TY - JOUR A1 - Komann, Steffen A1 - Wille, Frank T1 - Zur Prüfung verpflichtet - Die BAM-Gefahrgutregel 016 (BAM-GGR 016) N2 - 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. KW - Gefahrgut KW - Radioaktive Stoffe KW - Normen KW - Regeln PY - 2018 SN - 0944-6117 VL - 2018 IS - 1-2 SP - 10 EP - 12 PB - Springer Fachmedien München GmbH CY - München AN - OPUS4-44168 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen T1 - Verpackungen für den Transport radioaktiver Stoffe T2 - 14. Gefahrgut-Technik-Tage CY - Berlin, Germany DA - 2013-11-11 PY - 2013 AN - OPUS4-29754 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Gröke, Carsten T1 - Verpackungen Bestandsaufnahme - Pflichten und Vorschriften bei schwach radioaktiven Stoffen PY - 2012 SN - 0944-6117 IS - 6 SP - 20 EP - 25 PB - Vogel CY - München AN - OPUS4-26362 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen T1 - Unterstützende Richtlinien und Normen in der sicherheitstechnischen Behälterprüfung T2 - 2. RAM-Behältersicherheitstage, BAM CY - Berlin, Germany DA - 2010-03-18 PY - 2010 AN - OPUS4-21322 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Neumann, Martin A1 - Wille, Frank T1 - Transportbehälter für Klasse 7 - Ein Überblick N2 - Die Präsentation gibt einen Überblick über die Tranportbehälter der Klasse 7. Behandelt werden neben der Versandstückklassifikation auch die Anforderungen an diese Verpackungen gemäß der gefahrgutrechtlichen Vorschriften. T2 - 30. Münchner Gefahrguttage CY - Online meeting DA - 20.07.2020 KW - Gefahrgutrecht KW - Transportbehälter KW - Radioaktives Material KW - Zulassungspflichtige Versandstücke PY - 2020 AN - OPUS4-51104 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Komann, Steffen A1 - Wille, Frank A1 - Droste, Bernhard T1 - Transport von Großkomponenten aus der Stilllegung kerntechnischer Anlagen in Deutschland-Erfahrungen bei der mechanischen Intergritätsbewertung N2 - Im Zuge des Rückbaus kerntechnischer Anlagen in Deutschland müssen u. a. Großkomponenten (Dampferzeuger, Reaktordruckbehälter) über öffentliche Verkehrswege in Zwischenlager transportiert werden, in denen sie zerlegt bzw. zwischengelagert werden. Da es sich hierbei um aktivierte und/oder oberflächenkontaminierte Objekte handelt, ist eine Begutachtung unter Berücksichtigung der Gefahrgutbeförderungsvorschriften notwendig. In den meisten Fällen handelt es sich um oberflächenkontaminierte Gegenstände, welche als Industrieversandstücke des Typs IP-2 zu befördern sind. KW - Zwischenlagerung KW - Entsorgung KW - Kerntechnik KW - Radioaktive Stoffe PY - 2016 SN - 0005-6650 VL - Mai 2016 IS - Band 91 SP - 188 EP - 192 PB - Springer VDI Verlag CY - Düsseldorf AN - OPUS4-36852 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Komann, Steffen A1 - Wille, Frank A1 - Droste, Bernhard T1 - Transport von Großkomponenten aus der Stilllegung kerntechnischer Anlagen in Deutschland-Erfahrungen bei der mechanischen Integritätsbewertung N2 - Im Zuge des Rückbaus kerntechnischer Anlagen in Deutschland müssen u. a. Großkomponenten (Dampferzeuger, Reaktordruckbehälter) über öffentliche Verkehrswege in Zwischenlager transportiert werden, in denen sie zerlegt bzw. zwischengelagert werden. Für diese Versandstücke ist eine mechanische Integritätsbewertung, nach den Vorschriften der Internationalen Atomenergie-Organisation (IAEO), notwendig, die in der Regel unter dem Einsatz einer komplexen numerischen Berechnung erfolgt. ---------------------------------------------------------- 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. For these packages a safety evaluation considering the International Atomic Energy Agency (IAEA) transport regulations is necessary, preferably by a complex numerical analysis. KW - Radioaktive Stoffe KW - Transport KW - Mechanik KW - Numerische Analyse PY - 2016 SN - 0005-6650 VL - 91 IS - 5 SP - 188 EP - 192 PB - Springer VDI Verlag CY - Düsseldorf AN - OPUS4-36016 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank T1 - Transport von Großkomponenten aus der Stilllegung kerntechnischer Anlagen - Erfahrungen bei der mechanischen Bewertung T2 - Jahrestagung Kerntechnik 2010 CY - Berlin, Germany DA - 2010-05-04 PY - 2010 AN - OPUS4-21228 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank T1 - Transport von Großkomponenten aus der Stilllegung kerntechnischer Anlagen - Erfahrungen bei der mechanischen Bewertung T2 - Jahrestagung Kerntechnik 2010 CY - Berlin, Deutschland DA - 2010-05-04 KW - Kerntechnische Anlagen KW - Rückbau KW - Mechanische Bewertung KW - Großkomponenten