• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.309-313
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• 2018

In this study, we prepared Li-K-Cd alloy, which meets the requirement of eutectic ratio of Li:K, to maintain the operating temperature of the drawdown process at $500^{\circ}C$ and to achieve the reuse of LiCl-KCl molten salt. The prepared Li-K-Cd alloys were added to LiCl-KCl salt bearing U and Nd at $500^{\circ}C$ to investigate the removal of $UCl_3$ in the salt. The reduction of $UCl_3$ in the salt was examined by measuring the OCP value of salt and analyzing the salt composition by ICP-OES. Reduction was also visually confirmed by change of salt color from dark purple to white. The experimental results reveal that the prepared Li-K-Cd alloy has reductive extractability for $UCl_3$ in salt. By improving the preparation method, the Li-K-Cd alloy can be applied to the drawdown process.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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• pp.176-181
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• 2005

The electrolytic reduction process and the electrorefining process, which are being developed at the Korea Atomic Energy Research Institute (KAERI), are to generate molten waste salts such as LiCI salt and LiCI-KCI eutectic salt, respectively. Our goal in waste salt management is to minimize a total waste generation and fabricate a very low­leaching waste form such as a ceramic waste form. Zeolite has been known to one of the most desirable media to immobilize waste salt, which is water soluble and easily radiolyzed. Zeolite can be also used to the removal of fission products from the spent waste salt. Molten LiCI salt is mixed with zeolite A at $650^{\circ}C$ to form a salt-loaded zeolite, and then thermally treated in above $900^{\circ}C$ to become an immobilized product with crystal phase of $Li_{8}Cl_{2}$-Sodalite. In this work, a crystal phase changes of immobilization medium, zeolite, during immobilization of molten LiCI salt using zeolite A is introduced.

• Korean Journal of Materials Research
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• v.9 no.10
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• pp.985-991
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• 1999

Corrosion behaviors of Incoloy 800H, KSA(Kaeri Superalloy)-6, Inconel 600 and Hastelloy C-276 in molten salts were investigated in the temperature range of 650 ~ $850^{\circ}C$. Due to $\textrm{Li}_{2}\textrm{O}$-induced basic fluxing mechanism, the corrosion rates of the alloys in mixed molten salt of LiC1-$\textrm{Li}_{2}\textrm{O}$ were significantly higher than those in molten salt of LiCl. In the mixed molten salt, Fe-base alloys showed higher corrosion resistance than the Ni-base alloys. and Hastelloy C-276 with high Mo and W contents exhibited the highest corrosion rate among the examined alloys. The single layer of $\textrm{LiCrO}_{2}$ was formed in molten salt of LiCl and two phase structure of a scale consisted of oxides and Ni was formed in the mixed molten salt.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3957-3961
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• 2022

Deposition behaviors of Sr and Cs in various liquid cathodes, such as Zn, Bi, Cd, and Pb, were examined to evaluate their recovery from LiCl-KCl eutectic salt. Cations in the salt were deposited on the liquid cathode, exhibiting potential of -1.8 to -2.1 V (vs. Ag/AgCl). Zn cathode had successful deposition of Sr and exhibited the highest recovery efficiency, up to 55%. Meanwhile, the other liquid cathodes showed low current efficiencies, below 18%, indicating LiCl-KCl salt decomposition. Sr was recovered from the Zn cathode as irregular rectangular SrZn₁₃ particles. A negligible amount of Cs was deposited on the entire liquid cathode, indicating that Cs was hardly deposited on liquid cathodes. Based on these results, we propose that liquid Zn cathode can be used for cleaning Sr in LiCl-KCl salt.

• Korean Journal of Materials Research
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• v.9 no.2
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• pp.211-216
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• 1999

Corrosion behavior of austenitic stainless steels of SUS 316L and SUS304L in molten salt of LiCl and $LiCl/Li_2O$ has been investigated in the temperature range of $650~850^{\circ}C$. Corrosion products of SUS316L and 304L in hot molten salt consisted of two layers-an outer layer of Li(CrFe)$O_2$and an inner layer of$Cr_2O_3$. The corrosion layer was uniform in molten salt of LiCl, but the intergranular corrosion occurred in addition to the uniform corrosion in mixed molten salt of LiCl/$Li_2O$. The corrosion rate increased slowly with the increase of temperature up to $750^{\circ}C$, but above $750^{\circ}C$ rapid increase in corrosion rate observed. SUS316L stainless steel showed slower corrosion rate than SUS 304L, exhibiting higher corrosion resistance in the molten salt.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.2 no.3
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• pp.211-217
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• 2004

A Li recovery technology has been developed and related experimental verification efforts were carried out to improve the economical viability and environmental friendliness of the 'Advanced Spent Fuel Conditioning Process' being developed at KAERI. This technology is characterized by the combination of 1) the electrolysis of $Li_2O$ in a molten salt by using a porous non-conducting magnesia container at the cathode, 2) the separation of the Li in the container from the molten salt by elevating the container above the level of a molten salt, 3) the transport of the Li in the container by using a vacuum siphon to a separated reservoir. Li was semi-continuously recovered from a LiCl-$Li_2O$ molten salt with a more than 95% yield by using the developed technology.

