• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.1
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• pp.83-90
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• 2017

The pyroprocessing of spent nuclear fuel generates LiCl-KCl eutectic waste salt containing radioactive rare earth nuclides. It is necessary to develop a simple process for the treatment of LiCl-KCl eutectic waste in a hot-cell facility. In this study, capture and solidification of a rare earth nuclide (Nd) in LiCl-KCl eutectic salt using an inorganic composite with a $Li_2O-Al_2O_3-SiO_2-B_2O_3$ system was conducted to simplify the existing separation and solidification process of rare earth nuclides in LiCl-KCl eutectic waste salt from the pyroprocessing of spent nuclear fuel. More than 98wt% of Nd in LiCl-KCl eutectic salt was captured when the mass ratio of the composite was 0.67 over $NdCl_3$ in the eutectic salt. The content of $Nd_2O_3$ in the Nd captured-composite reached about 50wt%, and this composite was directly fabricated into a homogeneous and chemical resistant glass waste in a monolithic form. These results will be utilized in designing a process to simplify the existing separation and solidification process.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.426-430
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• 2003

Fissionable uranium will be separated from long-lived nuclear materials in pyroprocess for transmutation. This study was measured decomposition voltage and deposition rate on cathode of uranium in $LiF-BeF_2$ molten salt by electrowinning. The result of experimental is that decomposition voltage of $UF_4$ and $LiF-BeF_2$ molten salt is -1.4 and -1.5 volt at $500^{\circ}C$ Deposition rate of uranium on cathode increases with increase of uranium concentration in molten salt.

• Korean Journal of Materials Research
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• v.15 no.3
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• pp.155-160
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• 2005

The electrolytic reduction of spent oxide fuel involves the liberation of oxygen in a molten LiCl electrolyte, which results in a chemically aggressive environment that is very corrosive fir typical structural materials. So, it is essential to choose the optimum material f3r the process equipment handling molten salt. In this study, the corrosion behavior of Al-Y coated Haynes 263 in a molten salt of $LiCl-Li_2O$ under oxidation atmosphere was investigated at $650^{\circ}C$ for $72\~168$ hours. The corrosion rate of Al-Y coated Haynes 263 was low while that of bare Haynes 263 was high in a molten salt of $LiCl-Li_2O$. Al-Y coated Haynes 263 improved the corrosion resistance better than bare Haynes 263 alloy. An Al oxide layer acts as a protective film which Prohibits Penetration of oxygen. Corrosion Products were formed $Li(Ni,Co)O_2$ and $LiTiO_2$ on bare Haynes 263, but $LiAlO_2,\;Li_5Fe_5O_8\;and\;LiTiO_2$ on Al-Y coated Haynes 263.

• Korean Journal of Materials Research
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• v.14 no.11
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• pp.813-820
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• 2004

The electrolytic reduction of spent oxide fuel involves the liberation of oxygen in a molten LiCl electrolyte, which is a chemically aggressive environment that is very corrosive for typical structural materials. So, it is essential to choose the optimum material for the process equipment handling molten salt. In this study, corrosion behavior of Haynes 263, 75, and Inconel X-750, 718 in molten salt of $LiCl-Li_{2}O$ under oxidation atmosphere was investigated at $650^{\circ}C\;for\;72\sim360$ hours. At $3\;wt\%\;of\;Li_{2}O$, Haynes 263 alloy showed the highest corrosion resistance among the examined alloys, and up to $8\;wt\%\;of\;Li_{2}O$, Haynes 75 exhibited the highest corrosion resistance. Corrosion products were formed $Li(Ni,Co)O_2,\;LiNiO_2\;and\;LiTiO_2\;and\;Cr_{2}O_3$ on Haynes 263, $Cr_{2}O_3,\;NiFe_{2}O_4,\;LiNiO_2,\;Li_{2}NiFe_{2}O_4,\;Li_{2}Ni_{8}O_10$ and Ni on Haynes 75, $Cr_{2}O_3,\;(Al,Nb,Ti)O_2,\;NiFe_{2}O_4,\;and\;Li_{2}NiFe_{2}O_4$ on Inconel X-750 and $Cr_{2}O_3,\;NiFe_{2}O_4\;and\;CrNbO_4$ on Inconel 718, respectively. Haynes 263 showed local corrosion behavior and Haynes 75, Inconel X-750, 718 showed uniform corrosion behavior.

