친환경적인 전기자동차, 전기 추진 선박, 하이브리드 자동차, 전철 등의 구성 요소 중 기존 파워 디바이스에서 사용 중인 실리 콘(Si)을 실리콘 카바이드(SiC, silicon carbide)로 대체하려는 연구가 진행 중이다. 고품질의 SiC 결정 성장을 시키기 위해 다양한 방법 중 상 부 종자 용액 성장(top seeded solution growth, TSSG)법이 큰 주목을 받고 있다. 그러나 SiC 결정 성장 시, 느린 성장 속도뿐만 아니라 많은 결함을 갖는 문제를 갖고 있다. 그래서 본 연구에서는 SiC 단결정을 성장 시키는 TSSG법의 개선을 위한 기초 연구를 진행하였다. 기존에 많이 사용되는 Si, Si0.6Cr0.4 용융 물질와 탄소 도가니와 관계를 가열 온도에 따른 접촉각과 자연 냉각 후 시료의 단면의 차이점을 통해 비 교 분석하였다. 젖음성 분석 시험 장비를 이용하여 탄소 도가니로 쓰이는 카본판 위에 Si과 Si0.6Cr0.4를 놓고 가열 및 용융 시키며 접촉각의 변화를 측정하였고, 가열 종료 후 자연 냉각된 시료의 단면을 관찰하였다. 결과적으로 1800 ℃에서 Si, Si0.6Cr0.4와 탄소판 간의 접촉각이 10°정도 차이를 나타냈다. 단면 관찰에서는 Si의 경우, 탄소판 안으로 스며든 후 굳은 모습을 확인할 수 있었다. 반면, Si0.6Cr0.4의 경우는 탄 소판 안으로 스며든 범위가 훨씬 더 적게 나타냈다. 본 연구의 결과는 TSSG법을 활용한 SiC 단결정 성장을 위한 연구의 기초 자료로 활용 될 것으로 기대된다.

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.

Liquid Bi pool is a candidate electrode for an electrometallurgical process in the molten LiCl-KCl eutectic to treat the spent nuclear fuels from nuclear power plants. The electrochemical behavior of Bi3+ ions and the electrode reaction on liquid Bi pool were investigated with the cyclic voltammetry in an environment with or without BiCl3 in the molten LiCl-KCl eutectic. Experimental results showed that two redox reactions of Bi3+ on inert W electrode and the shift of cathodic peak potentials of Li+ and Bi3+ on liquid Bi pool electrode in molten LiCl-KCl eutectic. It is confirmed that the redox reaction of lithium with respect to the liquid Bi pool electrode would occur in a wide range of potentials in molten LiCl-KCl eutectic. The obtained data will be used to design the electrometallurgical process for treating actinide and lanthanide from the spent nuclear fuels and to understand the electrochemical reactions of actinide and lanthanide at liquid Bi pool electrode in the molten LiCl-KCl eutectic.

As an approach for estimation of the droplet size in the molten salt-liquid metal extraction process, a droplet formation experiment at room temperature was conducted to evaluate the applicability of the Scheele-Meister model with water-mercury system as a surrogate that is similar to the molten salt-liquid metal system. In the experiment, droplets were formed through the nozzle and the droplet size was measured using a digital camera and image analysis software. As nozzles, commercially available needles with inner diameters (ID) of 0.018 cm and 0.025 cm and self-fabricated nozzles with 3-holes (ID: 0.0135 cm), 4-holes (ID: 0.0135 cm), and 2-holes (ID: 0.0148 cm) were used. The mercury penetration lengths in the nozzles were 1.3 cm for the needles and 0.5 cm for the self-fabricated nozzles. The droplets formed from each nozzle maintained stable spherical shape up to 20 cm below the nozzle. The droplet size measurements were within a 10% error range when compared to the Scheele-Meister model estimates. The experimental results show th

본 연구에서는 LiCI-KCl/Cd계의 전해제련 공정을 대상으로 악티늄 및 희토류족 원소들의 전해이동을 모텔링하고 해석하였다. 이 공정에서 용융염 전해질과 액체 카드륨 음극간의 확산 경계층 계면에서 확산제한 전기화학반웅 및 물질수지를 고려한 단순화된 통적모댐을 수립하였다 제안된 모델링 접근방법은 옴극에서 일어나는 금속엽의 반쪽 전 지 환원반용에 기초릎 둔 모델이다 이 모탤올 사용하여 정전류 전해공정에서 주어진 인가전류 조건을 만족하는 시간까지의 전해 이 동과 연계된 농도거동， 각 원소의 패러데이 전류 그리고 시간 함수의 전기화학 전위를 예측하는 가놓성을 보여주었다 선택된 5성분 원소(U ， Pu, Am, La, Nd) 계의 결과를 예비 모사하여 전산모댈이 전기화학적 특성을 이해하고 개선된 전해 제련로를 개발하기 위한 정보를 제공할 수 있는가를 평가하였다.