaging wall. The investigation must consider the entire component including all penetration areas such as manholes or nozzles. According to the present IAEA regulations the drop position is to be examined, which causes the maximum damage to the package. In the case of a transport under special arrangement a drop only in an attitude representing the usual handling position (administratively controlled) is necessary. If dose rate values of the package are higher than maximum allowable values for a public transport, then it is necessary that additional shielding construction units are attached to the large component. Ashford Ramtrans Publ. 2009 Packaging, transport, storage & security of radioactive materials (RAMTRANS) 20 4 149 153 10.1179/174650909X12543085665266 2016-02-19 OPUS4-21239 Beitrag zu einem Tagungsband Komann, Steffen; Droste, Bernhard; Wille, Frank Transport von Großkomponenten aus der Stilllegung kerntechnischer Anlagen - Erfahrungen bei der mechanischen Bewertung Berlin Kerntechnische Gesellschaft e.V. (KTG) 2010 Jahrestagung Kerntechnik 2010 (Proceedings) Jahrestagung Kerntechnik 2010 Berlin, Deutschland 2010-05-04 2010-05-06 Sektion 7 / 706 1 5 2016-02-19 OPUS4-30487 Zeitschriftenartikel Komann, Steffen; Droste, Bernhard; Wille, Frank Dicht und sicher abgeschirmt BAM-GGR Transportverpackungen für radioaktive Stoffe müssen den BAM-Gefahrgutregeln entsprechen. Ein kurzer Einblick in die Richtlinien. München Vogel 2014 Gefahrgut 3 18 20 2016-02-20 OPUS4-23383 Beitrag zu einem Tagungsband Komann, Steffen; Droste, Bernhard; Wille, Frank Transport of large nuclear power plant components - experiences in mechanical design assessment PATRAM 2010 PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials (Proceedings) PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials London, UK 2010-10-03 2010-10-08 1 8(?) 2016-02-19