• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1534-1539
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• 2021

This study reports on the dissolution behavior of SrO in LiCl at varying SrO concentrations from low concentrations to excess. The amount of SrO dissolved in the molten salt and the species present upon cooling were determined. The thermal behavior of LiCl containing various concentrations of SrO was investigated. The experimental results were compared with results from the simulated results using the HSC Chemistry software package. Although the reaction of SrO with LiCl in the standard state at 650 ℃ has a slightly positive Gibbs free energy, SrO was found to be highly soluble in LiCl. Experimentally determined SrO concentrations were found to be considerably higher than those present in used nuclear fuel (<2 g/kg). As Sr-90 is one of the most important heat-generating nuclides in used nuclear fuel, this finding will be impactful in the development of fast, simple, and proliferation-resistant heat reduction processes for used nuclear fuel without the need for separating nuclear materials. Heat reduction is important as it decreases both the volume necessary for final disposal and the worker handling risk.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.4
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• pp.207-212
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• 2009

In this study, a thermal behavior of $PrCl_3$ as one of the lanthanide chlorides in LiCl-KCl molten salts was investigated in an oxidizing condition. First, a thermo-gravimetric analysis (TGA) of $PrCl_3$ was carried out by an injection of $O_2$ gas. Based on the results, an oxidation of $PrCl_3$ in the molten salts was performed by sparging $O_2$ gas with changing temperatures. According to the TGA data of $PrCl_3$, a dissociation of $PrCl_3$ occurred rapidly by about $380^{\circ}C$ and a conversion of $PrCl_3$ to $PrCl_3$ was completed at about $600^{\circ}C$. The thermal behavior of $PrCl_3$ in LiCl-KCl molten salts by sparging $O_2$ gas was similar to that of $PrCl_3$ in the TGA test, and PrOCl as a insoluble compound in the molten salts was precipitated into the bottom of the molten salts. A conversion of $PrCl_3$ to PrOCl in the molten salts occurred actively at a higher temperature than $650^{\circ}C$. And it would be possible to estimate a conversion status of $PrCl_3$ to PrOCl by measuring a $Cl_2$ concentration in a flue gas generated from an oxidation test of $PrCl_3$ in LiCl-KCl molten salts.

• Journal of the Korean Ceramic Society
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• v.38 no.2
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• pp.143-151
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• 2001

Ca/LiCl-KCl/CaCrO$_4$열 전지계의 양극재료로서 BCT(Body-Centered Tetragonal) 결정구조를 갖는 CaCrO$_4$분말을 GNP로 합성하고, SEM, TEM, XRD를 이용하여 그 미세구조를 분석하였다. GNP 공정에 의한 CaCrO$_4$분말은 단일상으로 0.5$mu extrm{m}$ 이하의 입자 크기를 가지며 균일하게 분포한 반면, 기존의 분말 혼합법은 높은 하수 온도 및 장시간의 하소 조건을 필요하므로 미세한 분말 합성이 어렵고 pellet 형태로 만들었을 때 GNP 분말에 비해 비표면적이 현저하게 작기 때문에 전극 재료로써 유리하지 못하다. Ca/LiCl-KCl/CaCrO$_4$계의 전기 화학적인 특성을 평가해본 결과 전지셀을 Ca/DEB(LiCl-KCl+CaCrO$_4$+SiO$_2$)와 같은 DEB 형태로 만들었을 때 $600^{\circ}C$의 온도에서 2.0 V이상 (<100 mA/㎤)의 안정한 전압이 5분 이상 유지되었다. 그러나 3층 전극 셀(Ca/LiCl/KCl/ CaCrO$_4$)에서는 동일한 온도에서 2.0 V이상 (<100 mA/㎤)의 전압이 7분 이상 유지되었으나 불안정한 전압 변동 및 낮은 peak voltage로 인해 DEB 셀의 전지 특성이 더 우수한 것으로 생각된다. 양극 재료의 제조 방법의 관점에서 볼 때, 동일한 DEB(Depolarizer : Electrolyte : Binder=25 : 70 : 5 wt%) 조성의 셀 구성시, GNP 분말은 분말 혼합법에 의한 분말보다 반응 표면적이 훨씬 크기 때문에 GNP 양극 활 물질의 DEB 셀에서의 전지 수명이 더 길었다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.2
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• pp.105-112
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• 2005

