• 방사성폐기물학회지
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• 제15권2호
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• pp.117-124
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• 2017

Densities of molten salt mixtures of eutectic LiCl-KCl with $UCl_3$, $CeCl_3$, or $LaCl_3$ at various concentrations (up to 13 wt%) were measured using a liquid surface displacement probe. Linear relationships between the mixture density and the concentration of the added salt were observed. For $LaCl_3$ and $CeCl_3$, the measured densities were significantly higher than those previously reported from Archimedes' method. In the case of $LiCl-KCl-UCl_3$, the data fit the ideal mixture density model very well. For the other salts, the measured densities exceeded the ideal model prediction by about 2%.

• Nuclear Engineering and Technology
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• 제42권1호
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• pp.73-78
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• 2010

The suppression of molten salt vaporization is one of the key technical issues in the electrolytic reduction process developed for recycling spent nuclear fuel from light-water reactors Since the Hertz-Langmuir relation previously applied to molten salt vaporization is valid only for vaporization into a vacuum, a diffusion model was derived to quantitatively assess the vaporization of LiCl, $Li_2O$ and Li from an electrolytic reducer operating under atmospheric pressure. Vaporization rates as a function of operation variables were calculated and shown to be in reasonable agreement with the experimental data obtained from thermogravimetry.

• 방사성폐기물학회지
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• 제20권3호
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• pp.259-268
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• 2022

Molten salt consisting primarily of eutectic LiCl-KCl is currently being used in electrorefiners in the Fuel Conditioning Facility at Idaho National Laboratory. Options are currently being evaluated for storing this salt outside of the argon atmosphere hot cell. The hygroscopic nature of eutectic LiCl-KCl makes is susceptible to deliquescence in air followed by extreme corrosion of metallic cannisters. In this study, the effect of occluding the salt into a zeolite on water sorption/desorption was tested. Two zeolites were investigated: Na-Y and zeolite 4A. Na-Y was ineffective at occluding a high percentage of the salt at either 10 or 20wt% loading. Zeolite-4A was effective at occluding the salt with high efficiency at both loading levels. Weight gain in salt occluded zeolite-4A (SOZ) from water sorption at 20% relative humidity and 40℃ was 17wt% for 10% SOZ and 10wt% for 20% SOZ. In both cases, neither deliquescence nor corrosion occurred over a period of 31 days. After hydration, most of the water could be driven off by heating the hydrated salt occluded zeolite to 530℃. However, some HCl forms during dehydration due to salt hydrolysis. Over a wide range of temperatures (320-700℃) and ramp rates (5, 10, and 20℃ min^-1), HCl formation was no more than 0.6% of the Cl^- in the original salt.

• 방사성폐기물학회지
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• 제18권4호
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• pp.549-562
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• 2020

Deionized water, methanol, and ethanol were investigated for their effectiveness at dissolving LiCl-KCl-UCl3 at 25, 35, and 50℃ using inductively coupled plasma mass spectrometry (ICP-MS) to study the concentration evolution of uranium and mass ratio evolutions of lithium and potassium in these solvents. A visualization experiment of the dissolution of the ternary salt in solvents was performed at 25℃ for 2 min to gain further understanding of the reactions. Aforementioned solvents were evaluated for their performance on removing the adhered ternary salt from uranium dendrites that were electrochemically separated in a molten LiCl-KCl-UCl3 electrolyte (500℃) using scanning electron microscopy with energy dispersive spectroscopy (SEM-EDS). Findings indicate that deionized water is best suited for dissolving the ternary salt and removing adhered salt from electrodeposits. The maximum uranium concentrations detected in deionized water, methanol, and ethanol for the different temperature conditions were 8.33, 5.67, 2.79 μg·L-1 for 25℃, 10.62, 5.73, 2.50 μg·L-1 for 35℃, and 11.55, 6.75, and 4.73 μg·L-1 for 50℃. ICP-MS analysis indicates that ethanol did not take up any KCl during dissolutions investigated. SEM-EDS analysis of ethanol washed uranium dendrites confirmed that KCl was still adhered to the surface. Saturation criteria is also proposed and utilized to approximate the state of saturation of the solvents used in the dissolution trials.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 Proceedings of The 6th korea-china joint workshop on nuclear waste management
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• pp.39-50
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• 2005

Electrolytic reduction of uranium oxide to uranium metal was studied in a $LiCl-Li_{2}O$ molten salt system. The reduction mechanism of the uranium oxide to a uranium metal has been studied by means of a cyclic voltammetry. Effects of the layer thickness of the uranium oxide and the thickness of the MgO on the overpotential of the cathode and the anode were investigated by means of a chronopotentiometry. From the cyclic voltamograms, the decomposition potentials of the metal oxides are the determining factors for the mechanism of the reduction of the uranium oxide in a $LiCl-3\;wt{\%} Li_{2}O$ molten salt and the two mechanisms of the electrolytic reduction were considered with regards to the applied cathode potential. In the chronopotentiograms, the exchange current and the transfer coefficient based on the Tafel behavior were obtained with regard to the layer thickness of the uranium oxide which is loaded into the porous MgO membrane and the thickness of the porous MgO membrane. The maximum allowable currents for the changes of the layer thickness of the uranium oxide and the thickness of the MgO membrane were also obtained from the limiting potential which is the decomposition potential of LiCl.