KW - Numerische Analyse PY - 2010 IS - Sektion 7 / 706 SP - 1 EP - 5 CY - Berlin AN - OPUS4-21239 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank T1 - Transport of large nuclear power plant components: experiences in mechanical design assessment N2 - 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. KW - Radioactive material KW - Large components KW - Decommissioning KW - Nuclear power plant KW - Mechanical assessment KW - Numerical analysis PY - 2009 U6 - https://doi.org/10.1179/174650909X12543085665266 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 20 IS - 4 SP - 149 EP - 153 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-20820 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank T1 - Transport of Large Nuclear Power Plant Components - Experiences in Mechanical Design Assessment - T2 - Workshop BAM/US.NRC CY - Washington, D.C., USA DA - 2011-07-11 PY - 2011 AN - OPUS4-24230 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank T1 - Transport of Large Nuclear Power Plant Components - Experiences in Mechanical Design Assessment - T2 - Radioactive Materials Transport Conference 2009 CY - Manchester, England DA - 2009-05-12 PY - 2009 AN - OPUS4-19220 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank T1 - Transport of Large Nuclear Power Plant Components - Experiences in Mechanical Design Assessment - T2 - 16.th International Symposium on the Packaging and Transport of Radioactive Materials CY - London, England DA - 2010-10-03 PY - 2010 AN - OPUS4-22216 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank T1 - Transport of large nuclear power plant components - experiences in mechanical design assessment T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Mechanische Bewertung KW - Radioaktives Material KW - Numerische Analyse PY - 2010 SP - 1 EP - 8(?) AN - OPUS4-23383 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank T1 - Transport of large nuclear power plant components - experiences in mechanical design assessment T2 - Ramtransport 09 - 8th International Conference on radioactive materials transport 2009 CY - Manchester, Great Britain DA - 2009-05-12 KW - Rückbau KW - Kerntechnische Anlagen KW - Großkomponenten KW - Transport KW - Mechanik PY - 2009 IS - Session 5 / PAP 12 SP - 1 EP - 8 AN - OPUS4-20557 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank A1 - Linnemann, Konrad T1 - Transport of Large Nucklear Power Plant Components - Experiences in Mechanical Desig Assessment - T2 - 3. RAM-Behältersicherheitstage, BAM CY - Berlin, Germany DA - 2012-03-22 PY - 2012 AN - OPUS4-26336 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Neumann, Martin A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Weber, Mike A1 - Qiao, Linan A1 - Droste, Bernhard T1 - Static and dynamic calculation approaches for mechanical design assessment of type B packages for radioactive material transport - 10193 T2 - WM2010 Conference CY - Phoenix, Arizona, USA DA - 2010-03-07 KW - Mechanische Bewertung KW - Radioaktives Material KW - Numerische Analyse KW - Typ B(U)-Verpackung PY - 2010 SP - 1 EP - 9 AN - OPUS4-21240 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Neumann, Martin A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Weber, Mike A1 - Droste, Bernhard A1 - Qiao, Linan T1 - Static and Dynamic Calculation Approaches for Mechanical Design Assessment of Type B Packages for Radioactive Material Transport T2 - Waste Management Conference 2010 CY - Phoenix, AZ, USA DA - 2010-03-07 PY - 2010 AN - OPUS4-21227 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Droste, Bernhard A1 - Koch, Frank A1 - Schubert, Sven A1 - Komann, Steffen T1 - State-of-the-art assessment of package design safety analyses T2 - PATRAM 2007, 15th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Miami, Florida, USA DA - 2007-10-21 KW - Transportation of radioactive materials KW - Safety KW - Design assessment KW - Mechanical tests KW - Thermal tests KW - Drop tests KW - Numerical calculations KW - Quality assurance PY - 2007 SP - 1 EP - 8 PB - Institute of Nuclear Materials Management AN - OPUS4-16105 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Müller, Lars A1 - Komann, Steffen A1 - Wille, Frank A1 - Droste, Bernhard T1 - Sicherheitstechnische Begutachtung der strukturmechanischen Auslegung von Castor-Behältern N2 - Die Sicherheitsanforderungen, die an Behälter für den Transport radioaktiver Stoffe auf öffentlichen Verkehrswegen gestellt werden, orientieren sich an dem Gefährdungspotenzial des radioaktiven Inhalts. So müssen für den