Molten Salt Reactor, which employs molten salt mixture as fuel, has many advantages in reactor size and operation compared to conventional nuclear reactor. In developing Molten Salt Reactor, the behavior of fission product in operation should be preliminary evaluated for the correct design of reactor and its associated system including off-gas treatment. In this study, for 100 Mw 46 KCl- 54 UCl3 based Molten Salt Reactor with operating life time of 20 year, the fission product behavior was estimated by thermodynamic modeling employing FactSage 8.2. Total inventory of all fission product were firstly calculated using OpenMC code allowing depletion during neutronic calculation. Then, among all inventory, 46 element species from Uranium to Holmium were chosen and given to the input for equilibrium module of Factsage with its mass. In phase equilibrium calculation, for the correct description of solution phase, KCl-UCl3 solution database based on modified quasichemical model in the quadruplet approximation (ANL/CFCT-21/04) was employed and the coexisting solid phase was assumed to pure state. With the assumption of no oxygen and moisture ingress into reactor system, equilibrium calculation showed that 1% of solid phase and of gas phase were newly formed and, in gas phase, major species were identified : ZrCl4 (47%), Xe (33%), UCl4 (14%), Kr (5%), Ar (1%) and others. This result reveals that off-gas treatment of system should account for the appropriate treatment of ZrCl4 and UCl4 besides treatment of noble gas such as Xe and Kr.

In KAERI, the nuclide management technology is currently being developed for the reduction of disposal area required for spent fuel management. Among the all fission products of interest, Cs, I, Kr, Tc are considered to be significantly removed by following mid-temperature and hightemperature treatment, however, a difficulty of real spent-fuel thermal treatment experiment limits the development of such thermal treatment. The test employing SimFuel (Simulated Spent Fuel) can be an alternative for such condition, however, the fabrication of SimFuel containing semivolatile species such as Cs, I and Re (substitute for Tc) was not achieved for conventional sintering method since such species are easily removed during hot temperature treatment. In this study, for the prevention of volatilization of such species and the inclusion of semi-volatile species in fabrication of SimFuel, argon-based high pressurizing up to Max 100 bar was considered to be applied in high temperature treatment. For this, lab-scale hot-isostatic press applicable up to 1,500°C was fabricated and is being waiting for the approval for high-pressure test. After approval of license, UO2 baesd SimFuel containing CsI will be fabricated and its micro-structure and composition will be evaluated through SEM-EDX and XRD

In KAERI’s previous phosphate precipitation tests, the dispersed powder of lithium phosphate (Li3PO4) as a precipitation agent reacted with various metal chlorides in a simulated LiCl-KCl molten salt. The reaction of metal chlorides composed of actinides such as uranium and three rare earths (Nd, Ce and La) with lithium phosphate is a solid-liquid reaction. A phosphorylation reaction rate is very fast and the metal phosphates as a reaction product precipitated on the bottom of the molten salt crucible. One of the recovery methods of the metal phosphate precipitates is segregation the lower part (precipitates) of the salt ingot using the various cutting tools. Recently, a new phosphorylation experiment using lithium phosphate ingots carried out in order to collect the metal phosphate precipitates into a small recovering vessel, and the test result of this new method was feasible. However, the reaction rate of test using lithium phosphate ingot is extremely slower than that of test using lithium phosphate powder. In this study, the precipitation reactor design (a tapered crucible with polished inner surface) used for phosphorylation reaction showed that the salt ingot with metal phosphate precipitates could be detached from a tapered stainless steel crucible. We propose that the recovery of precipitates from a salt ingot is possible by introducing a dividing plate structure into a molten salt and by positioning it at the interface between salt and precipitated metal phosphate.

