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Packages for the transport of radioactive materials shall fulfil the requirements of the IAEA regulations for the safe transport. The requirements define mechanical and thermal test conditions including criteria ensuring the package design’s ability to withstand severe accidents and provide a high level of technical safety. Different methods can be used for safety demonstration showing compliance with the regulations.
The central part of a safety demonstration which is presented in this paper was a comprehensive drop test program with a full-scale model of a transport package accompanied by pre- and post-test FE analyses. Using full-scale drop test models allow the benefit that similarity and scaling issues become a significant smaller issue, additional material investigations can be limited and analyses for transferring test results to the original package design are reduced. Additionally, experience for the future serial packaging manufacturing and handling procedures can be collected in a very early state of the design approval process. The pre-test finite element analyses derived and justified the drop test program consisting of several drop sequences with different drop orientations of the specimen. The performance and the results of the drop test sequences shows the manageability and the advantage e.g., in view of the direct availability of test results for the package licensing. On the other hand, the drop test performance shows the difficulties during handling and the need for additional equipment during preparation of the specimen.
The package presented was intended for the transport and storage of compacted radioactive waste from reprocessing of spent nuclear fuel assemblies - designed and applied for approval by the AGC consortium. The project ended in 2021. The package design was characterized by a cask body made of a forged thick stainless-steel shell, a bolted double lid system with metallic gaskets and wood filled shock absorbers at both ends. The total mass of the entire transport package including content was 120,000 kg, the total length was about 7000 mm and the diameter approximately
3000 mm, both measures include the shock absorbers.
The paper provides an insight into the performance of a full-scale drop testing campaign within the package safety evaluation and shows some selected test results.
Characterization is a very important step in dealing with materials and waste streams generated during the operational and decommissioning phases of nuclear installations, including nuclear power plants. Characterization allows differentiation between materials that can be released from regulatory control and those that require further treatment and conditioning to become a stable waste form suitable for future storage and final disposal according to its waste classification. Characterization is also needed in the pre-disposal stages of radioactive waste management to demonstrate compliance with the waste acceptance criteria of the facilities that will accept the different waste forms. This work will present the strategies developed and implemented by the three projects for in-depth and accurate waste characterization and investigation of the different radioactive waste packages considered. CHANCE, MICADO, and PREDIS will present their goals, the methods developed, the technologies used and the (preliminary) results contributing to the improvement of the safety and the data and information quality of the waste packages analyzed at the different stages of the waste management process. Special emphasis will also be given to complementary approaches highlighting the usability of the technologies, the accessibility of the data, and the problem-solving of the three projects within the European panorama.
Suitable containers for receiving high-level radioactive waste are a key elements of every repository system concerning different host rocks like rock salt, claystone and crystalline rock as being considered in Germany. The disposal container design is decisive regarding boundary conditions for the transport, emplacement techniques, and the operational and long-term safety assessment. The systematic derivation of the requirements to be placed on disposal containers forms the basis for a targeted, comprehensible and transparent development of containers that meet the requirements. The presentation summarizes the outcomes of the research project KoBrA.
Before the final disposal of radioactive wastes, various processes can be implemented to optimise the waste form. This can include different chemical and physical treatments, such as thermal treatment for waste reduction, waste conditioning for homogenisation and waste immobilisation for stabilisation prior to packaging and interim storage. Ensuring the durability and safety of the waste matrices and packages through performance and condition assessment is important for waste owners, waste management organisations, regulators and wider stakeholder communities. Technical achievements and lessons learned from the THERAMIN and PREDIS projects focused on low- and intermediate-level waste handling is shared here. The recently completed project on Thermal Treatment for Radioactive Waste Minimization and Hazard Reduction (THERAMIN) made advances in demonstrating the feasibility of different thermal treatment techniques to reduce volume and immobilise different streams of radioactive waste (LILW) prior to disposal. The Pre-Disposal Management of Radioactive Waste (PREDIS) project addresses innovations in the treatment of metallic materials, liquid organic waste and solid organic waste, which can result from nuclear power plant operation, decommissioning and other industrial processes. The project also addresses digitalisation solutions for improved safety and efficiency in handling and assessing cemented-waste packages in extended interim surface storage.
This presentation summarizes the opbjectives and the progress of the EURATOM project PREDIS (pre-disposal management of radioactive waste) and its workpackage 7 in particular. Focus is on the development of wireless sensors for monitoring radioactive waste packages, measuring radiation, temperature, pressure and humidity.