Transport abgebrannter Brennelemente oder hochradioaktiver Abfalle aus nuklearen Anlagen Behälter eingesetzt werden, die auch schweren Unfällen standhalten. Zu ihnen gehören z. B. die CASTOR®-Behälter (Cask for Storage and Transport of Radioactive Material). T2 - BTU Stahlbau-Symposium 2014 CY - Cottbus, Germany DA - 23.05.2014 PY - 2014 SN - 1611-5023 N1 - Serientitel: Schriftenreihe Stahlbau – Series title: Schriftenreihe Stahlbau IS - 8 SP - 33 EP - 40 AN - OPUS4-31148 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Droste, Bernhard A1 - Zencker, Uwe A1 - Komann, Steffen T1 - Reflektion der verkehrsrechtlichen Bauartprüfung des CASTOR HAW 28M aus Sicht der zuständigen Behörde BAM T2 - Jahrestagung Kerntechnik 2010 CY - Berlin, Deutschland DA - 2010-05-04 KW - Bauartprüfung KW - Fallprüfungen KW - Transportbehälter KW - FEM PY - 2010 SP - 1 EP - 5 CY - Berlin AN - OPUS4-21241 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Gröke, Carsten T1 - Qualitätssicherung und -überwachung bei Verpackungen für prüfpflichtige Versandstücke zur Beföderung radioaktiver Stoffe N2 - Ziel dieses Beitrages soll es sein, zum einen das Vorgehen bei der Erlangung einer Anerkennung zur qualitätsgesicherten Herstellung einer Verpackung für prüfpflichtige Versandstücke der zuständigen Behörde im Sinne des ADR Unterabschnittes 1.7.3 'Qualitätssicherung' darzustellen und zum anderen Probleme bei der Umsetzung der rechtsverbindlichen Vorschriften des Gefahrgutrechts aufzuzeigen. T2 - Jahrestagung Kerntechnik 2013 CY - Berlin, Germany DA - 14.05.2013 PY - 2013 IS - Sektion 5: Entsorgung radioaktiver Abfälle, Lagerung SP - 1 EP - 6 AN - OPUS4-28653 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen T1 - Qualitätssicherung prüfpflichtiger Versandstücke T2 - 4. RAM-Behältersicherheitstage CY - Berlin, Germany DA - 2014-03-26 PY - 2014 AN - OPUS4-30710 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen T1 - Qualitätssicherung pfüfpflichtiger Versandstücke T2 - 4. RAM-Behältersicherheitstage CY - Berlin, Germany DA - 2014-03-26 PY - 2014 AN - OPUS4-30679 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Gröke, Carsten A1 - Droste, Bernhard T1 - Prüfung der qualitätssichernden Maßnahmen an Verpackungen püfpflichtiger Versandstücke durch die BAM T2 - 3. RAM-Behältersicherheitstage, BAM CY - Berlin, Germany DA - 2012-03-22 PY - 2012 AN - OPUS4-26335 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Günther, U. A1 - Baden, M. A1 - Komann, Steffen A1 - Nitz, Thilo T1 - Performance and restrictions of non-destructive testing (NDT) within the quality surveillance during manufacturing of type B-packages T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Mechanische Bewertung KW - Radioaktives Material KW - Numerische Analyse KW - Qualitätsüberwachung PY - 2010 SP - 1 EP - 8 AN - OPUS4-23675 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Quercetti, Thomas A1 - Scheidemann, Robert A1 - Komann, Steffen A1 - Ballheimer, Viktor A1 - Wille, Frank T1 - Package testing of a dual purpose cask for SNF from German research reactors N2 - A new dual purpose cask design was developed for the safe transport and interim storage of spent fuel elements of German research reactors. In the framework of the safety assessment within the licensing procedure the Bundesanstalt für Materialforschung und –prüfung (BAM) as competent authority performed a series of drop tests according to the IAEA-Regulations [1]. The package consists of a cylindrical thick-walled ductile cast iron cask body closed by a double lid system with metallic seals. A lid and bottom sided impact limiter consisting of a wood/steel construction limit the mechanical impact loading. The full-scale test specimen was equipped with a basket and assembled with dummy-fuel elements. The package and test specimen, respectively have a total mass of approximately 24 metric tons. The mechanical test program included three 9m free drop tests, in horizontal, vertical and oblique cask orientation onto the lid system. Additionally, a 1m-puncture drop test followed the horizontal drop test to consider an IAEA-test sequence. The horizontal and vertical drop tests were performed at a temperature of minus 40°C. During the oblique drop test the upper impact limiter was heated to +80°C. The tests were conducted onto an unyielding target, fulfilling the requirements of the IAEA regulations [1]. The test specimen was considerably instrumented with strain gauges and accelerometers. Transient strains at selected locations of the inner and outer container