• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1711-1716
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• 2017

Uranium tetrafluoride ($UF_4$) is the most used nuclear material for producing metallic uranium by reduction with Ca or Mg. Metallic uranium is a raw material for the manufacture of uranium silicide, $U_3Si_2$, which is the most suitable uranium compound for use as nuclear fuel for research reactors. By contrast, ammonium uranyl carbonate is a traditional uranium compound used for manufacturing uranium dioxide $UO_2$ fuel for nuclear power reactors or $U_3O_8-Al$ dispersion fuel for nuclear research reactors. This work describes a procedure for recovering uranium and ammonium fluoride ($NH_4F$) from a liquid residue generated during the production routine of ammonium uranyl carbonate, ending with $UF_4$ as a final product. The residue, consisting of a solution containing high concentrations of ammonium ($NH_4^+$), fluoride ($F^-$), and carbonate ($CO_3^{2-}$), has significant concentrations of uranium as $UO_2^{2+}$. From this residue, the proposed procedure consists of precipitating ammonium peroxide fluorouranate (APOFU) and $NH_4F$, while recovering the major part of uranium. Further, the remaining solution is concentrated by heating, and ammonium bifluoride ($NH_4HF_2$) is precipitated. As a final step, $NH_4HF_2$ is added to $UO_2$, inducing fluoridation and decomposition, resulting in $UF_4$ with adequate properties for metallic uranium manufacture.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.237.1-237
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• 1999

• 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.

• Economic and Environmental Geology
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• v.56 no.5
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• pp.603-618
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• 2023

This study focused on evaluating the suitability of the WRK (waste repository Korea) bentonite as a buffer material in the SNF (spent nuclear fuel) repository. The U (uranium) adsorption/desorption characteristics and the adsorption mechanisms of the WRK bentonite were presented through various analyses, adsorption/desorption experiments, and kinetic adsorption modeling at various pH conditions. Mineralogical and structural analyses supported that the major mineral of the WRK bentonite is the Ca-montmorillonite having the great possibility for the U adsorption. From results of the U adsorption/desorption experiments (intial U concentration: 1 mg/L) for the WRK bentonite, despite the low ratio of the WRK bentonite/U (2 g/L), high U adsorption efficiency (>74%) and low U desorption rate (<14%) were acquired at pH 5, 6, 10, and 11 in solution, supporting that the WRK bentonite can be used as the buffer material preventing the U migration in the SNF repository. Relatively low U adsorption efficiency (<45%) for the WRK bentonite was acquired at pH 3 and 7 because the U exists as various species in solution depending on pH and thus its U adsorption mechanisms are different due to the U speciation. Based on experimental results and previous studies, the main U adsorption mechanisms of the WRK bentonite were understood in viewpoint of the chemical adsorption. At the acid conditions (＜pH 3), the U is apt to adsorb as forms of UO₂²⁺, mainly due to the ionic bond with Si-O or Al-O(OH) present on the WRK bentonite rather than the ion exchange with Ca²⁺ among layers of the WRK bentonite, showing the relatively low U adsorption efficiency. At the alkaline conditions (>pH 7), the U could be adsorbed in the form of anionic U-hydroxy complexes (UO₂(OH)₃^-, UO₂(OH)₄^2-, (UO₂)₃(OH)₇^-, etc.), mainly by bonding with oxygen (O^-) from Si-O or Al-O(OH) on the WRK bentonite or by co-precipitation in the form of hydroxide, showing the high U adsorption. At pH 7, the relatively low U adsorption efficiency (42%) was acquired in this study and it was due to the existence of the U-carbonates in solution, having relatively high solubility than other U species. The U adsorption efficiency of the WRK bentonite can be increased by maintaining a neutral or highly alkaline condition because of the formation of U-hydroxyl complexes rather than the uranyl ion (UO₂²⁺) in solution,and by restraining the formation of U-carbonate complexes in solution.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.2
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• pp.77-84
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• 2005