To provide better means for a safe and effective monitoring of cemented waste packages including prediction tools to assess the future integrity development during pre-disposal activities, several digital tools are evaluated and improved in the frame of the EC funded project PREDIS. Safety enhancement (e. g. less exposure of testing personnel) and cost effectiveness are part of the intended impact.
The work includes but is not limited to inspection methods such as muon imaging, wireless sensors integrated into waste packages as well as external package and facility monitoring such as remote fiber optical sensors. The sensors applied will go beyond radiation monitoring and include Proxy parameters important for long term integrity assessment (e. g. internal pressure). Sensors will also be made cost effective to allow the installation of much more sensors compared to current practice.
The measured data will be used in digital twins of the packages for specific simulations (geochemical, integrity) providing a prediction of future behavior. Machine Learning techniques trained by the characterization of older packages will help to connect the models to the actual data.
All data (measured and simulated) will be collected in a joint data base and connected to a decision framework to be used at actual facilities.
The presentation includes detailed information about the various tools under consideration, their connection and first results of our research.
As part of the evaluation of a package for the safe transport of radioactive waste the regulations of the IAEA International Atomic Energy Agency shall be fulfilled. The regulations define requirements for the package and specify mechanical and thermal test conditions. Different methods are allowed for the test performance to demonstrate compliance with the regulations. Next to calculational approaches and the use of models of an appropriate scale, the performance of full-scale testing with prototype packages respectively full-scale models is applied.
The use of full-scale models has several advantages within the complete safety assessment procedure for a transport package approval. Scaling and corresponding similarity questions don’t have to be considered, additional material investigations can be limited and analyses to transfer test results to the original package design are reduced in number and complexity. Additionally, experience for future serial design procedures can be built up during manufacturing and assembling of the test model.
BAM operates different drop and fire test facilities south of Berlin, Germany. BAM has started to perform a drop test campaign with a full-scale model of 120 metric tons weight for a transport package approval procedure. The paper describes experience with test preparation, drop performance and additional analyses. The measurement concept is explained and test goals regarding the package safety assessment and evaluation of safety margins are introduced.
The European Joint Programme on Radioactive Waste Management (EURAD) has been launched to coordinate activities on agreed priorities of common interest between European partners. The main actions include scientific and technical activities on radioactive waste characterization and processing, interim storage of radioactive waste as well as predisposal and disposal solutions. As part of the programme, the work package on Spent Fuel Characterization and Evolution until Disposal (SFC) was proposed and accepted by the European Commission. The objectives of this work package include the development of beyond-state-of-the-art characterization techniques for spent nuclear fuel during its evolution from reactor unloading until disposal, the understanding of the spent fuel performance during prolonged storage, transport and emplacement in a deep geological repository, the understanding of the behaviour of fuel, cladding, pellet-cladding interaction and ageing effect under normal and postulated accident scenarios until disposal, and contributions to operational safety concepts for fuel handling at spent nuclear fuel packaging facilities.
Ongoing R&D work at BAM
(2018)
The presentation provides an update on the German nuclear waste management policy towards final disposal and related reasearch at BAM. Topics include long term investigations of metal seals, investigations of fuel cladding failure mechanisms, and the development of requirements and concepts for high level radioactive waste and spent fuel disposal containers.
The Bundesanstalt für Materialforschung und -prüfung (BAM) in Germany has gathered extensive expertise over decades in the fields of design testing and safety assessment of transport and storage containers for radioactive materials. On behalf of the German commission for the disposal of highly radioactive waste, BAM had been assigned the elucidation of requirements for high level waste disposal containers specifically in terms of their retrievability from deep geological formations during repository operation and recovery after its closure over a period of time between 500 to 1000 years. The output of the initial assessment is presented taking into account the frame of the current German repository site selection procedure together with the definition of repository conditions corresponding to the potential deep geological formations to be considered for disposal.
This presentation addresses the spectrum of relevant stress factors for disposal containers and reflects their relevance and consequences during the repository operation phase including the emplacement and their potential retrieval. In addition, further aspects and consequences for container recovery after repository closure are discussed taking into consideration potential repository conditions of different geological formations.
Finally, potential disposal container concepts are briefly addressed comparing the consequences of an improvement of existing dual purpose casks versus completely new disposal canister designs adjusted to the specific geological conditions of the host rock formation. Pros and cons of these two concepts are depicted in order to address major challenges. However, precise container design concepts cannot be consolidated before repository conditions and derived specific container requirements are designated.