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 OPUS4-44168 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 OPUS4-36852 Zeitschriftenartikel Komann, Steffen; Wille, Frank; Droste, Bernhard Transport von Großkomponenten aus der Stilllegung kerntechnischer Anlagen in Deutschland-Erfahrungen bei der mechanischen Intergritätsbewertung 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. Düsseldorf Springer VDI Verlag 2016 Bauingenieur Mai 2016 Band 91 188 192 2016-07-14 OPUS4-36016 Zeitschriftenartikel Komann, Steffen; Wille, Frank; Droste, Bernhard Transport von Großkomponenten aus der Stilllegung kerntechnischer Anlagen in Deutschland-Erfahrungen bei der mechanischen Integritätsbewertung 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. Düsseldorf Springer VDI Verlag 2016 Bauingenieur 91 5 188 192 2016-05-13 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 OPUS4-20820 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 OPUS4-23383 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 OPUS4-20557 Beitrag zu einem Tagungsband Komann, Steffen; Droste, Bernhard; Wille, Frank Transport of large nuclear power plant components - experiences in mechanical design assessment The Nuclear Institute (NI), London 2009 RAMTRANSPORT 09 - 8th International Conference on radioactive materials transport 2009 (Proceedings) Ramtransport 09 - 8th International Conference on radioactive materials transport 2009 Manchester, Great Britain 2009-05-12 2009-05-14 Session 5 / PAP 12 1 8 2016-02-19 OPUS4-21240 Beitrag zu einem Tagungsband Komann, Steffen; Neumann, Martin; Ballheimer, Viktor; Wille, Frank; Weber, Mike; Qiao, Linan; Droste, Bernhard Static and dynamic calculation approaches for mechanical design assessment of type B packages for radioactive material transport - 10193 2010 WM2010 Conference (Proceedings) WM2010 Conference Phoenix, Arizona, USA 2010-03-07 2010-03-11 1 9 2016-02-19 OPUS4-16105 Beitrag zu einem Tagungsband Wille, Frank; Droste, Bernhard; Koch, Frank; Schubert, Sven; Komann, Steffen State-of-the-art assessment of package design safety analyses Institute of Nuclear Materials Management 2007 PATRAM 2007 - 15th International symposium on the packaging and transportation of radioactive materials (Proceedings) PATRAM 2007, 15th International Symposium on the Packaging and Transportation of Radioactive Materials Miami, Florida, USA 2007-10-21 2007-10-26 1 8 2016-02-19 OPUS4-31148 Beitrag zu einem Tagungsband Müller, Lars; Komann, Steffen; Wille, Frank; Droste, Bernhard Sicherheitstechnische Begutachtung der strukturmechanischen Auslegung von Castor-Behältern 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). Brandenburgische Technische Universität Cottbus - Senftenberg 2014 BTU Stahlbau-Symposium 2014 BTU Stahlbau-Symposium 2014 Cottbus, Germany 23.05.2014 23.05.2014 8 33 40 2016-02-20 OPUS4-21241 Beitrag zu einem Tagungsband Wille, Frank; Droste, Bernhard; Zencker, Uwe; Komann, Steffen Reflektion der verkehrsrechtlichen Bauartprüfung des CASTOR HAW 28M aus Sicht der zuständigen Behörde BAM Berlin Kerntechnische Gesellschaft e.V. (KTG) 2010 Jahrestagung Kerntechnik 2010 (Proceedings) Jahrestagung Kerntechnik 2010 Berlin, Deutschland 2010-05-04 2010-05-06 1 5 2016-02-19 OPUS4-50619 Beitrag zu einem Tagungsband Quercetti, Thomas; Scheidemann, Robert; Komann, Steffen; Ballheimer, Viktor; Wille, Frank Package testing of a dual purpose cask for SNF from German research reactors 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. ANS 2019 International High-Level Radioactive Waste Management 2019 2019 978-0-89448-761-3 IHLRWM2019 Knoxville, TN, USA 14.04.2019 18.04.2019 paper 27283, 1 7 2020-04-06 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 OPUS4-57750 Beitrag zu einem Tagungsband Komann, Steffen; Linnemann, Konrad; Wille, Frank; Reiche, Ingo; Ramsay, J. New SCO-iii regulations to ship large comonents 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. 2023 PATRAM22 PATRAM22 Juan les Pins, France 11.06.2023 15.06.2023 1 8 2023-06-26 OPUS4-58577 Zeitschriftenartikel Komann, Steffen; Wille, Frank; Schubert, Sven Neue Regel für das Alterungsmanagement 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. München Heinrich Vogel 2023 Gefahrgut 2023 9 16 18 2023-10-16 OPUS4-26448 Zeitschriftenartikel Komann, Steffen; Neumann, Martin; Ballheimer, Viktor; Wille, Frank; Weber, Mike; Qiao, Linan; Droste, Bernhard Mechanical assessment within type B packages approval: application of static and dynamic calculation approaches 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. Ashford Ramtrans Publ. 2011 Packaging, transport, storage & security of radioactive materials (RAMTRANS) 22 4 179 183 10.1179/1746510911Y.0000000012 2016-02-19 OPUS4-23384 Beitrag zu einem Tagungsband Komann, Steffen; Neumann, Martin; Ballheimer, Viktor; Wille, Frank; Weber, Mike; Qiao, Linan; Droste, Bernhard Mechanical assessment within type B packages approval - the application of static and dynamic calculation approaches 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 OPUS4-18440 Beitrag zu einem Tagungsband Komann, Steffen; Droste, Bernhard; Wille, Frank; Rittscher, D. Mechanical assessment of large nuclear power plant components in transport licensing Institute of Nuclear Materials Management (INMM) 2008 49th INMM Annual Meeting, July 13-17, 2008, Nashville, TN, USA (Proceedings) 49th INMM Annual Meeting Nashville, TN, USA 2008-07-13 2008-07-17 1 8 2016-02-19 OPUS4-24345 Beitrag zu einem Tagungsband Kuschke, Christian; Ballheimer, Viktor; Wille, Frank; Komann, Steffen Mechanical assessment criteria of spent fuel assemblies basket design 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. 