• 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 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.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 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.

• 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.

• 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%.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.4
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• pp.325-332
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• 2013

Pyroporcessing of spent nuclear fuel generates a considerable amount of LiCl-KCl eutectic waste salt containing radioactive rare earth (RE) chlorides. In this study, a series of processes, which consist of a phosphorylation/distillation process and a solidification process, were performed to minimize volume of the LiCl-KCl eutectic waste salt and solidify a residual waste into a stable form at a relatively low temperature. Over 99wt% of RE chlorides in LiCl-KCl eutectic salt was converted and separated into $REPO_4$ in the phosphorylation/distillation process using a mixture of $Li_3PO_4-K_3PO_4$. The separated $REPO_4$ was solidified into a homogeneous and fine-grained form at $1,050^{\circ}C$ using LIP(Lead Iron Phosphate) as a solidification agent. The final waste volume was reduced below about 10% through the series of the processes.

• 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.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.05a
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• pp.102-103
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• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.600-601
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• 2018