• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.3
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• pp.189-197
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• 2007

The solubility of U(VI) hydrolysis products was determined by using a laser-induced breakdown detection (LIBD) technique. The experiments were carried out at uranium concentrations in range from $2{\times}10^{-4}\;M\;to\;4{\times}10^{-6}\;M$, pH values between 3.8 and 7.0, the constant ionic strength of 0.1 M $NaClO_4$ and the temperature of $25.0{\pm}0.1^{\circ}C$. The solubility product of U(VI) hydrolysis products was calculated from LIBD results by using the hydrolysis constants selected in NEA-TDB. The solubility product extrapolated to zero ionic strength, ${\log}K^{\circ}_{sp}=-22.85{\pm}0.23$ was calculated by using a specific ion interaction theory (SIT). The spectral features of ionic species in uranium solutions were investigated by using a conventional UV-visible absorption spectrophotometer and a fluorophotometer, respectively, $(UO_2)_2(OH)_2^{2+}\;and\;(UO_2)_3(OH)_5^+$ were dominant species at uranium concentration of $2{\times}10^{-4}\;M$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.8 no.4
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• pp.261-267
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• 2010

The pyrometallurgical nuclear fuel recycle process, called pyroprocessing, has been known as a promising nuclear fuel recycling technology. Pyroprocessing technology is crucial to advanced nuclear systems due to increased nuclear proliferation resistance and economic efficiency. The basic concept of pyroprocessing is group actinide recovery, which enhances the nuclear proliferation resistance significantly. One of the key steps in pyroprocessing is "electrowinning" which recovers group actinides with lanthanide from the spent nuclear fuels. In this study, a vertical cadmium distiller was manufactured. The evaporation rate of pure cadmium in vertical cadmium distiller varied from 12.3 to $40.8g/cm^2/h$ within a temperature range of 773 923 K and pressure below 0.01 torr. Uranium - cadmium alloy was fabricated by electrolysis using liquid cadmium cathode in a high purity argon atmosphere glove box. The distillation behavior of pure cadmium and cadmium in uranium - cadmium alloy was investigated. The distillation behavior of cadmium from this study could be used to develop an actinide recovery process from a liquid cadmium cathode in a cadmium distiller.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.321-332
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• 2017

In order to build a general model of a high-throughput uranium electrorefining process according to the electrode configuration, numerical analysis was conducted using the COMSOL Multiphysics V5.3 electrodeposition module with Ordinary Differential Equation (ODE) interfaces. The generated model was validated by comparing a current density-potential curve according to the distance between the anode and cathode and the electrode array, using a lab-scale (1kg U/day) multi-electrode electrorefiner made by the Korea Atomic Energy Research Institute (KAERI). The operating temperature was $500^{\circ}C$ and LiCl-KCl eutectic with 3.5wt% $UCl_3$ was used for molten salt. The efficiency of the uranium electrorefining apparatus was improved by lowering the cell potential as the distance between the electrodes decreased and the anode/cathode area ratio increased. This approach will be useful for constructing database for safety design of high throughput spent nuclear fuel electrorefiners.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.3
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• pp.161-165
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• 2009

To evaluate the efficiency of the decontamination plant for the removal of uranium compounds deposited on the internal surface of $UF_6$ cylinder for its reuse, two demonstration tests of the plant with different ratio of ${Na_2}{CO_3}$ and ${H_2}{O_2}$ were carried out, and each test had 5 steps. The main chemical form removed by the tests was to be identified as ${Na_4}{UO_2}(CO_3)_3$. More than 50% of uranium was removed by water of the first step, and at the following steps the removal amounts were exponentially decreased. On the other hand, the result shows that the injected amount of ${Na_2}{CO_3}$, compared with that of the removed uranium, was stoichiometrically excessed. This suggests that the injected amounts of ${Na_2}{CO_3}$, the generation rate of decontaminated waste, and the decontamination steps could be reduced by a process optimization of the plant.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.1
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• pp.1-6
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• 2014

A large amount of acidic waste solution is generated from the practical electrokinetic decontamination equipments for the remediation of soil contaminated with uranium. After filtration of uranium hydroxides formed by adding CaO into the waste solution, the filtrate was recycled in order to reduce the volume of waste solution. However, when the filtrate was used in an electrokinetic equipment, the low permeability of the filtrate from anode cell to cathode cell due to a high concentration of calcium made several problems such as the weakening of a fabric tamis, the corrosion of electric wire and the adhension of metallic oxides to the surface of cathode electrode. To solve these problems, sulfuric acid was added into the filtrate and calcium in the solution was removed as $CaSO_4$ precipitate. A decontamination test using a small electrokinetic equipment for 20 days indicated that Ca-removed waste solution decreased uranium concentration of the waste soil to 0.35 Bq/g, which is a similar to a decontamination result obtained by distilled water.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.1 no.1
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• pp.11-23
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• 2003

