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The Large Aperture Ultrasound System (LAUS) developed at BAM is known for its ability to penetrate thick objects, especially concrete structures commonly used in nuclear waste storage and other applications in civil engineering. Although the current system effectively penetrates up to ~9 m, further optimization is imperative to enhance the safety and integrity of disposal structures for radioactive or toxic waste. This study focuses on enhancing the system’s efficiency by optimizing the transducer spacing, ensuring that resolution is not compromised. An array of twelve horizontal shear wave transducers was used to find a balance between penetration depth and resolution. Systematic adjustments of the spacing between transmitter and receiver units were undertaken based on target depth ranges of known reflectors at depth ranges from 5 m to 10 m. The trade-offs between resolution and artifact generation were meticulously assessed. This comprehensive study employs a dual approach using both simulations and measurements to investigate the performance of transducer units spaced at 10 cm, 20 cm, 30 cm, and 40 cm. We found that for depths up to 5 m, a spacing of 10 cm for LAUS transducer units provided the best resolution as confirmed by both simulations and measurements. This optimal distance is particularly effective in achieving clear reflections and a satisfactory signal-to-noise ratio (SNR) in imaging scenarios with materials such as thick concrete structures. However, when targeting depths greater than 10 m, we recommend increasing the distance between the transducers to 20 cm. This increased spacing improves the SNR in comparison to other spacings, as seen in the simulation of a 10 m deep backwall. Our results emphasize the critical role of transducer spacing in achieving the desired SNR and resolution, especially in the context of depth imaging requirements for LAUS applications. In addition to the transducer spacing, different distances between individual sets of measurement positions were tested. Overall, keeping the minimal possible distance between measurement position offsets provides the best imaging results at greater depths. The proposed optimizations for the LAUS in this study are primarily relevant to applications on massive nuclear structures for nuclear waste management. This research highlights the need for better LAUS efficiency in applications such as sealing structures, laying the foundation for future technological advances in this field.
Data science (digitalisation and artificial intelligence) became more than an important facilitator for many domains in fundamental and applied sciences as well as industry and is disrupting the way of research already to a large extent. Originally, data sciences were viewed to be well-suited, especially, for data-intensive applications such as image processing, pattern recognition, etc. In the recent past, particularly, data-driven and physics-inspired machine learning methods have been developed to an extent that they accelerate numerical simulations and became directly usable for applications related to the nuclear waste management cycle. In addition to process-based approaches for creating surrogate models, other disciplines such as virtual reality methods and high-performance computing are leveraging the potential of data sciences more and more. The present challenge is utilising the best models, input data and monitoring information to integrate multi-chemical-physical, coupled processes, multi-scale and probabilistic simulations in Digital Twins (DTw) able to mirror or predict the performance of its corresponding physical twins. Therefore, the main target of the Topical Collection is exploring how the development of DTw can benefit the development of safe, efficient solutions for the pre-disposal and disposal of radioactive waste. A particular challenge for DTw in radioactive waste management is the combination of concepts from geological modelling and underground construction which will be addressed by linking structural and multi-physics/chemistry process models to building or tunnel information models. As for technical systems, engineered structures a variety of DTw approaches already exist, the development of DTw concepts for geological systems poses a particular challenge when taking the complexities (structures and processes) and uncertainties at extremely varying time and spatial scales of subsurface environments into account.
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.
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.
Two hybrid alkaline cements (HAC) based on Portland clinker, ground granulated blast furnace slag (GGBFS), fly ash and sodium sulfate, as well as an alkali-activated GGBFS/fly ash blend and a Portland cement paste were exposed to a saturated saline solution for 70 days. The combined chemical attack of chloride, magnesium and sulfate ions and the associated changes of the phase assemblage of the materials were studied by X-ray diffraction, thermal analysis and spatially resolved X-ray fluorescence spectroscopy. The experimental results revealed dissolution of ettringite, C-N-A-S-H and calcite, and the formation of gypsum, Kuzel's salt and Friedel's salt; thermodynamic modeling indicated the formation of M-S-H. The resistance of the HAC against attack by the saline solution increased with Portland clinker fraction. The capacity of portlandite to maintain pH at values above 10 is found to be a major factor controlling the resistance of HAC against corrosion in the saline solution.
Recovery is an important measure for seal applications representing to which extent the elastomer regains its initial shape after deformation and release of an applied force. Compression set (CS) indicates the degree of recovery. Ethylene propylene diene rubber (EPDM) was aged at 75 ◦C, 100 ◦C, 125 ◦C and 150 ◦C for different ageing times up to five years and compression set measurements were performed at different times after disassembly and after additional tempering. Short- and long-term recovery up to one year after release for samples aged at 125 ◦C and 150 ◦C was also studied. To assess the curvature in the Arrhenius diagram that may occur due to non-sufficiently aged specimens, a degradation-rate based model was fitted to the CS data after tempering. For each ageing temperature, two decay fit functions were proposed, each with an activation energy and a corresponding degradative process. The influence of ageing on the leak-tightness after fast small partial release is investigated and estimated through the analysis of the shift factors from time temperature superposition (TTS) of CS measurements at different times after disassembly. Shift factors of CS measurement after 1 s and after additional tempering are in good agreement.
