Molten Salt Reactor (MSR) is one of Generation-IV nuclear reactors that uses molten salts as a fuel and coolant in liquid forms at high temperatures. The advantages of MSR, such as safety, economic feasibility, and scalability, are attributed from the fact that the molten salt fuel in a liquid state is chemically stable and has excellent thermo-physical properties. MSR combines the fuel and coolant by dissolving the actinides (U, Th, TRU, etc.) in the molten salt coolant, eliminating the possibility of a core meltdown accident due to loss of coolant (LOCA). Even if the molten salt fuel leaks, the radioactive fission products dissolved in the molten salt will solidify with the fuel salt at room temperature, preventing potential leakage to the outside. MSR was first demonstrated at ORNL starting with the Aircraft Reactor Experiment (ARE) in 1954 and was extended to the 7.4 MWth MSRE developed in 1964 and operated for 5 years. Recently, various start-ups, including TerraPower, Terrestrial Energy, Moltex Energy, and Seaborg, have been conducting research and development on various types of MSR, particularly focusing on its inherent safety and simplicity. While in the past, fluoride-based molten salt fuels were used for thermal neutron reactors, recently, a chlorine-based molten salt fuel with a relatively high solubility for actinides and advantageous for the transmutation of spent nuclear fuel and online reprocessing has been developing for fast neutron spectrum MSRs. This paper describes the development status of the process and equipment for producing highpurity UCl3, a fuel material for the chlorine-based molten salt fuel, and the development status of the gas fission product capturing technologies to remove the gaseous fission products generated during MSR operation. In addition, the results of the corrosion property evaluation of structural materials using a natural circulation molten salt loop will also be included.

Based on the results of a review for various precipitation methods phosphorylation (phosphate precipitation) of metal chlorides considered as a proper treatment method for recovering of the fission products in a molten salt. In previous precipitation tests, the powder of lithium phosphate (Li3PO4) added into LiCl-KCl molten salt containing metal chlorides as a precipitation agent. The reaction of metal chlorides containing actinides and rare earths to recover with lithium phosphate in a molten salt known as solid-liquid reaction. The powder of lithium phosphate disperse in a molten salt by stirring thoroughly in order to enhance the precipitation reaction. As a result, metal phosphates as the reaction products precipitate on the bottom of the vessel and cutting at the lower part of the salt ingot considered as one of the recovery method of the precipitates. Recently, the vacuum distillation of upper part of the salt proposed as another recovering method. Cutting method of precipitate at the lower part of the salt ingot would be difficult to handle the increased size of the salt ingot produced from the practical scale equipment. In this presentation, a new method for collecting the precipitates of phosphorylation reaction into a small vessel is introduced with test results in a molten salt containing uranium and rare earths such as Nd, Ce, and La. As the first step of a series of test lithium phosphate ingot was prepared by melting the powder at a temperature 1,300°C, and the ingot put into LiCl-KCl molten salt at 500°C for more than three hours to examine the shape of ingot to be deformed or not. The phosphorylation experiments using lithium phosphate ingots carried out to collect the metal phosphate precipitates and the test result of this new method was feasible.

Under the circumstance of energy transition policy of the previous government in which nuclear energy portion will be gradually reduced, some R&D study looking for alternatives other than Pyro- SFR recycling could be very valuable and timely suitable. New alternative study started to evaluate the possibility of it if there are some advantages in terms of waste burden in case that the spent fuel are appropriately treated and disposed of in a disposal site, instead of recycling of spent nuclear fuels (SNF). The alternative study separate the fission products (minor actinides and rare earths) from SNF in a molten salt medium. The molten salt coming from the alternative study is radioactive and heat generating because it contains the fission products chlorides. It is necessary to collect the fission products from the waste molten salt for minimization of the high-level waste volume and to generate a final waste form containing the fission products compatible to the disposal site. Based on the results of a review for various precipitation methods, phosphorylation (phosphate precipitation) of metal chlorides selected as a proper treatment method for recovering of the fission products in a molten salt. Phosphate precipitation has the potential for removing most of fission product elements from a molten salt arising from the treatment of spent nuclear fuel. The performance of phosphate precipitation method evaluated using a salt mixture with the actinide and rare earth chlorides. The molten salt containing uranium as surrogate of the actinides and three rare earths (Nd, Ce, La) chloride was used for testing a phosphate precipitation method at experimental condition (temperature 500°C, salt stirring 200~300 rpm, and 1~1.2 eq. of phosphorylation agent). A cyclic voltammetry (CV) method monitored in-situ phosphate precipitation progress for determining the precipitation rate and conversion ratio evaluated. The phosphorylation reaction increased greatly at a salt stirring 300 rpm.