walls, of the primary and secondary lid, as well as of the corresponding lid bolts were measured during the drop tests. Furthermore, decelerations in different locations at the cask body and the lids were measured. The complex geometrical deformation of the impact limiters due to the impact were determined by optical 3d- measurements using the projected fringe method in combination with multi-image photogrammetry. Before and after the drop tests the leakage rate of the lid system was determined by helium leakage testing. The experimental results contribute to the evaluation of the package response to mechanical tests, demonstrating safety under normal and accident conditions of transport. Especially to the verification of the dynamic finiteelement model of the package used in the package design safety report [2]. The paper describes the performance of the drop tests, selected test results focusing on the lid screws and the cask body and the deformation of the impact limiters as well as impact kinematics, respectively. T2 - IHLRWM2019 CY - Knoxville, TN, USA DA - 14.04.2019 KW - Drop test KW - Package testing KW - Dual purpose cask PY - 2019 SN - 978-0-89448-761-3 VL - 2019 SP - paper 27283, 1 EP - 7 PB - ANS AN - OPUS4-50619 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 - Komann, Steffen A1 - Linnemann, Konrad A1 - Wille, Frank A1 - Reiche, Ingo A1 - Ramsay, Jeff 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 - 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 - 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 - JOUR A1 - Komann, Steffen A1 - Neumann, Martin A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Weber, Mike A1 - Qiao, Linan A1 - Droste, Bernhard T1 - Mechanical assessment within type B packages approval: application of static and dynamic calculation approaches N2 - This paper demonstrates exemplarily how numerical and experimental approaches can be combined reasonably in mechanical assessment of package integrity according to the IAEA regulations. The paper also concentrates on the question about how static mechanical approaches can be applied, and what their problems are in relation to dynamic calculation approaches. Under defined impact tests, which represent accident transport conditions, the package has to withstand impact loading, e.g. resulting from a 9 m free drop onto an unyielding target in sequence with a 1 m puncture drop test. Owing to the local character of the interaction between the puncture bar and the cask body, it is possible to develop a dynamic numerical model for the 1 m puncture drop which allows an appropriate simulation of the interaction area. Results from existing experimental drop tests with prototype or small scale cask models can be used for verification and validation of applied analysis codes and models. The link between analysis and experimental drop testing is described exemplarily by considering a regulatory 1 m puncture bar drop test onto the cask body of a recently approved German high level waste transport package. For the 9 m drop test of the package, it is difficult to develop a dynamic numerical model of the package due to the complexity of the interaction between cask body, impact limiters and unyielding target. Dynamic calculations require an extensive verification with experimental results. The simulation of a 9 m drop of a package with impact limiters is thereby often more complex than the simulation of a 1 m puncture drop onto the cask body. A different approximation method can be applied for the consideration of dynamic effects on the impact loading of the package. In a first step, maximum impact force and rigid body deceleration of the cask body during the impact process can be calculated with simplified numerical tools. This rigid body deceleration can subsequently be applied on a verified static numerical model. Dynamic effects, which cannot be covered by the static numerical analysis, have therefore to be considered by using an additional dynamic factor. The paper describes this approach exemplarily for a 9 m horizontal drop of a typical spent fuel cask design. KW - Package assessment KW - Package KW - Packaging KW - Structural analysis PY - 2011 U6 - https://doi.org/10.1179/1746510911Y.0000000012 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 22 IS - 4 SP - 179 EP - 183 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-26448 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Neumann, Martin A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Weber, Mike A1 - Qiao, Linan A1 - Droste, Bernhard T1 - Mechanical Assessment within Type B Packages Approval, The Application of Static and Dynamic Calculation Approaches T2 - 16th International Symposium on the Packaging and Transport of Radioactive Materials CY - London, England DA - 2010-10-03 PY - 2010 AN - OPUS4-22252 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Neumann, Martin A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Weber, Mike A1 - Qiao, Linan A1 - Droste, Bernhard T1 - Mechanical assessment within type B packages approval - the application of static and dynamic calculation approaches T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Mechanische Bewertung KW - Radioaktives Material KW - Numerische Analyse PY - 2010 SP - 1 EP - 8(?) AN - OPUS4-23384 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank A1 - Rittscher, D. T1 - Mechanical assessment of large nuclear power plant components in transport licensing T2 - 49. INMM-Konferenz CY - Nashville, TN, USA DA - 2008-07-13 PY - 2008 AN - OPUS4-17760 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Droste, Bernhard A1 - Wille, Frank A1 - Rittscher, D. T1 - Mechanical assessment of large nuclear power plant components in transport licensing T2 - 49th INMM Annual Meeting CY - Nashville, TN, USA DA - 2008-07-13 KW - Rückbau KW - Kerntechnische Anlagen KW - Großkomponenten KW - Transport KW - Mechanik PY - 2008 SP - 1 EP - 8 AN - OPUS4-18440 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Kuschke, Christian A1 - Ballheimer, Viktor A1 - Wille, Frank A1 - Komann, Steffen T1 - Mechanical assessment criteria of spent fuel assemblies basket design N2 - Packages for the transport of radioactive material are generally equipped with specific structures (basket) to support the radioactive content in defined position. The safety function of the basket depends on the kind of transported inventory. In case of transport cask for spent fuel, the basket design has to ensure the subcriticality of the fissile material in all conditions of transport in particular. Therefore the evaluation of structural integrity and neutron absorption capability of the basket is an important part of complete safety analysis. Sufficient heat transfer to maintain fuel assembly and cask temperature within allowable limits has to be verified as well. Corrosion resistance is an additional requirement on basket materials owing to contact with water during loading and unloading operations. Computational and experimental methods or their combination along with additional material and component tests can be used to analyse the mechanical and thermal basket behaviour under transport conditions defined in IAEA regulations. By deciding between the analysis methods, the design features (including material selection concept) as well as specific safety function should be accounted. In approval procedures of transport packages for radioactive materials, the competent authority mechanical and thermal safety assessment is carried out in Germany by BAM. Some questions of safety evaluation of basket designs are discussed in this paper based on the BAM experience within approval procedures. The paper focuses primarily on the mechanical behaviour of baskets with regard to the assumptions that have to be used in the criticality safety demonstration. The state of the art methodologies for computational basket stress and deformation analysis as well as for interpretation of drop tests results are presented. T2 - PATRAM 2010 - 16th International symposium on the packaging and transport of radioactive materials CY - London, UK DA - 2010-10-03 KW - Transport conditions KW - Spent fuel assemblies KW - Basket design PY - 2010 SP - 1 EP - 7 (Thursday T36/92) AN - OPUS4-24345 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - 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 - 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 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 - Schubert, Sven A1 - Reichardt, Adrian A1 - Wille, Frank A1 - Komann, Steffen A1 - Neumann, Martin T1 - Introduction of an ageing management approach for packages for the transport of radioactive materials N2 - 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. T2 - INMM & ESARDA Joint Annual Meeting CY - Online meeting DA - 30.08.2021 KW - Dual purpose casks KW - Ageing KW - SSR-6 KW - Interim storage KW - Ageing mechanisms KW - Transport of radaioactive materials PY - 2021 UR - https://www.abstractsonline.com/pp8/#!