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 7 (Thursday T36/92) 2016-02-19 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 OPUS4-57770 Beitrag zu einem Tagungsband Schubert, Sven; Reichardt, Adrian; Müller, Lars; Neumann, Martin; Komann, Steffen; Wille, Frank Introduction of the German ageing management guide for packages for transport of radioactive materials 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. London World Nuclear Transport Institute (WNTI) 2023 Proceedings of the 20th international symposium on the packaging and transportation of radioactive materials PATRAM22 - 20th International Symposium on the Packaging and Transportation of Radioactive Materials Juan-les-Pins, France 11.06.2023 15.06.2023 1 10 2023-06-28 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 OPUS4-39919 Beitrag zu einem Tagungsband Komann, Steffen; Wille, Frank; Schönfelder, Thorsten German concept to ensure transportability of SNF packages after interim storage 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. American Nuclear Society 2017 ANS IHLRWM Conference Proceedings International High-Level Radioactive Waste Management Conference, IHLRWM 2017 Charlotte, NC, USA 09.04.2017 13.04.2017 476 481 2017-04-25 OPUS4-32574 Beitrag zu einem Tagungsband Komann, Steffen; Wille, Frank; Droste, Bernhard Gap analysis examples from periodical reviews of transport package design safety reports of SNF/HLW dual purpose casks 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. 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 Paper 259, 1 9 2016-02-20 OPUS4-37641 Beitrag zu einem Tagungsband Komann, Steffen; Wille, Frank; Droste, Bernhard Gap analysis by periodical reviews of transport package design safety reports of German SNF/HLW dual purpose casks 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. 2016 Proceedings of the 18th International Symposium on the Packaging and Transportation of Radioactive Materials 18th International Symposium on the Packaging and Transportation of Radioactive Materials PATRAM2016 Kobe, Japan 18.09.2016 23.09.2016 Paper 5004, 1 9 2016-10-05 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 OPUS4-51546 Beitrag zu einem Tagungsband Müller, Lars; Scheidemann, Robert; Schönfelder, Thorsten; Komann, Steffen; Wille, Frank Drop tests assessment of internal shock absorbers for packages loaded with encapsulations for damaged spent nuclear fuel Damaged spent nuclear fuel (DSNF) can be loaded in German dual-purpose casks (DPC) for transport and interim storage. Encapsulations are needed to guarantee a safe handling and a tight closure, separated from the package enclosure. These encapsulations shall be durable and leak-tight for a long storage period, because they are usually not accessible within periodical inspections of the DPC. Due to the general design of DPCs for standard fuel assemblies, specific requirements have to be considered for the design of encapsulations for DSNF to ensure the loading in existing package designs. Especially the primary lid system of a DPC is designed for maximum loads due to the internal impact of the content during drop test conditions. The main difference of encapsulations for damaged spent nuclear fuel is that they have usually a much higher stiffness than standard fuel assemblies. Therefore the design of an internal shock absorber, e.g. at the head of an encapsulation is required to reduce mechanical loads to the primary lid system during impacts. BAM as part of the German competent authority system is responsible for the safety assessment of the mechanical and thermal package design, the release of radioactive material and the quality assurance of package manufacturing and operation. Concerning the mechanical design of the encapsulation BAM was involved in the comprehensive assessment procedure during the package design approval process. An internal shock absorber was developed by the package designer with numerical analyses and experimental drop tests. Experimental drop tests are needed to cover limiting parameters regarding, e.g. temperature and wall thickness of the shock absorbing element to enable a detailed specification of the whole load-deformation behavior of the encapsulation shock absorber. The paper gives an overview of the assessment work by BAM and points out the main findings which are relevant for an acceptable design of internal shock absorbers. The physical drop tests were planned on the basis of pre-investigations of the applicant concerning shape, dimension and material properties. In advance of the final drop tests the possible internal impact behavior had to be analyzed and the setup of the test facility had to be validated. The planning, performance and evaluation of the final drop tests were witnessed and assessed by BAM. In conclusion it could be approved that the German encapsulation system for damaged spent nuclear fuel with shock absorbing components can be handled similar to standard fuel assemblies in existing package designs. 