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.

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.

MCFC 작동온도인 650˚C에서 음극의 creep과 소결에 의한 구조적 변형을 막기 위해 기계적 합금법에 의한 Ni-WC분말을 합금화하여 변형에 대한 저항성을 증대시키고자 하였다. 80시간동안 어트리션 밀링을 실시한 분말은 XRD 분석결과 결정규칙이 파괴된 비정질 상을 보였다. 제조된 분말은 적당한 점도의 슬러리로 제조후 테이프 캐스팅법에 의해 green sheet를 제조하였다. 제조된 박판의 두께는 0.9mm였고, 평균 기공 크기는 3~5μm, 기공율은 55%였다. 소결체의 XRD 분석결과 2차성은 생성되지 않았으며, SEM 및 dot-Mapping image를 통해 Ni matrix 안에 W 입자가 미세하고 균일하게 분포되어 있어 고용강화 및 분산강화를 통해 Ni 음극의 기계적 특성을 향상시킬 것으로 기대된다.

In the development of Molten Carbonate Fuel Cell, one of the serious problems is the dissolution of cathode material. Therefore, the development of the alternative cathode which is stable in molten carbonate is needed. In this research, the LiCoO2 was chosen as alternative cathode material. LiCoO2 powder was synthesized by high temperature calcination method and by citrate sol-gel method. And its structure and physical characteristics were analyzed by XRD, IR TGA and porosimeter. The conductivity and solubility of LiCoO2 electrode were also measured Homogeneous LiCoO2 powder was obtained by citrate sol-gel method at 445℃, however, obtained above 750℃ by high temperature calcination method. Homogeneous particle size distribution and fine powder were obtained by the citrate sol-gel method. LiCoO2 electrode showed higher electric conductivity (1.7 Ω^-1 ㎝^1) than NiO (0.1 Ω^-1 ㎝^-1) at 650℃. The solubilities of LiCoO2 electrode in electrolyte were varies 0.6 to 1.0 ppm during 200 hours. So, the solubilities of LiCoO2 were much lower than that of NiO.