/10383/presentation/964 SP - 1 EP - 6 PB - Institute of Nuclear Materials Management (INMM) CY - Mount Laurel, NJ AN - OPUS4-53180 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Wille, Frank A1 - Schönfelder, Thorsten T1 - German concept to ensure transportability of SNF packages after interim storage N2 - In Germany the concept of dry interim storage of spent fuel in dual purpose metal casks is implemented, currently for periods of up to 40 years. The casks being used have an approved transport package design in accordance with the international IAEA transport regulations as well. The storage of spent nuclear fuel and high-level waste in dual purpose casks (DPC) is related with the challenge of maintaining safety for transportation over several decades of storage. Besides consideration of aging mechanisms by appropriate design, material selection and operational controls to assure technical reliability by aging management measures, an essential issue is the continuous control and update of the DPC safety case. Not only the technical objects are subject of aging but also the safety demonstration basis is subject of “aging” due to possible changes of regulations, standards and scientific/technical knowledge. To ensure a safe transport in future to a destination which is not known yet (because of not yet existing repository sites) periodical reviews of the Package Design Safety Report (PDSR), in connection with periodic renewals of package design approval certificates, have to be carried out. T2 - International High-Level Radioactive Waste Management Conference, IHLRWM 2017 CY - Charlotte, NC, USA DA - 09.04.2017 KW - Dual purpose casks KW - Long time storage KW - Aging management KW - Approval certificate PY - 2017 AN - OPUS4-39916 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Wille, Frank A1 - Schönfelder, Thorsten T1 - German concept to ensure transportability of SNF packages after interim storage N2 - In Germany the concept of dry interim storage of spent fuel in dual purpose metal casks is implemented, currently for periods of up to 40 years. The casks being used have an approved transport package design in accordance with the international IAEA transport regulations as well. The storage of spent nuclear fuel and high-level waste in dual purpose casks (DPC) is related with the challenge of maintaining safety for transportation over several decades of storage. Besides consideration of aging mechanisms by appropriate design, material selection and operational controls to assure technical reliability by aging management measures, an essential issue is the continuous control and update of the DPC safety case. Not only the technical objects are subject of aging but also the safety demonstration basis is subject of “aging” due to possible changes of regulations, standards and scientific/technical knowledge. To ensure a safe transport in future to a destination which is not known yet (because of not yet existing repository sites) periodical reviews of the Package Design Safety Report (PDSR), in connection with periodic renewals of package design approval certificates, have to be carried out. T2 - International High-Level Radioactive Waste Management Conference, IHLRWM 2017 CY - Charlotte, NC, USA DA - 09.04.2017 KW - Dual purpose casks KW - Interim storage KW - Transportability PY - 2017 SP - 476 EP - 481 AN - OPUS4-39919 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Wille, Frank A1 - Droste, Bernhard T1 - Gap analysis examples from periodical reviews of transport package design safety reports of SNF/HLW dual purpose casks N2 - Storage of spent nuclear fuel and high-level waste in dual purpose casks (DPC) is related with the challenge of maintaining safety for transportation over several decades of storage. Beside consideration of aging mechanisms by appropriate design, material selection and operational controls to assure technical reliability by aging management measures, an essential issue is the continuous control and update of the DPC safety case. Not only the technical objects are subject of aging but also the safety demonstration basis is subject of "aging" due to possible changes of regulations, standards and scientific/technical knowledge. The basic document, defining the transport safety conditions, is the package design safety report (PDSR) for the transport version of the DPC. To ensure a safe transport in future to a destination which is not known yet (because of not yet existing repository sites) periodical reviews of the PDSR, in connection with periodic renewals of package design approval