Decommissioning waste is generated at all stages during the decommissioning of nuclear facilities, and various types of radioactive waste are generated in large quantities within a short period. Concrete is a major building material for nuclear facilities. It is mixed with aggregate, sand, and cement with water by the relevant mixing ratio and dried for a certain period. Currently, the proposed treatment method for volume reduction of radioactive concrete waste was involved thermomechanical and chemical treatment sequentially. The aggregate as non-radioactive materials is separated from cement components as contaminated sources of radionuclides. However, to commercialize the process established in the laboratory, it is necessary to evaluate the scale-up potential by using the unit equipment. In this study, bench-scale testing was performed to evaluate the scale-up properties of the thermomechanical and chemical treatment process, which consisted of three stages (1: Thermomechanical treatment, 2: Chemical treatment, 3: Wastewater treatment). In the first stage, lab, bench, and pilot scale thermomechanical tests were performed to evaluate the treated coarse aggregate and fines. In the second stage, the fine particles generated by the thermomechanical treatment process, were chemically treated using dissolution equipment, after then the removal efficiency and residual of cement in the small aggregate was compared with laboratory results. The final stage, the secondary wastewater containing contaminant nuclides was treated, and the contaminant nuclides could be removed by chemical precipitation method in the scale-up reactors. Furthermore, an additional study was required on the solid-liquid separation, which connected each part of the equipment. It was conducted to optimize the separation method for the characteristics of the particles to be separated and the purpose of separation. Therefore, it is expected that the basic engineering data for commercialization was collected by this study.

Thermal decomposition of the uranyl phosphate mineral phase meta-ankoleite (KUO2PO4·3H2O) has been considered in relation to high temperature thermal sintering for the immobilisation of a uranyl phosphate containing waste. Meta-ankoleite thermal decomposition was studied across the temperature range 25 – 1200℃ under an inert N2 atmosphere at 1 atm. It is shown that the meta-ankoleite mineral phase undergoes a double de-hydration event at 56.90 and 125.85℃. Subsequently, synthetically produced pure meta-ankoleite remains stable until at least 1150℃ exhibiting no apparent phase changes. In contrast, when present in a mixed waste the meta-ankoleite phase is not identifiable after thermal treatment indicating incorporation within the bulk waste either as an amorphous phase and/or as uranium oxide. Visual inspection of the waste post thermal treatment showed evidence of self-sintering owing to the presence of glass former materials, namely, silica (SiO2) and antimony(V) oxide (Sb2O5). Therefore, incorporation of the uranium phase into the waste as part of waste sentencing and immobilisation via high temperature sintering for the purpose of long-term disposal is deemed feasible.