TY - CONF A1 - Linnemann, Konrad A1 - Ballheimer, Viktor A1 - Sterthaus, Jens A1 - Wille, Frank T1 - Methodological aspects for numerical analysis of lid systems for SNF and HLW transport packages N2 - The regulatory compliance of the containment system is of essential importance for the design assessment of transport packages for radioactive materials. The requirements of the IAEA transport regulations SSR-6 for accident conditions implies high load on the containment system of Type B(U) packages. The integrity of the containment system has to be ensured under the mechanical and thermal tests. The containment system of German transport packages for spent nuclear fuel (SNF) and high level waste (HLW) usually includes bolted lids with metal gaskets. BAM Federal Institute for Materials Research and Testing as the German competent authority for the mechanical and thermal design assessment of approved transport packages has developed the guideline BAM-GGR 012 for the analysis of bolted lid and trunnion systems. According to this guideline the finite element (FE) method is recommended for the calculations. FE analyses provide more accurate and detailed information about loading and deformation of such kind of structures. The results allow the strength assessment of the lid and bolts as well as the evaluation of relative displacements between the lid and the cask body in the area of the gasket groove. This paper discusses aspects concerning FE simulation of lid systems for SNF and HLW transport packages. The work is based on the experiences of BAM within safety assessment procedures. The issues considered are the assessment methods used in the BAM-GGR 012 for bolted lid systems along with the nominal stress concept which is applied for bolts according to that guideline. Additionally, modeling strategies, analysis techniques and the interpretation of the results are illustrated by the example of a generalized bolted lid systems under selected accident conditions of transport. T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 PY - 2013 SP - 1 EP - 10 PB - Omnipress AN - OPUS4-30134 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Erhard, Anton A1 - Völzke, Holger A1 - Droste, Bernhard A1 - Wolff, Dietmar A1 - Wille, Frank T1 - Nuclear power plant waste management strategy in Germany N2 - The shut-down of eight nuclear power plants and the phase out of the nuclear energy genera� tion in Germany, due to a decision of the German government after the Fukushima accident, has rapidly forced the need for transportation and storage of spent nuclear fuel (SNF) from shut-down reactors in this country. In accordance with the present strategy in Germany, which is characterized by searching for appropriate final repository areas, only the dry interim stor� age facilities are available for storage nuclear waste of the decommissioned plants as well as the operating plants. Therefore, the transportation and the storage casks must be available for both types of light water reactors. But for the time being this is the bottleneck. Unfortunately the prognoses are neglecting the fact, that such a high number of casks are needed in a very short space of time. This requires the acceleration of formal activeness in relation to the li� censes but also an increase of the production. In the meanwhile to get this process started, the only possibility is the storage in the pools at the plant site. Anyway, for the time being, this is one of the problems for dismantling the eight plants which have been shut down due to the decision of the Government, which is now fixed in the last amended German Atomic Energy Act dated 24th of February 2012. Starting to dismantle in the radioactive regions of the plants requires empty pools. For emptying the pools of damaged SNF special procedures regarding the transportation and storage are necessary. The transportation of damaged SNF requires special casks or at least special procedures and therefore also a special licenses. This paper describes the challenge for the management in transportation and storage of SNF in Germany after the shut-down of the nuclear power plants by reason of the Fukushima acci� dent. T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 PY - 2013 SP - 1 EP - 11 PB - Omnipress AN - OPUS4-30135 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Komann, Steffen A1 - Kiyak, Yusuf A1 - Wille, Frank A1 - Zerbst, Uwe A1 - Weber, Mike A1 - Klingbeil, Dietmar T1 - Assessment of ductile cast iron fracture mechanics analysis within licensing of German transport packages N2 - 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. T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 PY - 2013 SP - 1 EP - 10 PB - Omnipress AN - OPUS4-30175 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Droste, Bernhard A1 - Wille, Frank T1 - Maintenance of package design safety report effectiveness of SNF and HWL dual purpose casks N2 - For interim storage of spent fuel or HLW in many countries transport casks are used. The design of these “dual purpose casks” has to be assessed and approved to transport