• 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.

• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.445-454
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• 2002

The sorption of U(Vl) on a domestic granite is studied as a function of experimental conditions such as contact time, solution-solid ratio, ionic strength, and pH using a batch procedure. The distribution coefficients, $K_{d}$＇s, of U(VI) are about 1-100mL/g depending on the experimental conditions. The sorption of U(VI) onto granite particles is greatly dependent upon the contact time, solution-solid ratio, and pH, but very little is dependent on the ionic strength. It is noticed that an U(VI)-carbonate ternary surface complex can be formed in the neutral range of pH. In the alkaline range of pH above 7, U(VI) sorption onto granite particles is greatly decreased due to the formation of anionic U(VI)-carbonate aqueous complexes.s.

• Bulletin of the Korean Chemical Society
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• v.20 no.1
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• pp.59-64
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• 1999

Based on the study of hydrolysis behaviour of U(Ⅵ) ion and major fission product metal ions such as Cs(Ⅰ), Ce(Ⅲ), Nd(Ⅲ), Mo(Ⅵ), Ru(Ⅱ), and ZR(Ⅳ) in the titration media, the performance of noncomplexing-alkalimetric titration method for the determination of free acid in the presence of these metal ions was investigated and its results were compared to those from the completing methods. The free acidities could be determined as low as 0.05 meq in uranium solutions in which the molar ratio of U(Ⅵ)/H+ was less than 5, when the end-point of titration was estimated by Gran plot. The biases in the determinations were less than 1% and about +3% respectively for 0.4 meq and 0.05 meq of free acid at the U(Vl)/H+ molar ratio of up to 5. Applicability of this method to the determination of free acid in spent fuel solutions was confirmed by the analysis of nitric acid content in simulated spent fuel solutions and in a real spent fuel solution.

• Analytical Science and Technology
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• v.16 no.2
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• pp.117-124
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• 2003

Oxygen to metal ratio has been measured by wet and dry chemical analysis to study the properties of sintered $UO_2$ pellets and $U_3O_8$ in the lithium reduction process of spent pressurized water reactor fuels. Uranium dioxide pellets simulated for the spent PWR fuels with burnup values of 20,000~60,000 MWd/MtU were prepared by mixing $UO_2$ powder and oxides of fission product elements, pelleting the powder mixture and sintering it at $1,700^{\circ}C$ under a hydrogen atmosphere. For wet chemical analysis, the simulated spent fuels were dissolved with mixed acid (10 M HCl : 8 M $HNO_3$, 2.5 : 1, v/v) using acid digestion bomb technique. The total amount of uranium and fission products added in the simulated spent fuels were measured using inductively coupled plasma atomic emission spectrometry. Weight change of the simulated fuel during its oxydation was measured by thermogravimetry and then the O/M ratio result was compared to that obtained by wet chemical analysis. Influence of $Mo_{0.4}-Ru_{0.4}-Rh_{0.1}-Pd_{0.1}$, quaternary alloy, on the determination of O/M ratio was investigated.

• Journal of the Korean Ceramic Society
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• v.42 no.9 s.280
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• pp.618-623
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• 2005

HTGR (High Temperature Gas-Cooled Reactor) is highlighted to next generation power plant for producing the clean hydrogen gas. In this study, the spherical $UO_2$ kernel via $UO_3$ gel particles was prepared by the sol-gel process. Raw material of slightly Acid Deficient Uranyl Nitrate (ADUN) solution, which has pH = 1.10 and $[NO_3]/[U]$ mole ratio = 1.93, was obtained from dissolution of $U_3O_8$ powder with conc.-$HNO_3$. The surface of these spherical $UO_3$ gel particles, which was prepared from the broth solution, consisted of 1 M-uranium, 1 M-HMTA, and urea, were covered with the fine crystallite aggregates, and these particles were so hard that crushed well. But the other $UO_3$ gel particles prepared with the broth solution, consisted of 2 M-uranium, 2 M-HMTA, and urea, have soft surface characteristics and an amorphous phase. This type of $UO_3$ gel particles is some chance of doing possibility of high density from the compaction. The amorphous $UO_3$ gel particles was converted to $U_3O_8$ and then $UO_2$ by calcination at $600^{\circ}C\;in\;4\%\;-\;H_2\;+\;N2$ atmosphere.