Ag-containing aluminosilicate sorbents capable of capturing iodine were prepared by sol-gelation from Na, Al, and Si alkoxides using co-solvent exchange, Ag/Na ion exchange, solvent exchange, and ambient-pressure drying. The Na+AlSi-OH gel was prepared using sodium methoxide (NaOMe): aluminum tri-sec-butoxide (Al(OsBu)3): tetraethyl orthosilicate (TEOS) molar ratios of 1.05:1:1, 1.3:1.1:1, 1.5:1.3:1. The solvent effect on textural properties such as Brunauer-Emmett-Teller (BET) surface areas and pore size distributions and Ag0 particle sizes was investigated using water with high surface tension, isopropanol and n-heptane with low surface tension. The BET surface area, average pore size, and cumulative pore volume for sorbents strongly increased with decreasing surface tension of solvents and increasing Al/Si atomic ratios. In addition, Ag0 particle sizes increased with decreasing surface tension of solvents.

Interests in molten salt reactor (MSR) using a fast spectrum (FS) have been increased not only for having a high power density but for burning the high-level waste generated from nuclear power plants. For developing the FS-MSR technologies, chloride-based fuels are considered due to the advantage of higher solubility of actinides and lanthanides over fluoride-based salts. Despite significant progress in development of MSR technology, the manufacturing technology for production of the fuel is still insufficiently understood. One of the option to prepare the MSR fuel is to use products from pyroprocessing where oxide form of spent nuclear fuel is reduced into metal form and useful elements can be collected via electrochemical methods in molten salt system at high temperature. In order to chlorinate the products into chloride form, previous study used NH4Cl to chlorinate U metal into UCl3 in an airtight reactor. It was found that the U metal was completely chlorinated into chloride forms; however, impurities generated by the reaction of NH4Cl and reactor wall were found in the product. Therefore, in this work, the air tight reactor was re-deigned to avoid the reaction of reactor wall by insertion of Al2O3 crucible inside of the reactor. In addition, the reactor size was increased to produce UCl3 over 100 g. Using the newly designed reactor, U metal chlorination experiments using NH4Cl chlorinating agent were performed to confirm the optimal experimental conditions. The detailed results will be further discussed.

It has been studied on the disposal area reduction for the used nuclear fuel by the management of high decay-heat nuclides, long-lived nuclides, and highly mobile nuclides. It was investigated on the management of the nuclides in KAERI. Strontium-90 is a high heat-generating nuclide in spent nuclear fuel. It is needed to separate the salt from the salt solution for the recovery of strontium after the chlorination of the strontium oxide in molten salt. Vacuum distillation was used for the separation of strontium from the molten salt. Potassium carbonate was chosen as a reactive distillation reagent for SrCl2 – LiCl – KCl system by the thermodynamic calculation. Reactive distillation experiments were carried out. The residual was mainly SrCO3 in the XRD analysis. It could be concluded that K2CO3 could be one of the suitable reagents for the reactive distillation. The salt in the long–lived nuclide powders should be removed to prepare the block for disposal. Experiments were carried out using W powders (surrogate) and U3O8 powders to develop a process for the removal of the residual salt from UOx powders. The salts were successfully removed from the W and U3O8 powders by distillation.