regulations (based on IAEA SSR-6), and to be assessed within the storage facility licensing procedure. Although the transport cask design differs from the storage cask design, e.g. by use of impact limiters, the majority of cask components is identical for both. Differences occur also in the acceptance criteria; these are for the transport case defined in IAEA SSR-6, and have to be developed for the storage case based on the storage conditions. Considering transport after several decades of storage requires the implementation of ageing behavior into the transport safety case. Additionally the transport package design safety case has to be maintained in an up-to-date state, considering potential regulatory changes and development of scientific and technical knowledge. The review of a transport package design safety case has to be done periodically, implemented in periodic re-assessment for extension of the package design approval certificate. The review process (as a kind of “intellectual periodic inspection”) should be part of the approved applicant’s management system. From experience we have seen that stability of regulatory requirements for Type B(U) packages was not a major problem, but consideration of ageing and developments regarding the state-of-the-art technology can cause necessary adjustments of specific technical evaluation, with the result of confirmation of package safety, or with the development of appropriate compensatory measures to reach the required level of safety. This paper will draw some examples from the past to explain the problems in periodic transport safety case reviews. T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 PY - 2013 SP - 1 EP - 9 PB - Omnipress AN - OPUS4-30138 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Wolff, Dietmar A1 - Droste, Bernhard A1 - Völzke, Holger A1 - Baden, M. T1 - German approach and experience feedback of transport ability 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 package design in accordance with the international transport regulations. The license for dry storage is granted on the German Atomic Energy Act with respect to the recently revised “Guidelines for dry interim storage of irradiated fuel assemblies and heat-generating radioactive waste in casks” by the German Waste management Commission (ESK). For transport on public routes between or after long term interim storage periods, it has to be ensured that the transport and storage casks fulfil the specifications of the transport approval or other sufficient properties which satisfy the proofs for the compliance of the safety objectives at that time. In recent years the validation period of transport approval certificates for manufactured, loaded and stored packages were discussed among authorities and applicants. A case dependant system of 3, 5 and 10 years was established. There are consequences for the safety cases in the Package Design Safety Report including evaluation of long term behavior of components and specific operating procedures of the package. Present research and knowledge concerning the long term behavior of transport and storage cask components have to be consulted as well as experiences from interim cask storage operations. Challenges in the safety assessment are e.g. the behavior of aged metal and elastomeric gaskets under IAEA test conditions to ensure that the results of drop tests can be transferred to the compliance of the safety objectives at the time of transport after the interim storage period. T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 KW - Storage KW - Transport KW - Lagerung KW - Verpackung KW - Radioactive material PY - 2013 SP - 1 EP - 9 PB - Omnipress AN - OPUS4-30159 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Schilling, O. A1 - Müller, Lars A1 - Ballheimer, Viktor A1 - Komann, Steffen A1 - Wille, Frank T1 - Analytical approaches within the assessment strategy of numerical analyses of components of RAM packages N2 - 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. T2 - PATRAM 2013 - 17th International symposium on the packaging and transportation of radioactive materials CY - San Francisco, CA, USA DA - 18.08.2013 PY - 2013 SP - Paper 217, 1 EP - 9 PB - Omnipress AN - OPUS4-31048 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin A1 - Nehrig, Marko A1 - Bletzer, Claus Wilhelm A1 - Musolff, André A1 - Wille, Frank T1 - Combustion of wood encapsulated in steel sheets during fire test N2 - Packages for the transport of radioactive material are often equipped with impact limiters consisting of wood, encapsulated by Steel sheets. These impact limiters shall ensure the transport cask meets the IAEA safety requirements. After damage caused by the mechanical tests the package has to withstand a severe fire scenario. According to the regulations during and following the fire test, the specimen shall not be artificially cooled and any combustion of materials of the package shall be permitted to proceed naturally. Due to results of the French institute IRSN it was shown that after the fire test an additional energy