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A sealed container for the geological disposal of spent nuclear fuel and vitrified high-level waste is the only component of a deep geological repository that provides complete containment of radionuclides. As such, attention is focused on its lifetime. The lifetime of the container is influenced by material degradation processes during disposal and is typically of the order of several millennia and, for some container materials, up to one million years. Designing, manufacturing, and predicting the performance of containers over such long periods requires an in-depth understanding of their material properties, fabrication processes, and degradation mechanisms. Scientific and technological progress can improve both the performance of containers and the robustness of lifetime predictions. Optimization of these aspects is of primary importance for many national radioactive waste disposal programs. In this article, the state of the art of complex coupled degradation processes, as well as the optimization potential of novel container materials, is presented. Furthermore, the existing tools allowing the prediction of long-term barrier integrity are discussed.

Nuclear power plays a major role in the generation of electricity with low carbon emissions. However, it generates significant amounts of radioactive waste, mainly from contaminated metallic components such as steam generators. Decontamination is essential for the safe handling and eventual recycling or disposal of these materials. Various decontamination techniques can be utilized but chemical processes are recommended for complex geometries such as the tubular parts of steam generators. COREMIX (Chemical Oxidation REduction using nitric permanganate and oxalic acid MIXture) is a process that is similar to the CORD (Chemical Oxidation Reduction Decontamination) process currently utilized in the industry which involves dissolving the contaminated oxide layers from metallic surfaces. This process generates a large quantity of radioactive effluent that requires appropriate treatment. The objective is to reduce metallic concentration and the radioactivity by precipitating metals in solution as hydroxides M(m-n)(OH)n (with m the oxidation number of the metal M). The optimization of a two-step precipitation protocol is presented here, with a study of the contact time (1–24 h) and the reagents used (NaOH and KOH). The resulting precipitates from this process are characterized using several techniques (FTIR, TGA and XRD). Tests were conducted on surrogate samples to demonstrate the viability of the process on more complex samples. Finally, the optimized protocols were implemented on radioactive Ni-alloy samples. Decontamination factors were calculated portraying the efficiency of both the COREMIX and the subsequent two-stage precipitation process. Characterization of the sludge produced during the process shows that the precipitate obtained at pH 8.5 consists mainly of iron (III) oxide-hydroxides, whereas the precipitate obtained at pH 12 is mainly composed of manganese (II,III) oxide. The optimization steps show that the contact time during the first precipitation and the choice of precipitants does not influence the efficiency of the protocol while the destruction of oxalic acid proves to be critical to quantitatively precipitate chromium. Ultimately, the COREMIX process can effectively decontaminate contaminated Ni-alloy samples, removing between 12% and 14% of the contamination in each cycle. Decontamination of effluent using the precipitation protocol results in a very high decontamination factor of between 3000 and 6000.

Oxalic acid is encountered within industrial processes, spanning from the nuclear sector to various chemical applications. The persistence and potential environmental risks associated with this compound underscore the need for effective management strategies. This article presents an overview of different approaches for the destruction of oxalic acid. The study explores an array of degradation methodologies and delves into the mechanistic insights of these techniques. Significant attention is channeled towards the nuclear industry, wherein oxalic acid arises as a byproduct of decontamination and waste management activities. An integral aspect of decommissioning efforts involves addressing this secondary waste-form of oxalic acid. This becomes imperative due to the potential release of oxalic acid into waste streams, where its accommodation is problematic, and its capacity to solubilize and transport heavy metals like Pu is a concern. To address this, a two-tiered classification is introduced: high concentration and low concentration scenarios. The study investigates various parameters, including the addition of nitric acid or hydrogen peroxide, in the presence of metallic ions, notably Mn2+ and Fe2+. These metallic ions are common components of effluents from metallic waste treatment. Additionally, the impact of UV light on degradation is explored. Investigations reveal that at high concentrations and with the influence of hydrogen peroxide, the presence of metallic cations accelerates the rate of destruction, demonstrating a direct correlation. This acceleration is further enhanced by exposure to UV light. At low concentrations, similar effects of metallic cations are observed upon heating the solution to 80°C. The rate of destruction increases proportionally with hydrogen peroxide concentration, with an optimal oxalic acid to hydrogen peroxide ratio of 1:100. Interestingly, a low-power UV light exerted no discernible effects on the destruction rate; heating alone proved sufficient. In essence, regardless of concentration, the degradation of oxalic acid with hydrogen peroxide experiences acceleration in the presence of metallic ions such as Mn2+ and Fe2+.

Abstract A Na-alumino-borosilicate glass and its Ca-doped counterpart were altered in vapor phase (98% relative humidity) and aqueous medium at 90 °C. Both the alteration media were enriched in 17O. Characterization of the altered samples pointed out some differences between glass alteration in aqueous medium and vapor phase in terms of the alteration kinetics, the effect of Ca-doping on glass chemical durability, the behavior of elements in the gel layer, and the structure of the gel layer. Some of the key results are the recondensation of boron in the gel layer formed in the vapor phase and the utility of Rotation Echo DOuble Resonance Nuclear Magnetic Resonance spectroscopy to qualitatively distinguish between signals from the pristine glass and hydrated gel layer within a sample that was not altered to the core. The results gave rise to inferences about glass alteration mechanisms in both the alteration media and the differences between them.