The effect of Li2O addition on precipitation behavior of uranium in LiCl-KCl-UCl3 has been investigated in this study. 99.99% LiCl-KCl eutectic salt is mixed with 10wt% UCl3 chips at 550°C in the Pyrex tube in argon atmosphere glove box, with 10 ppm O2 and 1 ppm H2O. Then, Li2O chunks are added in mixed LiCl-KCl-UCl3 and the system has been cooled down to room temperature for 10 hours to form enough UO2 particles in the salt. The solid salt has been taken out from the glove box, and cut into three sections (top, middle and bottom) by low-speed saw for further microscopic analysis. Three pieces of solid salt are dissolved in deionized water at room temperature and the solution is filtered by a filter paper to collect non-dissolved particles. The filter paper with particles is baked in vacuum oven at 120°C for 6 hours to evaporate remaining moisture from the filter paper. Further analysis was performed for the powder remaining on the filter paper, and periphery of the powder (cake) on the filter paper. Scanning electron microscopy (SEM), electron diffraction spectroscopy (EDS), and X-ray powder diffraction (XRD) are adopted to analysis the characteristic of the particles. From SEM analysis, the powders are consisted of small particles which have 5 to 10 m diameter, and EDS analysis shows they are likely UO2 with 23 at. % of uranium and 77 at. % oxygen. Cake is also analyzed by SEM and EDS, and needle like structures are widely observed on the particle. The length of needle is distributed from 5 to 20 m, and it has 6 to 10 at. % of chlorine, which are not fully dissolved into deionized water at room temperature. From XRD analysis, the particles show the peak position of UO2, and the result is well matched with the SEM-EDS results. We are planning to add more Li2O in the system for fully reacting uranium in UCl3, and compare the results to find the effect of Li2O concentration on UO2 precipitation.

Thermodynamically, TRUOx, REOx, and SrOx can be chlorinated using ammonium chloride (NH4Cl) as a chlorinating agent, whereas uranium oxides (U3O8 and UO2) remain in the oxide form. In the preliminary experiments of this study, U3O8 and CeO2 are reacted separately with NH4Cl at 623 K in a sealed reactor. CeO2 is highly reactive with NH4Cl and becomes chlorinated into CeCl3. The chlorination yield ranges from 96% to 100%. By contrast, U3O8 remains as UO2 even after chlorination. We produced U/REOx- and U/SrOx-simulated fuels to understand the chlorination characteristics of the oxide compounds. Each simulated fuel is chlorinated with NH4Cl, and the products are dissolved in LiCl-KCl salt to separate the oxide compounds from the chloride salt. The oxide compounds precipitate at the bottom. The precipitate and salt phases are sampled and analyzed via X-ray diffraction, scanning electron microscope-energy dispersive spectroscopy, and inductively coupled plasma-optical emission spectroscopy. The analysis results indicate that REOx and SrOx can be easily chlorinated from the simulated fuels; however, only a few of U oxide phases is chlorinated, particularly from the U/SrOx-simulated fuels.

An accumulation of spent nuclear fuel (SNF) has brought a considerable interest due to its energy and environmental issue. To effectively manage SNF, a pyroprocessing is introduced to separate useful resources from the spent fuels and to manufacture suitable fuels. In head-end process of pyroprocessing, spent fuels are thermally treated to prepare UO2 pellets, where various radioactive gases from SNFs are released during thermal treatment. Within these gases, C-14 as CO2 form is a radioactive fission product which had a long half-life of 5,730 years and emits beta radiation of 0.156 MeV. Generally, current CO2 capturing technologies include adsorption by solid materials, absorption by aqueous solutions, and membrane separation. Among these methods, absorption is an effective approach which traps CO2 effectively and and it is easy to operate at room temperature. In addition, it is highly recommended as immobilizing 14CO2 as CaCO3 formation due to the high thermal and chemical stability, and the relatively low solubility in water. Generally, a double alkali method has been proposed to capture low concentrated 14CO2 from the stream. This method for CO2 capture includes absorption process with NaOH solution and causticization using Ca(OH)2. In this study, CO2 emitted from SNF is captured using double alkali method, and the effects of operating conditions on capturing efficiency were investigated. Furthermore, considering the two-film theory, the effects of trapping conditions on the CO2 absorption performance were examined. The recovered CaCO3 from causticization was collected from the absorbing solution and analyzed.

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.