supply from a pre-damaged impact limiter should be taken into account. The combustion or smouldering of wood was of interest. BAM started a first test phase to examine the issue of combustion for such kind of package components. The goal was to understand the phenomena under the consideration of relevant regulatory boundary conditions. Several metal buckets were filled with wood and equipped with thermocouples. The test specimens have been prepared with different damage arrangements to take the influence of the mechanical tests into account. This paper shows the experimental Setup and the conduction of the tests. The first test shows that pre-damaged metal encapsulations can lead to smouldering of the wood and with this to a Supplement energy release after the end of the 30 minute fire. The consequence could be, to consider additional thermal loads of wood filled impact limiters to filled the IAEA regulations, if the conditions of the tests are transferable to the safety analysis of the package design. T2 - RAMTRANS 2015 - 10th International conference on radioactive materials transport and storage CY - Oxford, UK DA - 19.05.2015 KW - Typ-B KW - Impact limiter KW - Shock absorber KW - Wood KW - Thermal test KW - Fire test PY - 2015 SP - 1 EP - 9 AN - OPUS4-33407 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Ballheimer, Viktor A1 - Quercetti, Thomas A1 - Sterthaus, Jens T1 - Consideration of gaps between content and lid within package design assessment N2 - Type B(U) packages for the transport of radioactive material have to withstand accident conditions of transport defined in the regulations of the International Atomic Energy Agency in form of different mechanical (drop) tests with a subsequent thermal test. According to the regulatory requirements the orientation of the package in drop tests shall be such to cause the most damaged state in the components performing the safety functions. For the package Iid system a 9 m drop onto the unyielding target with Iid side downwards is often the most damaging orientation. The impact Ioads acting on the Iid in this orientation result mainly from interaction between Iid and internal content. In case of a movable content its impact onto the inner side of the Iid can cause additional Ioad peaks on the Iid and the Iid bolts. The intensity of the internal collision depends on the position of content relating to Iid at the time of package first contact with target. Due to physical limitations an axial gap, which could be set in "pre-drop" configuration of package or which could spontaneously appear during the drop test, usually does not cover the maximum size possible in specific package design. In this context, the combination of drop tests with post-test analysis can be helpful to better estimate the effect of internal impact. The paper summarizes some aspects of this issue based on the BAM experience in the design assessment of Type B(U) transport packages. Additionally the paper shall support applicants in German approval procedures to reduce rounds of questions and ensure delivery of reliable safety case documents to the authorities. T2 - RAMTRANS 2015 - 10th International conference on radioactive materials transport and storage CY - Oxford, UK DA - 19.05.2015 KW - Transport packages KW - Accident conditions of transport KW - Internal impacts KW - Lid/content interaction KW - Radioactive material PY - 2015 SP - 1 EP - 10 AN - OPUS4-33419 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 - Müller, Lars A1 - Scheidemann, Robert A1 - Schönfelder, Thorsten A1 - Komann, Steffen A1 - Wille, Frank T1 - Drop tests assessment of internal shock absorbers for packages loaded with encapsulations for damaged spent nuclear fuel N2 - Damaged spent nuclear fuel (DSNF) can be loaded in German dual-purpose casks (DPC) for transport and interim storage. Encapsulations are needed to guarantee a safe handling and a tight closure, separated from the package enclosure. These encapsulations shall be durable and leak-tight for a long storage period, because they are usually not accessible within periodical inspections of the DPC. Due to the general design of DPCs for standard fuel assemblies, specific requirements have to be considered for the design of encapsulations for DSNF to ensure the loading in existing package designs. Especially the primary lid system of a DPC is designed for maximum loads due to the internal impact of the content during drop test conditions. The main difference of encapsulations for damaged spent nuclear fuel is that they have usually a much higher stiffness than standard fuel assemblies. Therefore the design of an internal shock absorber, e.g. at the head of an encapsulation is required to reduce mechanical loads to the primary lid system during impacts. BAM as part of the German competent authority system is responsible for the safety assessment of the mechanical and thermal package design, the release of radioactive material and the quality assurance of package manufacturing and operation. Concerning the mechanical design of the encapsulation BAM was involved