• 방사성폐기물학회지
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• 제1권1호
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• pp.25-39
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• 2003

본 연구에서는 고온의 LiCl-Ll$_2$O 용융염계에서 우라늄 산화물의 금속전환과 Li$_2$O의 전해반응이 동시에 진행되는 통합 반응 메카니즘을 기초로 한 전기화학적 금속전환기술을 제안하였다. 본 실험에서는 전기화학적 환원반응에 의해 생성된 Li 금속이온이 음극에 전착과 동시에 우라늄 산화물과 반응하여 금속전환율 99 % 이상의 우라늄 감속을 생성하는 통합 반응 메카니즘을 확인할 수 있었다. 또한 전기화학적 금속전환기술의 공정 적용성 평가 일환으로 우라늄 산화물의 금속전환성, 반응 메카니즘 규명, Li$_2$O의 closed recycle rate 및 물질전달 특성 등의 기초 데이터를 확보하였다 향후 전기화학적 금속전환기술은 LiCl-Li 용융염계의 금속전환공정의 반응조건 제한성 해소, 금속전환율 향상 및 공정의 단순화 등의 기술성과 경제성 향상 측면에서 획기적인 방안으로 고려될 수 있을 것으로 판단된다.

• 분석과학
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• 제13권6호
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• pp.712-721
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• 2000

$LiF-BeF_2-ZrF_4$(63-30-7 mol%) 용융염은 상온에서 증류수 1ml당 최고 0.02g까지 용해율 99.9%로 용해되었다. 그리고 $ZrF_4$를 열가수분해시켜 제조된 $ZrO_2$ 산화물을 포함하는 $LiF-BeF_2-ZrF_4$ 용융염에서 $ZrO_2$ 산화물을 증류수로 침출시켜 회수하였다. 회수된 $ZrO_2$ 산화물의 결정모양은 손상되지 않았다.

• 자원리싸이클링
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• 제32권1호
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• pp.42-49
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• 2023

Li-Al-Si를 함유한 유리세라믹 순환자원은 인덕션, 방화유리, 비젼냄비 등 리튬의 전체 소비량 중 14%로 리튬이온전지 다음으로 많이 쓰인다. 따라서 리튬의 수요가 폭발하고 있는 현재 새로운 리튬 자원을 찾아야 하고 이로부터 리튬의 회수 연구가 필요하다. 본 연구는 이러한 맥락하에 Li을 함유한 새로운 순환자원인 Li-Al-Si 유리세라믹으로부터 리튬을 회수하기 위한 연구를 수행하였다. 본 연구에서는 1.5% Li, 9.4% Al, 28.9% Si를 함유한 Li-Al-Si 유리세라믹 중 방화유리를 원료물질로 사용하였다. 방화유리로부터 리튬을 회수하기 위한 공정은 크게 칼슘 염을 투입한 건식 배소 공정과 수침출 공정으로 나뉜다. 325 mesh 이하로 분쇄된 방화유리 시료를 열처리 전과 열처리 후 칼슘 염을 투입하여 침출 실험을 비교 진행하였고 칼슘 염과 Li-Al-Si 유리세라의 투입비율에 따른 침출율, 칼슘 염 배소 온도에 따른 침출 연구도 비교 수행하였다. 수침출 연구에서는 온도, 시간, 고액비, 그리고 연속 침출횟수에 따라 리튬의 침출율 및 회수율을 비교하였다. 그 결과 Li-Al-Si를 함유한 유리세라믹 방화유리는 열처리를 반드시 수행하여 베타 형태의 스포듀민으로 상변화 시켜야 하며 이로부터 CaCO₃ 염을 Li-Al-Si를 함유한 유리세라믹 방화유리와 6:1의 비율로 투입하여 1000℃이상에서 배소한 후 4회 이상 연속 침출하여 리튬의 회수율을 98% 이상 획득하였고 이때 리튬의 농도는 200mg/L였다.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2003년도 가을 학술논문집
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• pp.373-378
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• 2003

$LiCl-Li_2O_2$용융염계에서 용융염 취급장치의 구조재료를 위한 평가의 일환으로 오스테나이트 합금인 Fe-base 및 Ni-base 합금의 부식거동을 분위기온도; 650~$725^{\circ}C$, 부식시간: 24~168h, $Li_2O$농도; 3wt%, 혼합가스농도: Ar-10%$O_2$에서 조사하였다. $LiCl-Li_2O_2$ 용융염계에서 Ni-base 합금이 Fe-base 합금보다 높은 내부식성을 나타내었으며, 또한 Fe-base 합금에서 Fe의 함량이 낮고 Ni의 함량이 높은 경우 부식저항성이 증가하였다. 아울러 Fe-base 합금의 부식생성물은 $Cr_2O_3$, $FeCr_2O_4$ Ni-base 합금에서는 $Cr_2O_3$, $NiFe_2O_4$로 나타났다.

• Nuclear Engineering and Technology
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• 제42권2호
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• pp.183-192
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• 2010

An electrochemical reduction of a mixture of metal oxides was conducted in a LiCl molten salt containing 3 wt% $Li_2O$ at $650^{\circ}C$. The oxide reduction was carried out by applying a current to an electrolysis cell, and the $Li_2O$ concentration was analyzed during each run. The concentration of $Li_2O$ in the electrolyte bulk phase gradually decreases according to Faraday's law due to a slow diffusion of the $O^{2-}$ ions. A hindrance effect of the unreduced metal oxides was observed for the reduction of the uranium oxide. Cs, Sr, and Ba of high heat-load fission products were diffused into and accumulated in the salt phase as predicted with thermodynamic consideration.