• Korean Journal of Materials Research
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• v.23 no.2
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• pp.123-127
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• 2013

The electrolytic reduction of a spent oxide fuel involves liberation of the oxygen in a molten LiCl electrolyte, which is a chemically aggressive environment that is too crosive for typical structural materials. Therefore, it is essential to choose the optimum material for the process equipment for handling a molten salt. In this study, the corrosion behavior of pyro-carbon made by CVD was investigated in a molten LiCl-$Li_2O$ salt under an oxidation atmosphere at $650^{\circ}C$ and $750^{\circ}C$ for 72 hours. Pyro-carbon showed no chemical reactions with the molten salt because of its low wettability between pyro-carbon and the molten salt. As a result of XRD analysis, pyro-carbon exposed to the molten salt showed pure graphite after corrosion tests. As a result of TGA, whereas the coated layer by CVD showed high anti-oxidation, the non-coated layer showed relatively low anti-oxidation. The stable phases in the reactions were $C_{(S)}$, $Li_2CO_{3(S)}$, $LiCl_{(l)}$, $Li_2O$ at $650^{\circ}C$ and $C_{(S)}$, $LiCl_{(l)}$, $Li_2O_{(S)}$ at $750^{\circ}C$. $Li_2CO_{(S)}$ was decomposed at $750^{\circ}C$ into $Li_2O_{(S)}$ and $CO_{2(g)}$.

• Journal of the Korean Ceramic Society
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• v.46 no.6
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• pp.587-591
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• 2009

Lithium salt have been used mainly as electrolyte of thermal battery for electricity storage. Recently, The 3phase lithium salt(LiCl-LiF-LiBr) is tried to use as electrolyte of thermal battery for high electric power. It is reported that LiCl-LiF-LiBr salt have high ion mobility due to its high lithium ion concentration. Solid lithium salt is melt to liquid state at above $500{^{\circ}C}$. The lithium ion is easily reacted with support materials. Because the melted lithium ion has small ion size and high ion mobility. For the increasing mechanical strength of electrolyte pellet, the research was started to apply ceramic filter to support of electrolyte. In this study, authors used SiOC web and glass fiber filter as ceramic mat for support of electrolyte and impregnated LiCl-LiF-LiBr salt into ceramic mat at above $500{^{\circ}C}$. The fabricated electrolyte using ceramic mat was washed with distilled water for removing lithium salt on ceramic mat. The washed ceramic mat was observed for lithium ion reaction behavior with XRD, SEM-EDS and so on.

• Journal of Electrochemical Science and Technology
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• v.13 no.1
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• pp.78-89
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• 2022

ANODE-free Li-metal batteries (AFLMBs) operating with Li of cathode material have attracted enormous attention due to their exceptional energy density originating from anode-free structure in the confined cell volume. However, uncontrolled dendritic growth of lithium on a copper current collector can limit its practical application as it causes fatal issues for stable cycling such as dead Li formation, unstable solid electrolyte interphase, electrolyte exhaustion, and internal short-circuit. To overcome this limitation, here, we report a novel dual-salt electrolyte comprising of 0.2 M LiPF₆ + 3.8 M lithium bis(fluorosulfonyl)imide in a carbonate/ester co-solvent with 5 wt% fluoroethylene carbonate, 2 wt% vinylene carbonate, and 0.2 wt% LiNO₃ additives. Because the dual-salt electrolyte facilitates uniform/dense Li deposition on the current collector and can form robust/ionic conductive LiF-based SEI layer on the deposited Li, a Li/Li symmetrical cell exhibits improved cycling performance and low polarization for over 200 h operation. Furthermore, the anode-free LiFePO₄/Cu cells in the carbonate electrolyte shows significantly enhanced cycling stability compared to the counterparts consisting of different salt ratios. This study shows an importance of electrolyte design guiding uniform Li deposition and forming stable SEI layer for AFLMBs.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.3
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• pp.373-382
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• 2020

This study proposes a method of separating uranium (U) and minor actinides from rare earth (RE) elements in the LiCl-KCl salt system. Several RE metals were used to reduce UCl₃ and MgCl₂ from the eutectic LiCl-KCl salt systems. Five experiments were performed on drawdown U and plutonium (Pu) surrogate elements from RECl₃-enriched LiCl-KCl salt systems at 773 K. Via the introduction of RE metals into the salt system, it was observed that the UCl₃ concentration can be lowered below 100 ppm. In addition, UCl₃ was reduced into a powdery form that easily settled at the bottom and was successfully collected by a salt distillation operation. When the RE metals come into contact with a metallic structure, a galvanic interaction occurs dominantly, seemingly accelerating the U recovery reaction. These results elucidate the development of an effective and simple process that selectively removes actinides from electrorefining salt, thus contributing to the minimization of the influx of actinides into the nuclear fuel waste stream.