in the comprehensive assessment procedure during the package design approval process. An internal shock absorber was developed by the package designer with numerical analyses and experimental drop tests. Experimental drop tests are needed to cover limiting parameters regarding, e.g. temperature and wall thickness of the shock absorbing element to enable a detailed specification of the whole load-deformation behavior of the encapsulation shock absorber. The paper gives an overview of the assessment work by BAM and points out the main findings which are relevant for an acceptable design of internal shock absorbers. The physical drop tests were planned on the basis of pre-investigations of the applicant concerning shape, dimension and material properties. In advance of the final drop tests the possible internal impact behavior had to be analyzed and the setup of the test facility had to be validated. The planning, performance and evaluation of the final drop tests were witnessed and assessed by BAM. In conclusion it could be approved that the German encapsulation system for damaged spent nuclear fuel with shock absorbing components can be handled similar to standard fuel assemblies in existing package designs. T2 - Pressure Vessels & Piping Conference 2020 CY - Online meeting DA - 03.08.2020 KW - Encapsulations for damaged spent nuclear fuel KW - Drop tests KW - Internal shock absorber KW - Design assessment of RAM packages PY - 2020 SP - 1 EP - 9 AN - OPUS4-51546 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin A1 - Quercetti, Thomas A1 - Wille, Frank T1 - Outcomes of three large scale fire reference tests conducted in BAM fire test facility N2 - Packages for the transport of high-level radioactive material are designed to withstand severe accidents. Hypothetical severe accident conditions are defined in the IAEA Regulations for the Safe Transport of Radioactive Materials. One of these accident conditions is the thermal test, mainly consisting of a 30 minute fully engulfing 800°C pool fire or an equally severe fire test. The heat fluxes into the package depend substantially on the fire characteristics and the Surface temperature of the package. Fire tests can be performed at BAM on a propane gas fire test facility. In order to investigate the heat fluxes over a wide range of surface temperatures in this test facility a fire reference package was designed for multiple use. The package represented the outer geometry of a specific transport cask for radioactive waste. The fire reference package is a closed steel sheet cylinder with a wall thickness of 10 mm, a length of 1500 mm and a diameter of 1050 mm. The package was instrumented with thermocouples and filled with heat resistant insulation material. Three open-air fire tests were performed in the BAM propane gas fire test facility. The flames exposure time period varied slightly for the fire tests. The wind direction as well as the wind Speed were measured and changed between and during the tests. Test stand parameters such as wind shield location and propane gas volume flow were chosen constant for the three tests. The locally measured fire reference package steel sheet temperatures were used for the calculation of heat fluxes as function of time and surface temperature. The measured temperatures allowed further calculations. In a first approach effective fire characteristics of the propane gas fire, including the flame temperature, the fire convection coefficient and a Radiation exchange coefficient mathematically describing the determined average heat flux over the surface temperature were calculated. T2 - Pressure Vessels & Piping Conference 2020 CY - Online meeting DA - 03.08.2020 KW - Fire KW - Testing KW - Large scale testing KW - Calorimeter KW - Heat flux PY - 2020 SP - 1 EP - 9 PB - ASME CY - New York AN - OPUS4-51192 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Nehrig, Marko A1 - Scheidemann, Robert A1 - Wille, Frank A1 - Ballheimer, Viktor T1 - Investigation of the internal impact during a 9 m drop test of an accident-safe waste package N2 - The safety assessment of packages for the transport of radioactive material follows the IAEA regulations and guidance. The specified regulatory tests cover severe accidents and demonstrate the package containment system integrity. Special attention must be drawn to the behaviour of the content which could move inside the package due to unpreventable gaps caused by the loading procedure and the structure of the content. A possible internal impact of the content which occurs during the drop tests onto the lid system is investigated. The IAEA regulations SSR-6 and the Guidance SSG-26, revised recently, consider input from Germany and France related to the significance of internal gaps. In the context of a waste package design assessment, a model was equipped with a representative content to conduct a drop test with an internal impact. The weight and kinetic impact of this content covered all possible real contents. The objective of the test was to maximize the load onto the lid system and to prove the