A sealed container for the geological disposal of spent nuclear fuel and vitrified high-level waste is the only component of a deep geological repository that provides complete containment of radionuclides. As such, attention is focused on its lifetime. The lifetime of the container is influenced by material degradation processes during disposal and is typically of the order of several millennia and, for some container materials, up to one million years. Designing, manufacturing, and predicting the performance of containers over such long periods requires an indepth understanding of their material properties, fabrication processes, and degradation mechanisms. Scientific and technological progress can improve both the performance of containers and the robustness of lifetime predictions. Optimization of these aspects is of primary importance for many national radioactive waste disposal programs. In this article, the state of the art of complex coupled degradation processes, as well as the optimization potential of novel container materials, is presented. Furthermore, the existing tools allowing the prediction of long-term barrier integrity are discussed. [ABSTRACT FROM AUTHOR]

Abstract Understanding the durability of nuclear glass in the vapor phase is essential for evaluating the long-term safety of the nuclear glass. This work studies the vapor hydration of borosilicate and iodine-bearing glass at 90 °C and relative humidity at 99%. Hydration kinetics was monitored by analyzing the boron and iodine released during the vapor hydration process. It is found that boron becomes more easily released after iodine being added to the glass matrix. The results related to boron release and retention indicate the possible transformation of tetrahedral [4]B to trigonal [3]B on the hydrated glass surface, which may enhance the release of boron in the form of boric acid afterward. Release of iodine occurred but at a far lesser extent than boron with 92% of iodine retained in the alteration layer, thus opening good perspectives with regard to the disposal of high-level waste containing iodine.

Nuclear power plays a major role in the generation of electricity with low carbon emissions. However, it generates significant amounts of radioactive waste, mainly from contaminated metallic components such as steam generators. Decontamination is essential for the safe handling and eventual recycling or disposal of these materials. Various decontamination techniques can be utilized but chemical processes are recommended for complex geometries such as the tubular parts of steam generators. COREMIX (Chemical Oxidation REduction using nitric permanganate and oxalic acid MIXture) is a process that is similar to the CORD (Chemical Oxidation Reduction Decontamination) process currently utilized in the industry which involves dissolving the contaminated oxide layers from metallic surfaces. This process generates a large quantity of radioactive effluent that requires appropriate treatment. The objective is to reduce metallic concentration and the radioactivity by precipitating metals in solution as hydroxides M(m-n)(OH)n (with m the oxidation number of the metal M). The optimization of a two-step precipitation protocol is presented here, with a study of the contact time (1–24 h) and the reagents used (NaOH and KOH). The resulting precipitates from this process are characterized using several techniques (FTIR, TGA and XRD). Tests were conducted on surrogate samples to demonstrate the viability of the process on more complex samples. Finally, the optimized protocols were implemented on radioactive Ni-alloy samples. Decontamination factors were calculated portraying the efficiency of both the COREMIX and the subsequent two-stage precipitation process. Characterization of the sludge produced during the process shows that the precipitate obtained at pH 8.5 consists mainly of iron (III) oxide-hydroxides, whereas the precipitate obtained at pH 12 is mainly composed of manganese (II,III) oxide. The optimization steps show that the contact time during the first precipitation and the choice of precipitants does not influence the efficiency of the protocol while the destruction of oxalic acid proves to be critical to quantitatively precipitate chromium. Ultimately, the COREMIX process can effectively decontaminate contaminated Ni-alloy samples, removing between 12% and 14% of the contamination in each cycle. Decontamination of effluent using the precipitation protocol results in a very high decontamination factor of between 3000 and 6000. [ABSTRACT FROM AUTHOR]

This study aims to investigate the groundwater vulnerability concerning potentially toxic elements in the vicinity of the abandoned iron mine of Ouixane (Morocco). A modified DRASTIC method (DRSTI) is proposed with satisfactory results. High vulnerability zones represent 40% of the study area, while medium and low vulnerability represent, respectively, 42% and 18% of the study area. These results have been validated by groundwater geochemical analyses of potentially toxic elements carried out in the framework of previous studies in the same area. Thus, the superposition of the waste rock and tailings map with the vulnerability map showed that the latter is located in areas of high to medium vulnerability and therefore constitutes the main cause of the deterioration of the geochemical quality of groundwater. Otherwise, the vulnerability method showed that the main parameters that significantly affect the vulnerability are: the depth of water (D), net recharge (R), and the unsaturated zone (I), while the other parameters do not significantly affect the model used and do not have much influence on the results of the vulnerability assessment. The method used allowed us to locate the most vulnerable areas to potentially metallic toxic elements pollution resulting from the abandoned iron mine of Ouixane, and it constitutes a tool for decision support and for developing effective action plans to mitigate and monitor the effects of the transfer of potentially toxic elements pollution to groundwater.

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.