To estimate the removal efficiency of TRU and rare earth elements in an oxide spent fuel, basic dissolution experiments were performed for the reaction of rare earth elements from the prepared simfuel with chlorination reagents in LiCl-KCl molten salt. Based on the literature survey, NH4Cl, UCl3, and ZrCl4 were selected as chlorination reagent. CeO2 and Gd2O3 powders were mixed with uranium oxide as a representative material of rare earth elements. Simfuel pellets were prepared through molding and sintering processes, and mechanically pulverized to a powder form. The experiments for the reaction of the simfuel powder and chlorination reagents were carried out in a LiCl-KCl molten salt at 500°C. To observe the dissolution behavior of rare earth elements, molten salt samples were collected before and after the reactions, and concentration analysis was performed using ICP. After the reaction completed, the remaining oxide was washed with water and separated from the molten salt, and XRD was used for structural analysis. As a result of salt concentration analysis, the dissolution performance of rare earth elements was confirmed in the reaction experiments of all chlorination reagents. In an experiment using NH4Cl and ZrCl4, the uranium concentration in the molten salt was also measured. In other words, it seemed that not only rare elements but also uranium oxide, which is a main component of simfuel, was dissolved. Therefore, it is thought that the dissolution of rare earth elements is also possible due to the collapse of the uranium oxide structure of the solid powder and the reaction with the oxide of rare earth elements exposed to molten salt. As a result of analyzing the concentration changes of Simfuel before and after each reaction, there was little loss of uranium and rare earth elements (Ce/Gd) in the NH4Cl experiment, but a significant amount of rare earth elements were found to be reduced in the UCl3 experiment, and a large amount of rare earth elements were reduced in the ZrCl4 reaction.

Strontium-90 is a high heat-generating nuclide in spent nuclear fuel. The removal of the nuclide separation is indispensable to reduce the burden of storage and disposal of high-level radioactive waste. Korea Atomic Energy Research Institute has developed the molten salt immersion technique to separate the strontium by the chlorination of the strontium oxide in molten salt. It is needed to separate the salt for the recovery of strontium from the salt solution after the chlorination reaction. In this study, it was investigated on the recovery of the strontium from the salt. Vacuum distillation was used for the separation of strontium from the molten salt. The vapor pressures of the candidate salts were calculated by HSC chemistry and the apparent evaporation rates (AER) were measured at 830°C to evaluate the salts for strontium recovery. The candidate salts were LiCl, KCl, MgCl2, NaCl and CaCl2. The AERs of MgCl2 and NaCl were 1.9 and 1.3 g/cm2-h, respectively. Those two salts can be separated from the strontium compound even though the AER values are much lower than those of LiCl-KCl (~ 8 g/cm2-h). CaCl2 salt was rarely evaporated (AER < 0.03 g/cm2-h) and it is not suitable to use as a strontium recovery salt. Therefore, MgCl2, NaCl, LiCl and KCl can be regarded as candidates for a strontium recovery salt.

Molten Salt Reactor (MSR) is one of the generation-IV advanced nuclear reactors in which hightemperature molten salt mixture is used as the primary coolant, or even the fuel itself unlike most nuclear reactors that adopt solid fuels. The MSR has received a great attention because of its excellent thermal efficiency, high power density, and structural simplicity. In particular, since the MSR uses molten salts with boiling points higher than the exit temperature of the reactor core, there is no severe accident such as a core melt-down which leads to a hydrogen explosion. In addition, it is possible to remove the residual heat through a completely passive way and when the fuel salt leaks to the outside, it solidifies at room-temperature without releasing radioactive fission products such as cesium, which make the MSR inherently safe. Both fluoride and chloride mixtures can be used as liquid fuel salts by adding actinide halides for MSRs. However, the MSRs using chloride-based salt fuels can be operated for a long time without adding nuclear fuel or online reprocessing because the actinide solubility in chloride salts is about six times higher than that in fluoride salts. Therefore, the chloride-based MSRs are more effective for the transmutation of long-lived radionuclides such as transuranic elements than the fluoride-based MSRs, which is beneficial to resolve the high radioactive spent nuclear fuel generated from light water reactors (LWRs). This paper examines liquid fuel fabrication using an improved U chlorination process for the chloride-based MSRs and presents the strategy for the management of gaseous fission products generated during the operation of MSR.