mechanical integrity by complying with the required leak tightness. The test was conducted conservatively at a package temperature lower than -40 °C at the BAM Test Site Technical Safety. This paper gives an overview of efforts to address internal gaps and their consequences, and the BAM efforts with the implementation of this topic into IAEA regulations and guidance material. The paper then focuses on the conduction of a drop test and investigation of internal component impact. T2 - ASME 2021 Pressure Vessels & Piping Conference (PVP2021) CY - Online meeting DA - 13.07.2021 KW - Internal gaps KW - Drop test KW - IAEA PY - 2021 SN - 978-0-7918-8535-2 DO - https://doi.org/10.1115/PVP2021-60996 SP - 1 EP - 6 PB - American Society of Mechanical Engineers (ASME) CY - New York, NY, USA AN - OPUS4-54742 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, 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 - CONF A1 - Rolle, Annette A1 - Komann, Steffen A1 - Wille, Frank T1 - Ageing aspect in design approval of special form radioactive material N2 - 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. T2 - PATRAM22 Conference CY - Juan-Les-Pins, Antibes, France DA - 11.06.2023 KW - Ttransport KW - Radioactive material KW - Sealed sources KW - Ageing PY - 2023 SP - 1 EP - 6 AN - OPUS4-57786 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Wille, Frank A1 - Quercetti, Thomas A1 - Scheidemann, Robert T1 - Test Facilities for Radioactive Materials Transport and Storage Packagings at BAM N2 - BAM acts as authority and for service in safety assessment of packages for transport and storage of radioactive materials. We offer extensive test capabilities and application of analytical methods for design verification and simulation for all types of packages for the transport and storage of radioactive materials according with the international IAEA Regulations for the safe transport and for national storage acceptance criteria. BAM operates several test facilities for drop and stacking testing, leak testing and thermal testing. The large drop test tower allows dropping full-scale specimens up to 200,000 kg in any drop orientation as requested. The comprehensive test facilities combined with long-term experience, newest equipment and measurement devices according to the latest state-of-the-art technology ensures realisation of complex test campaigns for package safety evaluation. Beyond that, non-destructive and destructive material test devices and experts are available. Equipment and application of all kinds of typical measurement categories can be offered for testing campaigns. In recent years we performed testing of full-scale type B package models with complex handling and preparation procedures. The results were contributed for different package design approval procedures. Type A packages mainly designed for medical related transport purposes, were continuously tested according to the transport regulations over recent years as well. Moreover, we work on research topics with relevance to package safety. The mechanical behaviour of lid closure systems under transport and storage conditions and the thermal behaviour of impact limiters were recently of special importance for the assessment competencies of BAM and were investigated under use of our test facilities. The paper describes the test facilities and capabilities for package design safety evaluation at BAM and shows examples from our recent work. T2 - PATRAM 2022, 20th International Symposium on the Packaging and Transportation of Radioactive Materials CY - Juan-les-Pins, France DA - 11.06.2023 KW - IAEA KW - Fire test KW - Drop testing KW - Transport KW - Package PY - 2023 SP - 1 EP - 12 AN - OPUS4-57967 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin A1 - Quercetti, Thomas A1 - Gleim, Tobias A1 - Nehrig, Marko A1 - Musolff, André A1 - Wille, Frank T1 - R&D Activities by BAM Related to Transport Package Fire Testing N2 - Packages for the transport of radioactive material shall meet the mechanical and thermal test requirements of the International Atomic Energy Agency (IAEA) regulations for package design approval. Besides mechanical testing, the Federal Institute for Materials Research and Testing (BAM) performs thermal tests in accordance with the IAEA regulations. The thermal test includes a 30-minute 800°C fully engulfing fire. BAM continuously performs various thermal experiments for the investigation of the thermal response of packages with respect to the IAEA fire. The purpose of this paper is to give an overview of the already performed, ongoing and future physical tests and experiments of BAM in the field of thermal investigations. These research and development works shall support our competencies for the authority package design assessment. BAM operates a propane gas fire test facility. To be able to carry out comparative investigations and validity between the propane fire and the in detail prescribed pool fire test in the regulations, BAM carries out various calorimetric tests and investigates the boundary conditions of the fire with the help