Currently, the Korea Atomic Energy Research Institute is conducting research on the development of technology to reduce the disposal area for SF (Spent nuclear Fuel). If the main radionuclides contained in SF can be separated and recovered according to their characteristics (long half-life, high mobility and high heat load) and uranium oxide which is expected to be the final residue, can be made into solids, the burden of the permanent disposal area of the SF will be greatly reduced. The waste form that end up in the repository must be verified for ease of manufacture and stability of the block. And, in order to increase the loading efficiency, a large block manufacturing technology is needed. This study describes the background of introducing PSA (Particle Size Analyzer) which is one of the necessary equipment for manufacturing UO2 blocks using slip casting, the method of using the equipment and performance verification of the equipment using standard samples. The particle size affects the sintering quality by the way the particles rearrange themselves during sintering. Powders of small particles are generally less free flowing and more difficult to compress, they form thin pores between the particles and sinter to higher density. In contrast, larger particle has a lower sintered density. Therefore, accurate particle size measurement and the selection of a suitable particle size are important. For this purpose, a PSA was installed in nuclear cycle experiment research center. To verify the performance of the equipment, a standard sample of 1.025 μm was analyzed. We got an average particle size of 1.0293 μm and standard deviation of 0.0668 μm. This value was within the uncertainty(±0.018 μm) of the sample’s certificate. In the future, this equipment will measure the size of UO2 (depleted uranium) powder and to produce large scale uranium oxide blocks.

Pyroprocessing technology has emerged as a viable alternative for the treatment of metal/oxide used fuel within the nuclear fuel cycle. This innovative approach involves an oxide reduction process wherein spent fuel in oxide form is placed within a cathode basket immersed in a molten LiCl-Li2O salt operating at 923 K. The chemical reduction of these oxide materials into their metallic counterparts occurs through a reaction with Li metal, which is electrochemically deposited onto the cathode. However, during process, the generation of Li2O within the fuel basket is inevitable, and due to the limited reduction efficiency, a significant portion of rare earth oxides (REOx) remains in their oxide state. The presence of these impurities, specifically Li2O and REOx, necessitates their transfer into the electrorefining system, leading to several challenges. Both Li2O and REOx exhibit reactivity with UCl3, the primary electrolyte within the electrorefining system, causing a continuous reduction in UCl3 concentration throughout the process. Furthermore, the formation of fine UO2 powder within the salt system, resulting from chemical reactions, poses a potential long-term operational and safety concern within the electrorefining process.Various techniques have been developed to address the issue of UO2 fine particle removal from the salt, utilizing both chemical and mechanical methods. However, it is crucial that these methods do not interfere with the core pyroprocessing procedure. This study aims to investigate the impact of Li2O and REOx introduced from the electrolytic reduction process on the electrorefining system. Additionally, we propose a method to effectively eliminate the generated UO2 fine powder, thereby enhancing the long-term operational stability of the electrorefining process. The efficiency of this proposed solution in removing oxidized powder has been confirmed through laboratory-scale testing, and we will provide a comprehensive discussion of the detailed results.

It is known that ZrCl4 can be used in the chlorination process of spent nuclear fuel. However, its solubility in high temperature molten salt is very small, making it difficult to dissolve a large amount of ZrCl4. Therefore, in this study, a flange-type sealed reactor was manufactured to observe the reaction characteristics of LiCl-KCl salt and ZrCl4. LiCl-KCl salt and ZrCl4 were placed in each alumina crucible, the reactor was sealed, and heated. The temperature at the reactor surface was above 500°C and maintained at that temperature for 48 hours. After completion of the reaction, the reactor was opened and the reaction products were recovered from each alumina crucible. The crystal structure of the reaction product was identified through XRD analysis, and the concentration of Zr was analyzed using ICP. Reaction characteristics were observed according to the molar ratio of ZrCl4 added to the number of moles of KCl in LiCl-KCl salt. The molar ratios of ZrCl4 to KCl were 0.5, 1, 2, and 3, respectively. As a result of each experiment, more than 95% of the injected ZrCl4 was vaporized and there was almost no residue in the ZrCl4 crucible. In the LiCl- KCl crucible, the weight increased in proportion to the amount of ZrCl4 added. As a result of XRD analysis, K2ZrCl6 was confirmed in all LiCl-KCl salt product. When the ZrCl4/KCl molar ratio was 2 and 3, LiCl-KCl could not be confirmed. Additionally, when the ZrCl4/KCl molar ratio was 1, LiCl was identified, but KCl was not found. Almost all of the KCl appears to have reacted with ZrCl4. ICP analysis results showed that the Zr concentration was proportional to the amount of ZrCl4 added in each LiCl-KCl salt, and exceeding the number of moles of reaction with KCl in the LiCl-KCl salt was observed. Therefore, these experimental results showed that ZrCl4 can be dissolved in LiCl-KCl salt at a maximum concentration higher than its solubility.