SFC is a work package in Eurad that investigates issues related to the properties of the spent nuclear fuel in the back-end of the nuclear fuel cycle. Decay heat, nuclide inventory, and fuel integrity (mechanical and otherwise), and not least the related uncertainties, are among the primary focal points of SFC. These have very significant importance for the safety and operational aspect of the back-end. One consequence is the operation economy of the back-end, where deeper understanding and quantification allow for significant optimization, meaning that significant parts of the costs can be reduced. In this paper, SFC is described, and examples of results are presented at about half-time of the work package, which will finish in 2024. The DisCo project started in 2017 and finished in November 2021 and was funded under the Horizon 2020 Euratom program. It investigated if the properties of modern fuel types, namely doped fuel, and MOX, cause any significant difference in the dissolution behavior of the fuel matrix compared with standard fuels. Spent nuclear fuel experiments were complemented with studies on model materials as well as the development of models describing the solid state, the dissolution process, and reactive transport in the near field. This research has improved the understanding of processes occurring at the interface between spent nuclear fuel and aqueous solution, such as redox reactions. Overall, the results show that from a long-term fuel matrix dissolution point of view, there is no significant difference between MOX fuel, Cr+Al-doped fuel, and standard fuels.
Wood is widely used in impact limiters of Transport casks for radioactive material. Encapsulated by an outer and inner Steel structure, spruce wood is offen applied in layers of altemating direction. The friction at the interfaces between these layers is of crucial importance for the impact and energy absorption e.g., at an accidental impact of a cask against a hard component.
In order to get detailed Information for corresponding numerical calculations, in this study the friction coefficient for the combinations wood-wood and wood-steel was measured in the temperature ränge between -40 °C and 90 °C according to the relevant stress conditions for such casks. Results show decreasing friction with increasing temperature, ranging from 0.43 at -40 °C to 0.22 for 90 °C for wood-steel combinations and from 0.3 at -40 °C to 0.24 at 90 °C to for a wood-wood combination.
The thermal expansion of polymeric parts can be an issue in many applications where the available space is limited, or exact dimensions of the part are required. For this study, a device was designed and built that allowed measuring the thermal expansion simultaneously in all three spatial directions on cubic samples with real-scale dimensions (78 mm edge length). The results are shown between 25 °C and 125 °C for two PE materials, one HMWPE and one tempered UHMWPE, for non-irradiated samples as well as cubes that have been irradiated with 100 and 400 kGy. The results measured with the new device were very similar to those measured with conventional thermo-mechanical analysis equipment and to literature data of UHMWPE. The HMWPE material shows a much larger thermal expansion coefficient in one direction compared to the other two directions during the first heating due to frozen stresses from the pressing step during material manufacturing. These stresses are mostly released by the expansion during the first heating, so that the expansion during the second heating is more uniform. The overall volumetric expansion is the same for both heating runs. By contrast, the tempered UHMWPE material shows no significant difference between first and second heating run, as the stresses from processing could already relax in the tempering step. The irradiation treatment does not affect the values significantly for the given test set-up.
CASTOR-Behälter sollen den Auswirkungen schwerster Unfälle standhalten und müssen dabei ihre Sicherheitsfunktionen beibehalten. Dazu zählen der sichere Einschluss des radioaktiven Inventars mit nachgewiesener Behälterintegrität und -dichtheit, die ausreichende Abschirmung der radioaktiven Strahlung, die Ableitung der von den Brennelementen ausgehenden Wärme und der Ausschluss des Entstehens einer nuklearen Kettenreaktion, d.h. die Kritikalitätssicherheit sind zu gewährleisten.
Zum Nachweis dieser Fähigkeiten werden die Behälter gegen Stoß, Aufprall, Durchstoßen, Feuer sowie beim Eintauchen in Wasser geprüft. Durch diese Tests werden Beanspruchungen aus potentiellen schweren Unfällen abdeckend simuliert. Die hierfür erforderlichen experimentellen Tests, aber auch die Prüfung und Entwicklung komplexer rechnerischer Simulationen des Behälterverhaltens unter Anwendung der Finite-Elemente-Methode erfolgen bei der Bundesanstalt für Materialforschung und -prüfung (BAM).
Die Basis bildet das Regelwerk der Internationalen Atom- und Energieorganisation (IAEO) zum sicheren Transport von radioaktiven Stoffen, welches international harmonisierte Schutzziele und Prüfanforderungen definiert. Die seit 60 Jahren praktizierte stetige Weiterentwicklung der Sicherheitsanforderungen trägt aktuellen Erkenntnissen, Erfahrungen und Risikoanalysen Rechnung. Die Brennelementbehälter werden mechanischen Prüfungen sowie Brand- und Wasserdruckprüfungen unterzogen. Hierdurch wird nachgewiesen, dass sie den Auswirkungen eines schweren Unfalls standhalten können.
Nur Behälter die zweifelsfrei die kumulativen mechanischen und thermischen Tests und auch die Wasserdruckprüfung mit Erfüllen aller Schutzziele bestehen, bekommen eine Zulassung, die es erlaubt die Behälter zu transportieren.