of fire reference packages. At the same time, we are conducting various fire scenarios with wood-filled impact limiters. Large-scale fire tests of impact limiters are carried out on a full scale as well as on a small scale. Influencing variables are investigated in particular by means of geometric changes and the consideration of artificial damages, in particular holes. In addition to propane fire as a heat source, thermal scenarios are also investigated with hydrogen as heat source and an infrared radiator system to ignite test specimens. For these numerous test arrangements, the transferability to existing and newly developed transport package designs is essential and fruitful within the review of design approvals, especially for Dual Purpose casks with a long-lasting operation time. T2 - 20th International Symposium on the Packaging and Transportation of Radioactive Materials (PATRAM 22) CY - Juan-Les-Pins, France DA - 11.06.2023 KW - Fire KW - Testing KW - Hydrogen KW - Wood KW - Propane KW - Heat Flux KW - Fire Reference Package KW - Radioactive Material PY - 2023 SP - 1 EP - 10 AN - OPUS4-57721 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Gleim, Tobias A1 - Neumann, Martin A1 - Linnemann, Konrad A1 - Wille, Frank T1 - Welding Seam Safety Evaluation in a Thick-Walled Type B Transport Package N2 - The safety demonstration of heavy weight type B transport packages used for storing spent nuclear fuel (SNF) or radioactive waste is ensured by a combination of physical testing and numerical calculations. While experiments are performed in accordance with the IAEA regulations for selected drop scenarios, Finite-Element-Method (FEM) simulations are used to predict the most damaging case and to investigate additional drop positions. BAM as competent authority in Germany has performed different investigations of a welding seam for a typical large transport package made of A508 forged steel, where the bottom plate is welded to the cylindrical shell. The package has a mass of approx. 120 t. Results of physical drop tests with a full-scale model and accompanying preliminary FEM simulations are presented to determine the decisive stresses in the welding seam. A drop test only represents one set of a package and test parameters. A further parameter analysis is considered to account for allowable variations of packaging properties (e.g. resulting from the manufacturing process) and, based on IAEA requirements, the temperature dependence of the material behaviour. The results of the stress analyses from the drop test and the simulation form the basis and provide the input parameters for a fracture mechanics analysis. In addition to the IAEA specifications, further standards are taken into account for an in-depth investigation, see R6 [1], BS 7910 [2] and API 579-1/ASME FFS1 [3]. All the above-mentioned standards require a manufacturer-specific defect analysis with respect to size and position. Both result from the welding process and the following heat treatment regime. The maximum defect sizes are ensured with non-destructive test methods (such as ultrasonic or particle methods) as integral part of the manufacturing process of the welding seam. Another important parameter in the welding process is the residual stress (secondary stress). The combination of the primary and secondary stress determines the total stress in the welding seam. The most damaging case of the welding seam is determined and evaluated with help of the abovementioned standards and taking into account the IAEA requirements with respect to defect sizes, material properties, primary and residual stress, yield strength etc. T2 - PATRAM 22 - The International Symposium on the Packaging and Transportation of Radioactive Materials CY - Juan-Les-Pins, Antibes DA - 11.06.2023 KW - Welding KW - Transport Package KW - Fracture Mechanics PY - 2023 SP - 1 EP - 11 AN - OPUS4-59421 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Feldkamp, Martin A1 - Quercetti, Thomas A1 - Gleim, Tobias A1 - Wille, Frank T1 - Evaluation of Heat Fluxes in Fire Reference Test Conducted in BAM Propane Gas Fire Test Facility N2 - Packages for the transport of intermediate- and high-level radioactive waste are designed to withstand severe accidents. The International Atomic Energy Agency (IAEA) has established specific mechanical and thermal tests. Packages for the transport of radioactive material must withstand these tests to comply with the Regulations for the Safe Transport of Radioactive Materials IAEA [IAEA (2018)]. A fire reference package was developed with the primary objective to demonstrate that the fire meets the regulatory requirements. Another aim is to characterise the boundary conditions of the actual fire as input parameters for thermo-mechanical simulations. A simple method to characterise the boundary conditions of a real steady state fire with a fire reference package is presented. The thermal test mainly consists of a 30 minute fully engulfing 800°C pool fire or an equally severe fire, such as a propane gas fire. The fire reference tests are performed prior