2020 Pressure Vessels & Piping Conference 2020 Pressure Vessels & Piping Conference 2020 Online meeting 03.08.2020 07.08.2020 1 9 2020-11-16 OPUS4-50622 Beitrag zu einem Tagungsband Quercetti, Thomas; Scheidemann, Robert; Komann, Steffen; Ballheimer, Viktor; Wille, Frank Drop Testing of a New Package Design for the Transport of SNF from German Research Reactors 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 package approval procedure the Bundesanstalt für Materialforschung und -prüfung (BAM) as competent authority performed a series of drop tests according with the IAEA Transport Regulations. The package consists of a cylindrical thick-walled ductile cast iron cask body closed by a bolted 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 drop test program included three 9m free drop tests, in horizontal, vertical and oblique cask orientation onto the lid system. Additionally, a 1m-puncture bar drop test followed the horizontal drop test to consider an IAEA-drop 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 up to +80°C. The tests were conducted onto an unyielding target, fulfilling the IAEA requirements. 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. 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 and especially to the verification of the dynamic finite-element model of the package used in the package design safety report. 2019 Proceedings of the 18th International Symposium on the Packaging and Transportation of Radioactive Materials PATRAM 2019 PATRAM 2019 New Orleans, LA, USA 04.08.2019 09.08.2019 Paper 19-A-1142,1 10 2020-04-02 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 OPUS4-17780 Beitrag zu einem Tagungsband Wille, Frank; Droste, Bernhard; Koch, Frank; Komann, Steffen Design assessment of spent fuel and HLW Transport Casks Institute for Nuclear Materials Management, USA 2008 49th INMM Annual Meeting, July 13-17, 2008, Nashville, USA (Proceedings) 49th INMM Annual Meeting, July 13-17, 2008 Nashville, USA 2008-07-13 2008-07-17 1 8 2016-02-19 OPUS4-48685 Beitrag zu einem Tagungsband Schönfelder, Thorsten; Müller, Lars; Komann, Steffen; Wille, Frank Schönfelder, Thorsten Design assessment of a dual purpose cask for damaged spent nuclear fuel German package design approvals were granted recently for dual purpose casks (DPC) intended for loading with encapsulated damaged spent nuclear fuel (DSNF). Comprehensive assessment procedures were carried out by the authority BAM with respect to the mechanical and thermal package design, the activity release of radioactive material and quality assurance aspects for manufacturing and operation of each packaging. The objective of each procedure was to verify the Package Design Safety Report (PDSR) and the relevant guidelines fulfils the requirements of the IAEA regulations. Previous approvals of German SNF package designs consider mainly standard fuel assemblies with defined specifications and properties for transport and interim storage. Due to the nuclear power phase-out in Germany all kinds of SNF, e.g. damaged spent fuel rods shall be packed in DPC now. Therefore specific requirements shall be considered in accordance with international experiences including IAEA technical reports. The main requirement for DSNF is a tight encapsulation with specific defined properties under transport and storage conditions. Due to the interim storage period of currently up to 40 years the encapsulation with DSNF in the casks shall also be long term durable. Thus specific loading and drying procedures are necessary and had to be qualified during the approval process. BAM assessed these drying procedures and could confirm the long-term behaviour of the encapsulation and the suitability of the drying equipment. This special equipment was qualified in a "cold handling". In addition, it was shown that the behaviour of the test equipment used in the qualification process was comparable with the original equipment, e.g. test fuel rods or test encapsulation. In the development of the drying process, experience was obtained in how to put the requirements of the IAEA regulations and related IAEA technical reports into practice. The paper gives an overview of approval assessment and testing experience made by BAM and point out the main resulting requirements on drying processes for these kinds of encapsulations with DSNF. 2019 Proceedings of the 19th International Symposium on the Packaging and Transportation of Radioactive Materials PATRAM 2019 Proceedings of the 19th International Symposium on the