The electrolytic reduction of uranium oxide in a LiCl-Li$_{2}$O molten salt system has been studied in a 10 g U$_{3}$O$_{8}$ /batch-scale experimental apparatus with an integrated cathode assembly at 650$^{\circ}C$. The integrated cathode assembly consists of an electric conductor, the uranium oxide to be reduced and the membrane for loading the uranium oxide. From the cyclic voltammograms for the LiCl-3 wt$\%$ Li$_{2}$O system and the U$_{3}$O$_{8}$-LiCl-3 wt$\%$ Li$_{2}$O system according to the materials of the membrane in the cathode assembly, the mechanisms of the predominant reduction reactions in the electrolytic reactor cell were to be understood; direct and indirect electrolytic reduction of uranium oxide. Direct and indirect electrolytic reductions have been performed with the integrated cathode assembly. Using the 325-mesh stainless steel screen the uranium oxide failed to be reduced to uranium metal by a direct and indirect electrolytic reduction because of a low current efficiency and with the porous magnesia membrane the uranium oxide was reduced successfully to uranium metal by an indirect electrolytic reduction because of a high current efficiency.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.3
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• pp.181-186
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• 2015

It is necessary to develop an effective waste salt treatment technology for the minimization of radioactive waste generation from the pyroprocessing of spent nuclear fuel. For this reason, the separation characteristics of NdCl₃ from LiCl-KCl eutectic salt in a reactive distillation process using Li₂CO₃ or K₂CO₃ were observed. NdCl₃ was converted into oxychloride (NdOCl) or oxide (Nd₂O₃) in the reaction model between NdCl₃ and the carbonates using HSC-Chemistry, and this result was confirmed in the reactive distillation test of the LiCl-KCl-NdCl₃ system using the carbonates. Based on these results, the reactive distillation process conditions were determined to separate NdCl₃ into an oxide form (Nd₂O₃) which can be easily fabricated into a final waste form.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.285-291
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• 2004

As a part of assessment of the structural material for the molten salt handling system, corrosion behavior of Inconel 718, X-750, Haynes 75 and Haynes 263 alloys in the molten salt of LiCl-Li$_2$O-O$_2$was investigated in the range of temperature; $650^{\circ}C$, time; 24~168h, $Li_2O$; 3wt%, mixed gas; Ar~10%$O_2$. In the molten salt of LiCl-$Li_2O-O_2$, the order corrosion rate was Haynes 263 ＜ Haynes 75 ＜ Inconel X-750 ＜ Inconel 718. Haynes 263 alloy showed the highest corrosion resistance among the examined alloys. Corrosion products of alloys were as fellows: Haynes 75: $Cr_2O_4$, $NiFe_2O_4$, $LiNiO_2$, $Li_2NiFe_2O_4$, Inconel 718; $Cr_2O_4$, $NiFe_2O_4$, Haynes 263; $Li(Ni,Co)O_2$, $NiCr_2O_4$, $LiTiO_2$, Inconel X-750; $Cr_2O_3$, $NiFe_2O_4$,$FeNi_3$, (Al,Nb,Ti)$O_2$. Haynes 263 showed local corrosion behavior and Haynes 75, Inconel 718 and Inconel X-750 showed uniform corrosion behavior.

• Proceedings of the Materials Research Society of Korea Conference
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• 2003.11a
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• pp.51-51
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• 2003

• Korean Chemical Engineering Research
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• v.50 no.3
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• pp.427-434
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• 2012

Waterborne polyurethane dispersions (PUD) were synthesized from poly (carbonate diol), isophrone diisocyanate and dimethylol propionic acid at different NCO/OH molar ratios. Subsequently, the PUD was mixed with different types of alkali metal salts ($LiClO_4$, $NaClO_4$, and $KClO_4$) to prepare antistatic waterborne polyurethane coating solutions. Effects of the types and amounts of alkali metal salts were investigated on the surface resistances of the resulting coating films. The surface resistances of coating films were decreased with increasing the amounts of alkali metal salts added in the PUD. The coating films prepared with the same amount of alkali metal salts showed increased ionic conductivity with the order of $LiClO_4$ > $NaClO_4$ > $KClO_4$. Also, the surface resistances of coating films were increased with increasing the molar ratios of NCO/OH in the PUD.

• Proceedings of the KIEE Conference
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• 1994.07b
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• pp.1229-1232
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• 1994

The purpose of this study is to research and develop solid polymer electrolyte(SPE) for Li secondary battery. This paper describes the effects of lithium salts, plasticizer addition and temperature dependence of conductivity of PEO electrolytes. Polyethylene oxide(PEO) based polymer electrolyte films were prepared by solution casting an acetonitrile solution of preweighed PEO and Li salt. After solvent evaporation, the electrolyte films were vacuum-dried at $60^{\circ}C$ for 48h, the thickness of the films were $90{\sim}110{\mu}m$. The conductivity properties of prepared PEO electrolytes are summarized as follows. PEO electrolyte complexed with $LiClO_4$ shows the better conductivity of the others. $PEO-LiClO_4$ electrolyte when $EO/Li^+$ ratio is 8, showed the best conductivity. Optimum operating temperature of PEO electrolyte is $60^{\circ}C$. By adding propylene carbonate and ethylene carbonate to $PEO-LiClO_4$ electrolyte, its conductivity was higher than $PEO-LiClO_4$ without those. Also $PEO_8LiClO_4$ electrolyte remains static up to 4.5V vs. $Li/Li^+$.