to the actual fire test with the real package. The heat fluxes into the package depend substantially on the fire characteristics and the surface temperature of the package. To investigate local and overall heat fluxes over a wide range of surface temperatures in this test facility a fire reference package was designed for repeated use. The fire reference package presented in this paper represents the outer geometry of a small transport container for radioactive material and is used as a device in civil engineering. It is designed as a closed steel sheet cylinder with a wall thickness of 10 mm, a length of 182 mm and a diameter of 102 mm. The package was instrumented with thermocouples and filled with heat resistant insulation material. Open-air fire tests were performed in a BAM propane gas fire test facility with the fire reference package. The measured temperatures are used to determine the changes of heat fluxes into the fire reference package in relation to the package surface temperature. The calculated heat fluxes allow its fitting to express the thermal exposure as simple mathematical boundary condition. Therefore, in a first approach, fire properties such as adiabatic surface temperature (AST) as proposed by Wickström et al. (2007), convection coefficient and emissivity are determined mathematically fitting the heat flux development presented in this paper. The evaluated results provide an initial picture of local fire characteristics of the conducted propane gas fire and are a further development of previous works from Feldkamp et al. (2020). The results can be used in thermal and thermo-mechanical models to simulate the load on the real transport package in fire. The test shows that the examined propane gas fire covers the IAEA-fire over a wide range of surface temperatures with the chosen test stand parameters. T2 - SMiRT 27 (27th conference on Structural Mechanics in Reactor Technology) CY - Yokohama, Japan DA - 03.03.2024 KW - Fire KW - Propane KW - Heat Flux KW - Fire Reference PY - 2024 SP - 1 EP - 10 PB - IASMiRT AN - OPUS4-59679 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Rolle, Annette A1 - Wille, Frank ED - Czarwinski, R. T1 - Radioaktive Stoffe in besonderer Form-wichtige Aspekte im Zulassungsverfahren N2 - Für die Beförderung von umschlossenen radioaktiven Stoffen (Strahler) können Transporterleichterungen gelten, wenn das Strahlerdesign nachweislich unfallsicher ausgelegt ist und eine Zulassung als radioaktiver Stoff in besonderer Form vorliegt. Die in Deutschland zuständige Behörde für die Prüfung und Zulassung radioaktiver Stoffe in besonderer Form ist die Bundesanstalt für Materialforschung und –prüfung (BAM). Eine oft langfristige Nutzung der Strahler kann eine alterungsbedingte Schwächung der Umschließung des radioaktiven Stoffes zur Folge haben. Jeder Strahler einer zugelassenen Bauart muss jedoch im Beförderungsfall zu jeder Zeit, auch nach längerer Nutzung, den vorgeschriebenen Prüfanforderungen genügen. Die Alterungsbewertung auf der Grundlage einer vom Antragsteller zu spezifizierenden Nutzungsdauer des radioaktiven Stoffes in besonderer Form ist seit langer Zeit Bestandteil des Zulassungsverfahrens in Deutschland. Ein von Deutschland eingebrachter Vorschlag im Rahmen der anstehenden Revision der IAEA Empfehlungen für die sichere Beförderung von radioaktiven Stoffen (SSR-6) soll die Lesbarkeit des Regelwerkes bezüglich der Alterungsbewertung verbessern, Klarheit über die damit verbundenen Anforderungen erzeugen und damit zu einer Harmonisierung der Zulassungsverfahren auf internationaler Ebene beitragen. In diesem Beitrag sollen wichtige Einflussfaktoren auf das Alterungsverhalten von radioaktiven Stoffen in besonderer Form aufgezeigt und die Notwendigkeit für die Spezifikation einer Nutzungsdauer als Grundlage für die Alterungsbewertung begründet werden. Der deutsche Vorschlag für die Revision des IAEA Regelwerks zum sicheren Transport radioaktiver Stoffe (SSR-6) wird vorgestellt und erläutert. T2 - Jahrestagung des FS 2022 CY - Konstanz, Germany DA - 26.09.2022 KW - Strahlenschutz KW - Radioaktive Stoffe KW - Beförderung KW - Zulassung PY - 2022 SN - 1013-4506 SP - 15 EP - 20 PB - Fachverband für Strahlenschutz e.V. CY - Berlin AN - OPUS4-56134 LA - deu AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Quercetti, Thomas A1 - Feldkamp, Martin A1 - Gleim, Tobias A1 - Musolff, Andre A1 - Werner, Jan A1 - Wille, Frank T1 - Enhancement of the BAM Fire Test Stand for Testing a Large Transport Package for Radioactive Materials N2 - Packages for the transport of spent nuclear fuel are designed to endure severe accidents. To obtain approval, these transport packages must adhere to the specification-based criteria of the international transport regulations SSR-6 of the International Atomic Energy Agency (IAEA). To ensure compliance with these requirements, specific mechanical and thermal tests need to be addressed with respect to the package type. Typically, SSR-6 prescribes a mechanical test followed by a thermal fire test as part of different testing scenarios. To approve the