Packaging and Transportation of Radioactive Materials PATRAM 2019 New Orleans, USA 04.08.2019 09.08.2019 Paper 1204, 1 10 2019-08-13 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 OPUS4-32518 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 OPUS4-33428 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 OPUS4-48749 Beitrag zu einem Tagungsband Linnemann, Konrad; Müller, Lars; Schönfelder, Thorsten; Komann, Steffen; Wille, Frank Authority Experience during Design Approval Procedure for Packages Loaded with Special Encapsulations for Damaged Spent Nuclear Fuel The first German package design approval certificate for a dual purpose cask intended for loading with damaged spent nuclear fuel was issued recently. BAM as part of the competent authority system in Germany carried out a comprehensive assessment procedure with respect to the mechanical and thermal design, the release of radioactive material and the quality assurance aspects of manufacturing and operation. Packages for the transport and storage of radioactive material have been assessed by BAM for many years, thus the common assessment procedure is well-known and good practice. Up to now only SNF without defects or HLW with well-defined properties were designated for long-term Interim storage and transports afterwards. Due to Germany's nuclear phase out all other kinds of spent nuclear fuel in particular damaged spent nuclear fuel shall be packed as well. Damaged spent nuclear fuel needs a tight closure with Special encapsulations and clearly defined properties in Germany. In addition, these encapsulations shall be long-term durable, because they are not accessible after loading in a packaging within periodical inspections. The main difference to Standard package components is that encapsulations with a permanent closure achieve their specified conditions not after manufacturing but only during operation, after loading and closing. To ensure compliance with the specific conditions, special measures for quality assurance are necessary during operation of each encapsulation, e.g. drying and sealing, which were assessed by BAM. The present paper gives an overview of the conducted assessment from BAM and point out the findings concerning to the special closure lid of the approved encapsulation, which is screwed and welded. A wide verification concept is necessary to show the specific tightness under transport conditions. Together with quality assurance measures during first operation steps these encapsulations with damaged spent nuclear fuel can be handled like standard fuel assemblies in approved package designs. 2019 Proceeding of the International Conference on the Management of Spent Fuel from Nuclear Power Reactors International Conference on the Management of Spent Fuel from Nuclear Power Reactors Vienna, Austria 24.06.2019 28.06.2019 1 9 2019-10-10 OPUS4-31047 Beitrag zu einem Tagungsband Müller, Lars; Schilling, O.; Ballheimer, Viktor; Wille, Frank; Komann, Steffen Assessment strategy of numerical analyses of RAM package components - 14619 In Germany current package design safety cases include more and more advanced numerical methods, e. g. finite element analysis (FEA), often in combination with local concepts of strength evaluation of the structure. This approach requires extensive modeling and verification procedures. As a consequence the efforts of authority assessment of design safety analysis increase as well. Only the check of pre- and post-data of numerical calculations is often not sufficient for the safety assessment. On the other hand own analyses of the mechanical problem by performing an Independent numerical modeling and analyzing is not always realizable. Therefore it is necessary to look for optimized procedures of the assessment, without loss of safety. This paper shows possibilities for the assessment strategy of numerical analyses with focus on simple analytical approaches as comparative calculations. Such approaches can be helpful to support evaluation of numerical calculations in the whole assessment procedure. Three examples are considered to Show which possibilities and limits exist to support the assessment of numerical analyses using analytical comparative calculations. Two examples of bolt and lid analysis show the influence of component and boundary stiffness on the results. Thickness to length/width ratios are partially exceeded and only fixed or free boundary conditions can be analyzed analytical. Nevertheless These analytical approaches can help to evaluate the numerical results for the assessment. The example of a trunnion demonstrates the limits of analytical approaches. The trunnion shows a complex deformation behavior and local stresses. A single basic theory isn't matching and a construct of several approaches is not useable for calculations of local stresses. Therefore numerical calculations during assessment are necessary. Analytical approaches are not always purposeful but often effective to reduce the effort of assessment for numerical analysis of complex and safety relevant components