A voloxidizer for a hot cell demonstration, that handles spent fuels of a high radiation level in a limited space should be small and spent fuel powders should not be dispersed out of the equipment involved. In this study a density rate equation as well as the Stokes'equation has been proposed in order to obtain the theoretical terminal velocity of powders. The terminal velocity of U$_{3}$O$_{8}$ has been predicted by using the terminal velocity of SiO$_{2}$, and then determination has been the optimum air flow rate which is able to prevent powders from scattering. An equation which has shown a relationship between theoretical terminal velocities of U$_{3}$O$_{8}$ and SiO$_{2}$ has been derived with the help of the Stokes'equation, and then an experimental verification made for the theoretical Stokes' equation of SiO$_{2}$ by means of an experimental device made of acryl. The theoretical terminal velocity based on the proposed density rate equation has been verified by detecting U$_{3}$O$_{8}$ powders in a filter installed in the mock-up voloxidizer. As the results, the optimum air flow rates seem to be 20 LPM by the Stokes'equation while they are 14.5 L/min by the density rate equation. At the experiments with the mock-up voloxidizer, a trace amount of U$_{3}$O$_{8}$ seems to be detectable at the air flow rate of 14.5 L/min by the density rate equation, but U$_{3}$O$_{8}$ powders of 7$\mu$m diameter seem detectable at the air flow rate of 20 L/min by the Stokes'equation. It is revealed that 14.5 L/min is the optimum air flowe rate which is capable of preventing U$_{3}$O$_{8}$ powders from scattering in the UO$_{2}$ voloxidizer and the proposed density rate equation is proper to calculate the terminal velocity of U$_{3}$O$_{8}$ powders.

• Proceedings of the Korean Powder Metallurgy Institute Conference
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• 2004.11a
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• pp.3-4
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.268-268
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• 2002

• Nuclear Engineering and Technology
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• v.28 no.2
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• pp.150-161
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• 1996

A deformation model was developed to calculate the deformation of the uranium-silicide dispersion fuel (U$_3$Si-Al) elements for research reactors. The model was based on the elasto-plasticity theory and power-law creep theory. Also, isotopic swelling was assumed for the fuel meat and isotropic thermal expansion for the fuel meat and dadding. The new model calculated successfully the deformation of the fuels of HANARO and NRU (in Canada). As the most important result, it was shown that the primary deformation mechanism in the fuel meat was swelling and that in the cladding was creep. For all cases simulated, the maximum hoop stress at cladding outer surface was lass than 5MPa, probably well below the yield stress of the dadding, and finally, the volume change was predicted to be less than 10% in the whole burnup range.

• Korean Chemical Engineering Research
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• v.57 no.6
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• pp.874-882
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• 2019

In this study, the gasification characteristics of four types of unused woody biomass and one waste wood in a lab-scale bubbling fluidized bed gasifier (Diameter: 0.11 m, Height: 0.42 m) were investigated. Effect of equivalence ratio (ER) of 0.15-0.3 and gas velocity of $2.5-5U_0/U_{mf}$ are determined at the constant temperature of $800^{\circ}C$ and fuel feeding rate of 1 kg/h. The silica sand particle having an average particle size of $287{\mu}m$ and olivine with an average particle size of $500{\mu}m$ were used as the bed material, respectively. The average product gas composition of samples is as follows; $H_2$ 3-4 vol.%, CO 15-16 vol.%, $CH_4$ 4 vol.% and $CO_2$ 18-19 vol.% with a lower heating value (LHV) of $1193-1301kcal/Nm^3$ and higher heating value (HHV) of $1262-1377kcal/Nm^3$. In addition, it was found that olivine reduced most of C2 components and increased $H_2$ content compared to silica sand, resulting in cracking reaction of tar. The non-condensable tar decreases by 72% ($1.24{\rightarrow}0.35g/Nm^3$) and the condensable tar decreases by 27% ($4.4{\rightarrow}3.2g/Nm^3$).

• Nuclear Engineering and Technology
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• v.33 no.4
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• pp.386-396
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• 2001

Mixtures Of AUC-UO$_2$and Gd$_2$O$_3$ Powders doped With Cr$_2$O$_3$ or Cr$_2$O$_3$-SiO$_2$ were Pressed and sintered at 1730 t in hydrogen gas witk various water-vapor contents. The density of UO$_2$- 6wt% Gd$_2$O$_3$ pellets can be increased from 91% TD to 94.5% TD in 1 vol% $H_2O$-H$_2$ gases by the addition of 0.02wt% Cr$_2$O$_3$-(0.01~0.04) wt% SiO$_2$. The magnitude of density increase is much larger in (1~3 vol%) $H_2O$-H$_2$ gases than in 0.05 vol% $H_2O$-H$_2$ gas. The densification of U0$_2$- Gd$_2$O$_3$ compact is significantly delayed in the temperature range between 1300 and 1500 t , but that of compacts with Cr$_2$O$_3$-SiO$_2$ is not. The role of Cr$_2$O$_3$ and SiO$_2$ in densification is discussed.