• 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.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1866-1870
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• 2019

Residual salt separation is an essential step in pyroprocessing because its reaction products, as prepared by electrochemical unit processes, contain frozen residual electrolyte species, which are generally composed of alkali-metal chloride salts (e.g., LiCl, KCl). In this study, a simple technique that utilizes high-temperature gas flux as a driving force to melt and push out the residual salt in the reaction products was developed. This technique is simple as it only requires the use of a heating gun in combination with a gas injection system. Consequently, $LiNO_3-ZrO_2$ and $LiCl-ZrO_2$ mixtures were successfully separated by the high-temperature gas injection (separation efficiency > 93%), thereby demonstrating the viability of this simple technique for residual salt separation.

• Journal of the Korean Electrochemical Society
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• v.18 no.3
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• pp.102-106
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• 2015

The performance of Li-B alloy as anode for molten salt electrolysis was firstly investigated. The crystalline phase of the prepared Li-B alloy was identified as $Li_7B_6$. The potential profile of Li-B alloy anode was monitored during the electrodeposition of $Nd^{3+}$ onto an LCC (liquid cadmium cathode) in molten LiCl-KCl salt at $500^{\circ}C$. The potential of Li-B alloy was increased from -2.0 V to -1.4 V vs. Ag/AgCl by increasing the applied current from 10 to $50mA{\cdot}cm^{-2}$. It was found that not only the anodic dissolution of Li to $Li^+$ but also the dissolution of the atomic lithium ($Li^0$) into the LiCl-KCl eutectic salt was observed, following the concomitant reduction of $Nd^{3+}$ by the $Li^0$ in Li-B alloy. It was expected that the direct reduction could be restrained by maintaining the anode potential higher that the deposition potential of neodymium.

• Korean Journal of Materials Research
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• v.9 no.6
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• pp.556-563
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• 1999

The corrosion behavior of heat-resistant alloys, More 1 and Super 22H in molten salts of LiCl and $LiCl-Li_2$O was investigated in the temperature range of $650~850^{\circ}C$. In a molten salt of LiCl, a dense protective oxide scale of $LiCrO_2$ was formed, following growth of oxide scale with parabolic kinetics. But in a mixed molten salt of LiCl, a dense protective oxide scale of $LiCrO_2$ was formed, following growth of oxide scale with parabolic kinetics. But in a mixed molten salt of $LiCl-Li_2$O, a porous non-protective scale of Li\ulcorner(Cr, Ni, Fe)\ulcornerO$_2$was formed, following growth of oxide scale with linear kinetics. 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. The corrosion behavior of Super 22H alloy was similar to that of More 1 alloy, but Super 22H showed higher corrosion resistance than More 1.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2003.11a
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• pp.215-219
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• 2003

The technology to fix a molten LiCl waste, which would be generated from the process to convert spent fuel to metal, into zeolite and then make a final waste form is doing developed. The XRD results of salt-loaded zeolites with different mixing ratios showed that all zeolites transformed from zeolite A type into Li-A type, or also Sodalite type as a minor phase for some conditions. The optimum LiCl-to-zeolite ratio to bring a minimum free salt was 1.0 when the molten LiCl waste contained Cs and Sr.

• Analytical Science and Technology
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• v.13 no.6
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• pp.759-765
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• 2000

A study on the separation of Zr metal from $LiF-BeF_2-ZrF_4$ (67-27-6 mol%) molten salt was carried out using electrowinning. The decomposition potentials of the $LiF-BeF_2$ (72-28 mol%) and the $LiF-BeF_2-ZrF_4$ (67-27-6 mol%) molten salts were measured to be -1.55 and -1.35 volt, respectively. The Zr separation voltage from the salt were found to be in a range of -1.4 -1.5 volt. As increasing applied current, the deposition of molten salt on a cathode increases but the current efficiency decreases. In addition, the deposition and current efficiency decreases with increasing temperature.