A study on the electrolytic dissolution of SUS-304 and Inconel-600 specimen was carried out in neutral salt electrolyte to evaluate the applicability of electrochemical decontamination process for recycle or self disposal with authorization of large amount of metallic wastes contaminated with uranium compounds generated by dismantling a retired uranium conversion plant in Korea. Although the best electrolytic dissolution performance for the specimens was observed in a Na2s04 electrolyte, a NaNO$_3$ neutral salt electrolyte, in which about 30% for SUS-304 and the same for Inconel-600 in the weight loss was shown in comparison with that in a Na$_2$SO$_4$ solution, was selected as an electrolyte for the electrochemical decontamination of metallic wastes with the consideration on the surface of system components contacted with nitric acid and the compatibility with lagoon wastes generated during the facility operation. The effects of current density, electrolytic dissolution time, and concentration of NaNO$_3$ on the electrolytic dissolution of the specimens were investigated. On the basis of the results obtained through the basic inactive experiments, electrochemical decontamination tests using the specimens contaminated with uranium compounds such as UO$_2$, AUC (ammonium uranyl carbonate) and ADU (ammonium diuranate) taken from an uranium conversion facility were performed in 1M NaNO$_3$ solution with the current density or In mA/$\textrm{cm}^2$. it was verified that the electrochemical decontamination of the metallic wastes contaminated uranium compounds was quite successful in a NaNO$_3$ neutral salt electrolyte by reducing $\alpha$ and $\beta$ radioactivities below the level of self disposal within 10 minutes regardless of the type of contaminants and the degree of contamination.

• Nuclear Engineering and Technology
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• v.38 no.4
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• pp.359-374
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• 2006

Since 1991, Korea, Canada and United States have performed the direct use of spent pressurized water reactor (PWR) fuel in the Canada deuterium uranium (CANDU) reactors (DUPIC) fuel development project. Unlike the Tandem fuel cycle, which requires a wet reprocessing, the DUPIC fuel technology can directly refabricate CANDU fuels from the PWR spent fuel and, therefore, is recognized as a highly proliferation-resistant fuel cycle technology, which can be adopted even in non-proliferation treaty countries. The Korea Atomic Energy Research Institute (KAERI) has fabricated DUPIC fuel elements in a laboratory-scale remote fuel fabrication facility. KAERI has demonstrated the fuel performance in the research reactor, and has confirmed the operational feasibility and safety of a CANDU reactor loaded with the DUPIC fuel using conventional design and analysis tools, which will be the foundation of the future practical and commercial uses of DUPIC fuel.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.109-114
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• 2020

This work presents the results of laboratory scale tests of the CARBOFLUOREX (CARBOnate FLUORide EXtraction) process - a novel technology for the recovery of U and Pu from the solid fluorides residue (fluorination ash) of Fluoride Volatility Method (FVM) reprocessing of spent nuclear fuel (SNF). To study the oxidative leaching of U from the fluorination ash (FA) by Na₂CO₃ or Na₂CO₃-H₂O₂ solutions followed by solvent extraction by methyltrioctylammonium carbonate in toluene and purification of U from the fission products (FPs) impurities we used a surrogate of FA consisting of UF₄ or UO₂F₂, and FPs fluorides with stable isotopes of Ce, Zr, Sr, Ba, Cs, Fe, Cr, Ni, La, Nd, Pr, Sm. Purification factors of U from impurities at the solvent extraction refining stage reached the values of 10⁴-10⁵, and up to 10⁶ upon the completion of the processing cycle. Obtained results showed a high efficiency of the CARBOFLUOREX process for recovery and separating of U from FPs contained in FA, which allows completing of the FVM cycle with recovery of U and Pu from hardly processed FA.

• Journal of the Korean Ceramic Society
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• v.32 no.4
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• pp.471-481
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• 1995

Unirradiated UO2 pellets were pulverized by oxidation in air at 40$0^{\circ}C$, and the oxidized powders were reduced in H2 and CO atmospheres at $600^{\circ}C$. During the oxidation of UO2 at 40$0^{\circ}C$, intergranular cracks which caused the spallation were mainly developed by the volume contraction due to the formation of intermediate phase (U4O9 or U3O7). As oxidation proceeded, U3O8 finally formed. As the oxidation/reduction cycles were repeated, the powder surface became coarser, specific surface area was increased and average particle size was decreased. The sintered densities of the powder were increased by the oxidation/reduction cycle due to the characteristic changes of the powder.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.21-26
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• 1997

Recently, thorium is receiving increasing attention as an important fertile material for the expanding nuclear power programs around the world. The superior nuclear and physical properties of thorium-based fuels could lead to very low fuel cycle cost and make thorium reactors economically attractive. In addition, the use of thorium in reactors would permit more efficient utilization of low cost uranium reserves and reduction nuclear wastes. In this work, the nuclear characteristics of a new type thorium fueled reactor (Radkowsky Thorium Reactor) consisting seed-blanket assemblies are addressed and compared with those typical assemblies of a PWR (CE type). Also, an assessment on several advantages of thorium fueled reactors is provided. All these results are based on the HELIOS code calculation.