of RAM packages. WM Symposia, Inc. 2014 WM2014 Conference (Proceedings) 978-0-9836186-3-8 WM2014 Conference Phoenix, Arizona, USA 02.03.2014 06.03.2014 Paper 14619, 1 14 2016-02-20 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 OPUS4-26363 Beitrag zu einem Tagungsband Komann, Steffen; Kiyak, Yusuf; Wille, Frank; Zerbst, Uwe; Weber, Mike; Klingbeil, Dietmar Assessment of ductile cast iron fracture mechanics analysis within licensing of German transport packages New York American Society of Mechanical Engineers 2012 PVP2012 - Proceedings of the ASME 2012 Pressure vessels & piping division conference ASME 2012 Pressure vessels & piping division conference Toronto, Ontario, Canada 2012-07-15 2012-07-19 PVP2012-78213, 1-7 2016-02-19 OPUS4-25022 Beitrag zu einem Tagungsband Komann, Steffen; Kiyak, Yusuf; Wille, Frank; Zerbst, Uwe; Weber, Mike; Klingbeil, Dietmar Assessment of ductile cast iron fracture mechanics analysis within licensing of German transport packages In the approval procedure of transport packages for radioactive materials, the mechanical and thermal safety assessment is carried out in Germany by competent authority BAM. In recent years BAM was involved in several licensing procedures of new spent fuel and HLW package designs, where the cask body was made of Ductile Cast Iron (DCI). According to IAEA regulations package designs have to fulfill requirements for specific conditions of transport. Type B(U) packages must withstand the defined accident conditions of transport. The temperature range from -40°C up to the operational temperature has to be considered. For the cask material DCI, it is necessary to determine the brittle fracture behavior. The German guideline BAM-GGR 007 defines requirements for the fracture mechanics of DCI. Due to complex structure of the cask body and the dynamic loading a fracture mechanical assessment in an analytical kind is not always possible. Experience of recent design approval procedures show that the application of numerical calculations are applicable to determine the fracture mechanical load in the cask body. At the first step a numerical analysis has to be done to identify the loading state at the whole cask body. Secondly an analysis of a detail of the cask body is made considering the displacement boundary conditions of the global model. An artificial flaw is considered in this detailed model to calculate the fracture mechanical loading state. The finite element mesh was strongly refined in the area of flaw. The size of the artificial flaw is characterized by the ultrasonic inspection used for the quality assurance of the package. The applicant developed additional analysis tools for calculation of stress intensity factor and/or J-Integral. The assessment approach by BAM led to the decision to develop own tools to the possibility for independent check of the results. The paper describes the authority assessment approach for the DCI fracture mechanics analysis. The validation procedure incl. the development of own tools is explained. BAM developed a post-processor called JINFEM to determine the fracture mechanical loads. The regulatory 1 m puncture bar drop test is used to give an example of the assessment procedure. Institute of Nuclear Materials Management (INMM) 2011 52nd INMM Annual meeting (Proceedings) 52nd INMM Annual meeting Palm Desert, CA, USA 17.07.2011 21.07.2011 1 10 2016-02-19 OPUS4-30175 Beitrag zu einem Tagungsband Komann, Steffen; Kiyak, Yusuf; Wille, Frank; Zerbst, Uwe; Weber, Mike; Klingbeil, Dietmar Assessment of ductile cast iron fracture mechanics analysis within licensing of German transport packages In the design approval of transport packages for radioactive materials, the mechanical and thermal safety assessment is carried out in Germany by competent authority BAM. In recent years BAM was involved in several licensing procedures of new spent fuel and HLW package designs, where the cask body is of Ductile Cast Iron (DCI). According to IAEA regulations package designs have to fulfill requirements for specific conditions of transport. Type B(U) packages must withstand the defined accident conditions of transport. The temperature range from -40°C up to the operational temperature has to be considered. For the cask material DCI, it is necessary to determine safety against brittle fracture. The German guideline BAM-GGR 007 defines requirements for fracture mechanics of packagings made of DCI. Due to complex cask body structure and the dynamic loading a fracture mechanical assessment by analytical approaches is not always possible. Experience of recent design approval procedures show that the application of numerical calculations are applicable to determine the stresses and stress intensity factors in the cask body. At the first step a numerical analysis has to be done to identify the loading state at the whole cask body. Secondly an analysis of a detail of the cask body is made considering the displacement boundary conditions of the global model. An artificial flaw is considered in this detailed model to calculate