• 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.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.675-682
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• 2013

The sequential separation process, composed of an oxygen sparging process for separating lanthanides and a zone freezing process for separating Group I and II fission products, was evaluated and tested with a surrogate eutectic waste salt generated from pyroprocessing of used metal nuclear fuel. During the oxygen sparging process, the used lanthanide chlorides (Y, Ce, Pr and Nd) were converted into their sat-insoluble precipitates, over 99.5% at $800^{\circ}C$; however, Group I (Cs) and II (Sr) chlorides were not converted but remained within the eutectic salt bed. In the next process, zone freezing, both precipitation of lanthanide precipitates and concentration of Group I/II elements were preformed. The separation efficiency of Cs and Sr increased with a decrease in the crucible moving speed, and there was little effect of crucible moving speed on the separation efficiency of Cs and Sr in the range of a 3.7 - 4.8 mm/hr. When assuming a 60% eutectic salt reuse rate, over 90% separation efficiency of Cs and Sr is possible, but when increasing the eutectic salt reuse rate to 80%, a separation efficiency of about 82 - 86 % for Cs and Sr was estimated.

• Journal of Electrochemical Science and Technology
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• v.11 no.4
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• pp.392-398
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• 2020

Producing solar grade silicon using an inexpensive method is a key factor in lowering silicon solar cell costs; the direct electrochemical reduction of SiO₂ in molten salt is one of the more promising candidates for manufacturing this silicon. In this study, SiO₂ granules were electrochemically reduced in molten CaCl₂ (850℃) using Ag-Si eutectic droplets that catalyze electrochemical reduction and purify the Si product. When Ag is used as the working electrode, the Ag-Si eutectic mixture is formed naturally during SiO₂ reduction. However, since the Ag-Si eutectic droplets are liquid at 850℃, they are easily lost during the reduction process. To minimize the loss of liquid Ag-Si eutectic droplets, a cylindrical graphite container working electrode was introduced and Ag was added separately to the working electrode along with the SiO₂ granules. The graphite container working electrode successfully prevented the loss of the Ag-Si eutectic droplets during reduction. As a result, the Ag-Si eutectic droplets acted as stable catalysts for the electrochemical reduction of SiO₂, thereby producing one-dimensional Si rods through a mechanism similar to that of vapor-liquid-solid growth.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.3
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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.

• 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.

• Journal of Advanced Engineering and Technology
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• v.4 no.4
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• pp.431-436
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• 2011

This study was carried out to reduce the problem during distillation process, which separate U, TRU (TRans Uranium) metal electro deposit, Cd and LiCl-KCl eutectic salt generating from LCC (Liquid Cadmium Cathode) electro winning process. The cadmium recovering apparatus was manufactured to separate for each metal using solid-liquid separation method. The apparatus consists of the first sieve for the separation of U and TRU metal electrodeposit, the second sieve for the separation of LiCl-KCl eutectic salt, cadmium collection basket, and a heating furnace. In addition, the size of each sieve is 2 mm to 3 mm. In this experiment, a metal wire was employed to replace TRU metal electrodeposit and U, which exist actually in a LCC crucible. In the solid state, The LiCl-KCl is separated at 340℃ at which the solid and the liquid of the remaining cadmium and LiCl-KCl eutectic salt coexists in each, after the metal wire separated at 500℃. As a result, it seems that it would be beneficial to set the processing condition in the distillation process with the additional treatment process of cadmium and LiCl-KCl eutectic salt.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.1
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• pp.41-48
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• 2010

Uranium deposits from an electrorefining process contain about 30% salt. In order to recover pure uranium and transform it into an ingot, the salts have to be removed from the uranium deposits. Major process variables for the salt distillation process of the uranium deposits are hold temperature and vacuum pressure. Effects of the variables on the salt removal efficiency were studied in the previous study[1]. By applying the Hertz-Langmuir relation to the salt evaporation of the uranium deposits, the evaporation coefficients were obtained at the various conditions. The operational conditions for achieving above 99% salt removal were deduced. The salt distilled uranium deposits tend to form the eutectic melt with iron, nickel, chromium for structural material of salt evaporator. In this study, we investigated the hold temperature limitation in order to prevent the formation of the eutetic melt between urnaium and other metals. The reactions between the uranium metal and stainless steel were tested at various conditions. And for enhancing the evaporation rate of the salt and the efficient recovery of the distilled salt, the thermal analysis of the salt distiller was conducted by using commercial CFX software. From the thermal analysis, the effect of Ar gas flow on the evaporation of the salt was studied.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1727-1732
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• 2017

Pyroprocessing is an alternative method of reprocessing spent fuel, usually involving the dissolving spent fuel in a molten salt media. The National Nuclear Laboratory designed, built, and commissioned a molten salt dynamics rig to investigate the transfer characteristics of molten lithium chloride-potassium chloride eutectic salt. The efficacy and flow characteristics of a high-temperature centrifugal pump and argon gas lift were obtained for pumping the molten salt at temperatures up to $500^{\circ}C$. The rig design proved suitable on an industrial scale and transfer methods appropriate for use in future molten salt systems. Corrosion within the rig was managed, and melting techniques were optimized to reduce stresses on the rig. The results obtained improve the understanding of molten salt transport dynamics, materials, and engineering design issues and support the industrialization of molten salts pyroprocessing.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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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%.