latter test of the sequence, BAM performs calorimeter tests in advance with so-called fire reference packages for characterizing the actual fire and its impact on the package to be tested. Packages are designed with different geometry sizes depending on their purpose. For previous tests, the implemented test setups in the fire test stand were sufficiently dimensioned and could cover all requirements in this respect. However, to cover additional testing needs in the future, BAM is expanding the test setup for the purpose of testing significantly larger packages. In previous test setups one ring burner for propane surrounding the test specimen was sufficient. The limiting size in this configuration was the design height of the transport package to be tested. According to the thermal test of the IAEA-Regulations [1], a 30-minute fully engulfing 800°C pool fire or an equally severe fire, e.g., propane gas fire, must be applied to the transport package. Possible adjustments such as nozzle cross-section and propane mass flow can be adjusted to a certain extent. Further modifications, to cover significantly higher and larger packages all-around with a fully engulfing fire, must be accomplished with an additional, second burner ring. Both burner rings had the same dimensions and were mounted on top of each other at different heights to create a significantly larger volume of fire that completely engulfs the package including its impact limiter. To meet the IAEA regulatory boundary conditions, the enhanced fire test stand with the second burner ring is tested with a large fire reference package and will then be used for real-size transport packages after all parameters are successfully met. This fire reference package represents the external geometry of a generic transport cask for radioactive material and is equipped with numerous temperature sensors to record temperature curves at the interior wall surfaces. T2 - RAMTrans 2024 CY - London, United Kingdom DA - 14.05.2024 KW - Fire testing KW - Full-scale KW - Transport package PY - 2024 SP - 1 EP - 7 AN - OPUS4-60253 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER - TY - CONF A1 - Apel, Andreas A1 - Wille, Frank A1 - Moutarde, M. A1 - Sert, G. A1 - Caillard-Lizot, M.-T. A1 - Eckert, B. A1 - Konnai, A. A1 - Pstrak, D. A1 - Harvey, J. A1 - Desnoyers, B. A1 - Purcell, P. A1 - Fiaccabrino, V. T1 - Stowage during transport: a proposal for harmonization N2 - For stowage and retention during transport, the International Atomic EnergyAgency (IAEA) Transport regulations (SSR-6) only require that the package shall be securely stowed. Additional information is given in the IAEA guidance material (SSG-26), that includes specific load factors that should be considered in the safety demonstrations. Nevertheless, applicants commonly use other load factors to justify the design of the packaging attachment points. In particular, acceleration values vary between different countries and this may lead to difficulties during the validation of foreign approval certificates. Discussions with applicants identified that the load factors presented in the IAEA guidance material are not necessarily relevant for the different transport modes. For some modes, the load factors are representative of situations occurring in routine conditions of transport. For other modes, the load factors are more representative of situations occurring in normal or accident conditions of transport. Furthermore, the origins of the specified values are not clear. In this context, some IAEA TRANSSC members decided to review the stowage guidance. An international working group was constituted in 2013, including representatives of competent authorities, technical support organizations, and transport stakeholders. Several topics were discussed and many questions were raised during the two year review. For example, some discussions focused on Paper No.1031 2 the conditions which have to be considered for stowage design, both as relevant to the load factors used for strength and fatigue analysis, as well as the criteria which have to be considered for the package attachment points. In addition, related questions on operational aspects were also discussed. Overall, the working group concluded that new guidance material was warranted as input into SSG-26. As a result of those discussions, the international working group produced new guidance material for stowage in transport, addressing each of these topics. The proposal to modify the IAEA guidance material (SSG-26, Appendix IV), presented by France to the TRANSSC in the 2015 initiated Review cycle of the Regulations, was unanimously accepted and will be implemented in the next edition of the guidance material. T2 - PATRAM 2016 CY - Kobe, Japan DA - 18.09.2016 KW - Packages for radioactive material KW - IAEA regulations (SSR-6, SSG-26) KW - Tie-down and stowage KW - Load cases for transport KW - Design of the packaging attachment points PY - 2016 SP - Paper 1031, 1 EP - 8 AN - OPUS4-38644 LA - eng AD - Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin, Germany ER -