the fracture mechanical loading state. The finite element mesh was strongly refined in the area of the flaw. The size of the artificial flaw is based on the ultrasonic inspection acceptance criteria applied for cask body manufacture. The applicant (GNS) developed additional analysis tools for calculation of stress intensity factor and/or J-Integral. The assessment approach by BAM led to the decision to develop own tools to the possibility for independent proof of the results. The paper describes the authority assessment approach for DCI fracture mechanics analysis. The validation procedure incl. the development of own tools is explained. BAM developed a postprocessor to determine the fracture mechanical loads. A horizontal 1 m puncture bar drop test is used to give a detailed description of the assessment procedure. Omnipress Institute of Nuclear Materials Management (INMM) 2013 PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials (Proceedings) PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials San Francisco, CA, USA 18.08.2013 23.08.2013 1 10 2016-02-20 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 OPUS4-56343 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 OPUS4-31048 Beitrag zu einem Tagungsband Schilling, O.; Müller, Lars; Ballheimer, Viktor; Komann, Steffen; Wille, Frank Analytical approaches within the assessment strategy of numerical analyses of components of RAM packages Load attachment points, consisting of the lifting component and corresponding bolt connections, and lid systems, consisting of a lid, gaskets and bolt connections, are usually analyzed numerically using the Finite Element Method. Reasons for applying the FEM are complex geometry, specific load distribution and the violation of application criteria for analytical approaches. For example the elementary beam theory is not suitable for the calculation of compact trunnions and the Kirchhoff plate theory is not suitable for the calculation of lids with a large thickness to diameter ratio because the Bernoulli-hypothesis is violated. In addition for structural integrity assessments often local stress and for the evaluation of lid systems tightness local contact opening are needed, which can be obtained only through accurately discretized numerical models. Independent comparative calculations are essential for complex calculations. Simplified but appropriate analytical approaches are an efficient way of examination. Is it possible to provide conservative analytical estimates of such kind of analysis problems? Are analytical calculations an appropriate approach in today's authority assessment business? Using the example of bolted trunnion and lid systems of a heavy package the possibilities and limits of analytical comparative calculations are shown. Omnipress Institute of Nuclear Materials Management (INMM) 2013 PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials (Proceedings) PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials San Francisco, CA, USA 18.08.2013 23.08.2013 Paper 217, 1 9 2016-02-20 OPUS4-52493 Vortrag Wille, Frank; Komann, Steffen Aktuelle Entwicklungen und Prüfungen der BAM Es werden im Rahmen des Vortrages die aktuellen Prüftätigkeiten und Arbeitsschwerpunkte der BAM vorgestellt. Des Weiteren wird darüber berichtet was es Neues in der aktuellen Normen- und Richtlinienentwicklung gibt. 2021 Informationsveranstaltung zum Transport radioaktiver Stoffe Online meeting 14.04.2021 14.04.2021 2021-04-26 OPUS4-50969 Beitrag zu einem Tagungsband Rolle, Annette; Komann, Steffen; Wille, Frank Baraldi, P.; Di Maio, F.; Zio, E. Ageing aspect in the safety evaluation of special form radioactive material In accordance with the IAEA transport regulations Special Form Radioactive Material (SFRM) is either an indispersible solid radioactive material or a sealed capsule containing radioactive material. The design of special form radioactive material has to resist a severe transport accident without undue loss or dispersal of radioactive material. The safety assessment for design approval shall include besides the required test program (impact, percussion, bending and heat test) also the evaluation of the management system for design, manufacture, testing, documentation, use, maintenance and inspection. The specified quality assurance measures have to assure, that every specimen of the approved design is produced in the same verified quality and every specimen must be able to survive the severe mechanical and thermal tests at any time of its working life. Due to the long-term use of SFRM the consideration of ageing is an important aspect in the approval procedure by BAM, the competent authority for SFRM approval in Germany. Different fields of application imply a wide range of environmental conditions, from clean room atmosphere to highly aggressive industrial conditions. So, besides of radioactive content, corrosion is a main factor for possible design degradation. This paper will describe major influencing factors to be taken into account to assess the ageing behavior of a SFRM design