The separation efficiency of nuclides in molten salt systems was investigated, with a focus on the influence of apparatus configuration and experimental conditions. A prior study revealed that achieving effective Sr separation from simulated oxide fuel required up to 96 hours, reaching a separation efficiency of approximately 90% using a static dissolution reaction in a porous alumina basket. In this study, we explored the impact of agitation on improving Sr separation efficiency and dissolution rates. The simulated oxide fuel composition consisted of 2wt% Sr, 3wt% Ba, 2wt% Ce, 3wt% Nd, 3wt% Zr, 2wt% Mo, and 89wt% U. To quantify the Sr concentration in the salt, we utilized ICP analysis after salt sampling via a dip-stick technique. Furthermore, we conducted ICPOES analysis over a 55-hour duration to assess the separated nuclides. Complementing these analyses, SEM and XRD investigations were performed to validate the crystal structure and morphology of the oxide products.

Separating nuclides from spent nuclear fuel is crucial to reduce the final disposal area. The use of molten salt offers a potential method for nuclide separation without requiring electricity, similar to the oxide reduction process in pyroprocessing. In this study, a molten salt leaching technique was evaluated for its ability to separate nuclides from simulated oxide fuel in MgCl2 molten salts at 800°C. The simulated oxide fuel contained 2wt% Sr, 3wt% Ba, 2wt% Ce, 3wt% Nd, 3wt% Zr, 2wt% Mo, and 89wt% U. The separation of Sr from the simulated oxide fuel was achieved by loading it into a porous alumina basket and immersing it in the molten salt. The concentration of Sr in the salt was measured using ICP analysis after sampling the salt outside the basket with a dip-stick technique. The separated nuclides were analyzed with ICP-OES up to a duration of 156 hours. The results indicate that Ba and Sr can be successfully separated from the simulated fuel in MgCl2, while Ce, Nd, and U were not effectively separated.

Molten salt immersion technique has been tested with several Sr oxides, SrZrO3, SrMoO4 and U2SrOy, and MgCl2 based molten salts for the Sr nuclide separation. Reaction time, temperature, and salt composition were varied to effectively separate Sr in chloride forms. ICP-OES, XRD, and SEM analysis were conducted for the conversion efficiency and structure and morphology analysis. It is confirmed that all experiments of SrZrO3 with MgCl2 at 800°C for reaction time 5, 10, 20 hours showed higher conversion efficiency than 99% and in LiCl-KCl-MgCl2 and NaCl-MgCl2 molten salts at 500°C or 600°C, conversion efficiency higher than 97% was obtained. SrMoO4 in MgCl2 immersion experiments for 10 hours showed higher conversion efficiency than 99% when the molar ratio of salt/oxide powder is 7. U2SrOy was also tested with MgCl2 molten salt at 800°C and higher efficiency than 99% and mainly MgUO4 were produced as a reaction product.

Nuclide separation of commercial nanofiltration (NF) & reverse osmosis (RO) membranes is investigated to elucidate the permeation mechanism and possibility on applying treatment of wastewater from nuclear facility. Since wastewater from nuclear facility contains significant amount of dangerous nuclides, their treatment for recycle or clearance is necessary. Polyamide based commercial membranes are prepared to investigate their separation performance on single- & multi-component nuclide aqueous solutions. Similar to their desalination performance, NF shows higher flux, but RO has high rejection. With respect to rejection, NF and RO membranes have better performance comparing to desalination. This study indicates that NF and RO membranes can be potential candidate for nuclide separation.

Radioactive liquid waste has given soil & water pollution and further congenital defect to the humanity. With this regard, there are lots of interests in reducing concentration of radioactive material even if it is in low level (ppm). There are filtration driven by pressure such as ultra-filtration, nano-filtration, and reverse osmosis to separate radioactive liquid waste. Among them, Nano-Filtration membrane shows promising performance in both permeability and selectivity. Commercial nanofiltration membrane were applied to separate model solutions. Model seawater was desalinated to 6,100 ppm with flux of 90 L/ ㎡h. On the other hands, Model nuclides (Co, Cs, and Sr) solutions are separated from 10 ppm to 2 ppm, 0.23 ppm, and 0.13 ppm. Commercial nanofiltration membranes show the trade-off property between selectivity and flux.