certificates, have to be carried out. The main reviewing tool is a gap analysis. A gap analysis for a PDSR is the assessment of the state of technical knowledge, standards and regulations regarding safety functions of structures, systems and components. T2 - PSAM 12 - Probabilistic safety assessment and management CY - Honolulu, Hawaii, USA DA - 22.06.2014 KW - Dual Purpose Casks KW - Aging KW - Transportation KW - Periodical Review PY - 2014 UR - http://psam12.org/proceedings/paper/paper_259_1.pdf SP - Paper 259, 1 EP - 9 AN - OPUS4-32574 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Wille, Frank A1 - Droste, Bernhard T1 - Gap analysis by periodical reviews of transport package design safety reports of German SNF/HLW dual purpose casks N2 - Storage of spent nuclear fuel and high-level waste in dual purpose casks (DPC) is related with the challenge of maintaining safety for transportation over several decades of storage. Beside consideration of aging mechanisms by appropriate design, material selection and operational controls to assure technical reliability by aging management measures, an essential issue is the continuous control and update of the DPC safety case. Not only the technical objects are subject of aging but also the safety demonstration basis is subject of “aging” due to possible changes of regulations, standards and scientific/technical knowledge. The basic document, defining the transport safety conditions, is the package design safety report (PDSR) for the transport version of the DPC. To ensure a safe transport in future to a destination which is not known yet (because of not yet existing repository sites) periodical reviews of the PDSR, in connection with periodic renewals of package design approval certificates, have to be carried out. The main reviewing tool is a gap analysis. A gap analysis for a PDSR is the assessment of the state of technical knowledge, standards and regulations regarding safety functions of structures, systems and components. T2 - 18th International Symposium on the Packaging and Transportation of Radioactive Materials PATRAM2016 CY - Kobe, Japan DA - 18.09.2016 KW - Radioactiv material KW - Transportation KW - Periodoc inspection PY - 2016 AN - OPUS4-37748 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Wille, Frank A1 - Droste, Bernhard T1 - Gap analysis by periodical reviews of transport package design safety reports of German SNF/HLW dual purpose casks N2 - Storage of spent nuclear fuel and high-level waste in dual purpose casks (DPC) is related with the challenge of maintaining safety for transportation over several decades of storage. Beside consideration of aging mechanisms by appropriate design, material selection and operational controls to assure technical reliability by aging management measures, an essential issue is the continuous control and update of the DPC safety case. Not only the technical objects are subject of aging but also the safety demonstration basis is subject of “aging” due to possible changes of regulations, standards and scientific/technical knowledge. The basic document, defining the transport safety conditions, is the package design safety report (PDSR) for the transport version of the DPC. To ensure a safe transport in future to a destination which is not known yet (because of not yet existing repository sites) periodical reviews of the PDSR, in connection with periodic renewals of package design approval certificates, have to be carried out. The main reviewing tool is a gap analysis. A gap analysis for a PDSR is the assessment of the state of technical knowledge, standards and regulations regarding safety functions of structures, systems and components. T2 - 18th International Symposium on the Packaging and Transportation of Radioactive Materials PATRAM2016 CY - Kobe, Japan DA - 18.09.2016 KW - Radioactive material KW - Gap analysis KW - Periodical review KW - Approved packages PY - 2016 SP - Paper 5004, 1 EP - 9 AN - OPUS4-37641 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - 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 - Reichardt, Adrian A1 - Komann, Steffen A1 - Schubert, Sven A1 - Müller, Lars A1 - Wille, Frank A1 - Neumann, Martin A1 - Gröke, Carsten T1 - Experience regarding ageing management and maintenance of transport packages in Germany N2 - The presentation gives an overview about the current experiences of BAM regarding ageing management of transport packages in Germany. The concept of the BAM-GGR 023 for ageing management of transport packagings will be