and will emphasize that there is a need for a regulatory specification of a SFRM-working life as basis for the aging evaluations. Singapore Research Publishing European Safety and Reliability Association / International Association for Probalistic Safety Assessment and Management 2020 e-proceedings of the 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL2020 PSAM15) 987-981-14-8593-0 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL2020 PSAM15) Online meeting 01.11.2020 06.11.2020 Paper 3659,1 5 2020-07-03 OPUS4-57781 Vortrag Rolle, Annette; Komann, Steffen; Wille, Frank Ageing aspect in design approval of special form radioactive material In accordance with the IAEA transport regulations, the design of special form radioactive material (SFRM) shall resist a severe transport accident without undue loss or dispersal of radioactive material. The safety assessment for design approval includes besides the program for physical tests (impact, percussion, bending and heat test) also the evaluation of the management system for design, manufacture, testing, documentation, use, maintenance, and inspection. SFRM source design plus management system shall ensure, that every specimen of the approved design is able to survive the severe mechanical and thermal tests at any time of its SFRM-working life. Due to the long-term use of SFRM designs in most cases, the assessment of the source ageing behavior is an important aspect in the approval procedure. Different fields of application imply a wide range of environmental conditions, from clean room atmosphere to highly aggressive industrial conditions. Besides of radioactive content, corrosion is a main factor for possible SFRM design degradation. Although the IAEA Advisory Material SSG-26 already implies an indication of the need for considering ageing mechanisms, suitable amendments in the regulatory requirements of SSR-6 should be introduced to make the approval procedure more transparent and help to reduce rounds of questions by the authority. A supplementary requirement for considering of ageing mechanisms could be a helpful contribution to an international harmonization of the approval procedure. This paper will describe major influencing factors to be considered to assess the ageing behavior of a SFRM design and will identify the need for a regulatory specification of a SFRM-working life as basis for the assessment of the SFRM design regarding time-dependent weakening. A proposal for an explicit requirement for consideration of ageing mechanisms in safety assessment of SFRM, which should be considered in the ongoing SSR-6 revision cycle, will be explained. 2023 PATRAM22 Conference Juan-Les-Pins, France 11.06.2023 15.06.2023 2023-06-28 OPUS4-57786 Beitrag zu einem Tagungsband Rolle, Annette; Komann, Steffen; Wille, Frank Ageing aspect in design approval of special form radioactive material In accordance with the IAEA transport regulations, the design of special form radioactive material (SFRM) shall resist a severe transport accident without undue loss or dispersal of radioactive material. The safety assessment for design approval includes besides the program for physical tests (impact, percussion, bending and heat test) also the evaluation of the management system for design, manufacture, testing, documentation, use, maintenance, and inspection. SFRM source design plus management system shall ensure, that every specimen of the approved design is able to survive the severe mechanical and thermal tests at any time of its SFRM-working life. Due to the long-term use of SFRM designs in most cases, the assessment of the source ageing behavior is an important aspect in the approval procedure. Different fields of application imply a wide range of environmental conditions, from clean room atmosphere to highly aggressive industrial conditions. Besides of radioactive content, corrosion is a main factor for possible SFRM design degradation. Although the IAEA Advisory Material SSG-26 already implies an indication of the need for considering ageing mechanisms, suitable amendments in the regulatory requirements of SSR-6 should be introduced to make the approval procedure more transparent and help to reduce rounds of questions by the authority. A supplementary requirement for considering of ageing mechanisms could be a helpful contribution to an international harmonization of the approval procedure. This paper will describe major influencing factors to be considered to assess the ageing behavior of a SFRM design and will identify the need for a regulatory specification of a SFRM-working life as basis for the assessment of the SFRM design regarding time-dependent weakening. A proposal for an explicit requirement for consideration of ageing mechanisms in safety assessment of SFRM, which should be considered in the ongoing SSR-6 revision cycle, will be explained. 2023 Proceedings of the 20th International Symposium on the Packaging and Transportation of Radioactive Materials iio PATRAM22 Conference Juan-Les-Pins, Antibes, France 11.06.2023 15.06.2023 1 6 2023-06-29