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This presentation addresses safety aspects concerning dual purpose casks for transportation and storage of spent fuel and high level radioactive waste. The long term performance of casks and their safety relevant components like sealed lid systems and the long term performance of cask internals, especially spent fuel assemblies are discussed. Transportation after interim storage is another key issue and due to the fact of delayed disposal projects current interim storage periods need to be extended causing additional safety demonstration needs for longer periods of time. Regarding this situation present BAM research activities in this area and their preliminary outcomes are presented.
The presentation explains the requirements, process and challenges of the container design testing procedures for the Konrad repository, where low and intermediate level radioactive waste is planned to be disposed of beginnung end of 2022. BAM performs container design testing for more than 30 years whether analytically or experimentally. The challenges of the complex Konrad requirements concerning their proper interpretation and application is discussed in this presentation.
This presentation starts with a brief overview about the current situation of spent nuclear fuel and high level waste storage in Germany followed by a discussion of the expected timeline towards final disposal. Furthermore an overview is presented about research programmes for radioactive waste disposal and also extended interim storage issues due to the delay of the national repository siting procedure. In the second part of the presentation BAM research activities related to extended interim storage are addressed. This includes the status and preliminary results of proactive R&D projects concerning metal seals, degradation effects of polymers used as neutron shielding materials and finally elastomer seals.
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.
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.
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.
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.
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.
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.