• Journal of the Korean Chemical Society
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• v.32 no.3
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• pp.233-242
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• 1988

The uranium(VI) complexes with new unsaturated macrocyclic ligands of cryptand types and the neodymium(III) complexes with cryptand 222 and DBC ligands have been investigated polarographically in dimethylsulfoxide solvent. The reduction states, electron numbers involved in the reduction process, effects of the added acid on the polarograms of complexes, and the mechanisms of the reduction electrode reactions have been examined. The stability constants and mole-ratio of new complexes were also obtained by polarographic method. The reaction of ligands was controlled by the diffusion in the reduction with four electrons at a step, whereas the redox reaction with six electrons at three steps in $UO_2\;^{2+}$ complexes with macrocyclic ligands and the redox reaction with one electron at a step in $Nd^{3+}$ complexes with cryptand 222 and DBC have been observed. The imine ligands formed stable complexes with uranium(VI) above pH 7.0, and the neodymium(III) complexes with cryptand 222 and DBC ligands were stable above pH 4.0.

• Journal of the Korean Chemical Society
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• v.34 no.2
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• pp.143-158
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• 1990

Metal complexes were prepared by reacting uranium (Ⅵ), thorium (Ⅳ) and rare earth metal (Ⅲ) ions including Nd (Ⅲ), Sm (Ⅲ) and Ho (Ⅲ) with macrocyclic ligands including five crown ethers, nine crownands and one cryptand ligands, and subjected to NMR studies in order to examine coordination sites of the ligands and compositions of the complexes formed. Among the marcocyclic ligands, crown ethers and crownand ligands have shown down-field shifts of the methylene protons of the lcigands by forming stable complexes with all the metal ions and the differences of chemical shifts were decreased as increasing of the cavity-size of crown ethers for the same metal ions and decreasing of the atomic number of the rare earth metals for the same ligands. It has been found that crownand 22 gave a stable complex with uranium(Ⅵ) ion by the coordination through both oxygen and nitrogen atoms of the ligand whereas no complex was formed with the rare earth metal(Ⅲ) ions, which on the other hand were found to form stable complexes with cryptand 221. The rest of the crowand ligands have also been found to form stable complexes with uranium(Ⅵ) ion by coordinating through all the oxygen and nitrogen atoms of the ligands whereas no complexes were formed with the rare earth metal(Ⅲ) ions. It has also been shown by 1H-NMR study that uranium(Ⅵ), thorium(Ⅳ) and rare earth metal(Ⅲ) ions formed 1:1 complexes with the macrocyclic ligands except for thorium(Ⅳ) complex of 12C4 in which the mole ratio of metal to ligand is 1:2. More stable metal complexes show larger changes in chemical shifts of the coordinated ligand protons. Finally, the rare earth metal(Ⅲ) complexes of 18C6 have shown ligand exchange reaction with the solvent molecules in acetylacetone solution, which was not observed for the uranium (Ⅵ) complexes.

• Nuclear Engineering and Technology
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• v.31 no.2
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• pp.133-138
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• 1999

Research on the etching reaction of UO$_2$ with CF$_4$/O$_2$gas mixture plasma is carried out. The reaction rates are investigated as a function of CF$_4$/O$_2$ ratio, plasma power, and substrate temperature. It is found that there exists an optimum CF$_4$/O$_2$ ratio around 4:1 at all temperatures up to 37$0^{\circ}C$ and surface analysis using XPS X-ray Photoelectron Spectroscopy) confirms the result. Peak rate at the optimum gas composition increases with increasing temperature. Highest rate obtained in this study leaches 1050 monolayers/min. at 37$0^{\circ}C$ under r. f. power of 150 W, which is equivalent to about 0.5${\mu}{\textrm}{m}$/min. The rate also increases with increasing r. f. power, thus, higher power and higher substrate temperature will undoubtedly raise the etching reaction rate much further. This reaction seems to be an activated process, whose activation energy will be derived in the following experiments.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.601-610
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• 2024

This paper gives an extensive analysis on the characterization of xenon isotopic ratios for various nuclear reactors and nuclear explosions through neutronic depletion codes. The results of the characterization can be used for discriminating the sources of the xenon isotopes' release among the nuclear explosions and nuclear reactors. The considered sources of the xenon radionuclides do not only include PWR, CANDU, and nuclear explosions using uranium and plutonium bombs, but also IRT-200 and 5MWe Yongbyon (MAGNOX reactor) research reactors operated in North Korea. A new data base (DB) on xenon isotopic activity ratios was produced using the results of the characterization, which can be used in discrimination of the sources of xenon isotopes. The results of the study show that 5MWe Yongbyon reactor has quite different characteristics in ¹³⁵Xe/¹³³Xe ratio from the PWRs and the nuclear reactors have different characteristics in ¹³⁵Xe/¹³³Xe ratios from the nuclear explosions.