described. The fundamentals of BAM-GGR 023 will be shown, including the following key aspects: -the organization of an Ageing Management System (AMS), -the issue of an Ageing Management Plan (AMP), -the ageing evaluation as a part of the AMP, -the Ageing Surveillance Program (ASP) as a part of the AMP, -the Ageing Management Documentation (AMD). T2 - First Consultancy Meeting on Drafting the Safety Guide on Ageing Management and Maintenance of Radioactive Material Transport Packages CY - Vienna, Austria DA - 20.11.2023 KW - Ageing Management KW - Dual Purpose Cask KW - BAM-GGR 023 PY - 2023 AN - OPUS4-58863 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Qiao, Linan A1 - Zencker, Uwe A1 - Musolff, André A1 - Komann, Steffen T1 - Dynamic finite element analyses of a spent fuel transport and storage cask with impact limiters by 9 meter drop tests N2 - The 9 meter drop onto an unyielding target is one of the important mechanical tests within the safety assessment of transport casks for radioactive material. In general, the cask is equipped with impact limiters to reduce the dynamic load on the cask body by absorbing a major part of the kinetic energy. The impact limiters are often made of wood or aluminium. In this study an elastic-plastic material model with volume change was used to describe the stress-strain behaviour of wood found in crush tests. For aluminium, an elastic-incremental plastic material model with Cowper-Symonds parameters for strain rate depending material hardening was used to model the adiabatic stress-strain relations measured at specimens at constant ambient temperature. Hereafter simulations with a sophisticated finite element model were carried out and compared with different drop tests. Four drop tests of a half-scale cask model equipped with wood and aluminium impact limiters with different drop positions were selected to investigate the impact limiter behaviour during a 9 meter drop test. All drop tests were simulated with the same FE mesh but under different boundary and initial conditions. T2 - SIMULIA Customer conference CY - Barcelona, Spain DA - 17.05.2011 KW - Cask KW - Drop test KW - Impact limiter KW - Wood KW - Aluminium KW - Dynamics PY - 2011 SP - 932 EP - 944 AN - OPUS4-23706 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - JOUR A1 - Qiao, Linan A1 - Neumann, Martin A1 - Komann, Steffen A1 - Ballheimer, Viktor T1 - Dynamic and quasi-static FE calculation of impact from 9 metre drop of spent fuel transport cask N2 - The drop from 9 m height onto a rigid target is one of the required proofs of safety for packages of radioactive materials. Direct dynamic finite element method (FEM) calculations and combined two step analytical quasi-static finite element (FE) calculations are both applied for the simulation of a 9 m drop test. This paper gives a comparative example of both approaches for the 9 m side drop simulation of the spent fuel transport cask with wood filled impact limiter. The model for dynamic FEM consists of cask body and impact limiter. Detailed material properties and geometry descriptions from each component of the impact limiter are required. The results (stress fields in the cask body) are obtained directly from the calculation. The combined method provides as intermediate results the force-deformation characteristic of impact limiter. The maximum impact limiter force determined by the law of energy conservation during the drop is then - in a second step - applied on the cask body in a quasi-static FE model in order to calculate the stresses. In this paper, the rigid body deceleration and the maximum stress in the middle of the cask body are used for the comparison between the dynamic FEM and the combined method. Similar maximum rigid body deceleration-time curves were obtained by both methods for the horizontal 9 m free fall. Concerning the stress in the cask body the dynamic FEM results oscillate about values calculated by the combined quasi-static approach. If the combined quasi-static approach is used in the safety assessment of a cask, a suitable factor has to be applied on its results to take into account the additional dynamic effects. KW - Drop test KW - Spent fuel transport cask KW - Finite element calculation KW - Dynamic simulation KW - Impact limiter KW - Wood PY - 2009 U6 - https://doi.org/10.1179/174650909X12553336148026 SN - 1746-5095 SN - 1746-5109 SN - 0957-476X VL - 20 IS - 4 SP - 174 EP - 178 PB - Ramtrans Publ. CY - Ashford AN - OPUS4-20819 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -