• Journal of Radiation Protection and Research
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• v.8 no.2
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• pp.36-40
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• 1983

Uranium radionuclides are electrodeposited on inexpensive stainless steel cathode from a mixed oxalate-chloride electrolyte. The factors affecting the optimum condition for the deposition are determined by studying the effects of deposition time, initial current, electrode spacing, pH of electrolyte and uranium concentration in the electrolyte at fixed cathode area. The experiment which was repeated 3 times at each uranium concentration with 60 minutes of deposition time, gave an error of less than 4% standard deviation at the 90% confidence level with average yield greater than 99%.

• Journal of Radiation Protection and Research
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• v.28 no.1
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• pp.43-50
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• 2003

This study describes a practical method for interpretation of bioassay results of inhaled uranium to assess the committed effective doses both for chronic and acute intake situations. Organs in the body were represented by a series of mathematical compartments for analysis of the behavior of uranium in the body according to the gastrointestinal track model, respiratory track model and biokinetic model recommended by the ICRP. An analytical solutions of the system of balance equations among the compartments were obtained using the Birchall's algorithm, and the urinary excretion function and the lung retention function of uranium were obtained. An initial or total intakes by intake modes were calculated by applying excretion and retention functions to the urinary uranium concentration and the lung burden measured with a lung counter. The dose coefficients given in ICRP 78 are used to estimate the committed effective doses from the calculated intakes.

• Polymer(Korea)
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• v.27 no.3
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• pp.242-248
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• 2003

PP-g-(AN/Sty) was synthesized by grafting with acrylonitrile (AN) and styrene (Sty) onto PP staple fiber using an electron beam accelerator and followed by amidoximination and phosphorylation. Mole fraction of AN in the graft chain increased with the increase of the AN content in the monomer mixture. The highest AN grafting yield of 45% was obtained at a monomer ratio of 40 vol% AN/60 vol% Sty. Mole fraction of AN in the graft chain decreased with the increase of methanol amount used its solvent. As reaction temperature increased, the grafting yield of copolymer increased and reached equilibrium at 50$^{\circ}C$. Amount of amidoxime group in fibrous ion exchanger was increased as increasing amount of hydroxylamine, and the maximum content of amidoxime group was observed at 5.8 mmol/g with the 9 wt% hydroxylamine concentration. Content of phosphorous group in fibrous ion exchanger increased up to 0.5 N phosphoric acid concentration, and then leveled off. The adsorption ability of the copolymer for uranyl ion by the chelating adsorbents was in the following order : bifunctional PP-g-(AN/sty) > amidoximated PP-g-(AN/Sty) > phosphorylated PP-g-(AN/Sty).

• Resources Recycling
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• v.22 no.3
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• pp.57-64
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• 2013

Sulfurization reaction characteristics of $Eu_2O_3$, uranium oxides($UO_2$, $U_3O_8$), mixture of $Eu_2O_3$ and uranium oxides, Eu-doped uranium oxides($(U,Eu)O_2$, $(U,Eu)_3O_8$), and phase-separated products prepared by HOX (High temperature OXidation) of $(U,Eu)O_2$ were investigated in the temperature range from 400 to $800^{\circ}C$. Only $Eu_2O_3$ in the mixture of $Eu_2O_3$ and uranium oxides was converted into $Eu_3S_4$ by sulfurization reaction at $450^{\circ}C$ without reaction between them. Sulfurization reaction behavior of $(U,Eu)_3O_8$ and $(U,Eu)O_2$ up to $600^{\circ}C$ was similar to $U_3O_8$ and $UO_2$, respectively, while they were sulfurized into Eu-rich $(U,Eu)S_x$ and ${\alpha}-US_2$ at $800^{\circ}C$. In the sulfurization of RE-rich $(U,Eu)_4O_9$ and $U_3O_8$ prepared by high temperature oxidation, it was confirmed that RE-rich $(U,Eu)S_x$ and UOS phases were formed at $600^{\circ}C$. For Eu-rich $(U,Eu)O_2$ and $UO_2$ prepared by reduction of HOX products, it was identified that Eu-rich (U,Eu)OS was formed at $450^{\circ}C$ by sulfurization of Eu-rich $(U,Eu)O_2$, while $UO_2$ remained unreacted.

• The Sea:JOURNAL OF THE KOREAN SOCIETY OF OCEANOGRAPHY
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• v.4 no.3
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• pp.188-197
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• 1999

In order to estimate the uranium flux from seawater to sediments, we took pore water samples and deployed benthic chambers on seafloor of Chonsu Bay, Korea. The uranium flux across the sediment-water interface was estimated from the pore water to be 0.112-0.566 mg/$m^2yr$, corresponding to a removal flux of $4.3-21.5{\times}10^7$ gU/yr for the entire Yellow Sea. Nutrient fluxes from sediment to bottom water were estimated to be 135.6 mmol/$m^2yr$ for ammonia, 228.2 mmol/$m^2yr$ for nitrate, 36.8 mmol/$m^2yr$ for phosphate and 23.9 mmol/$m^2yr$ for silicate. The redox boundary, based on the distribution of pore water nitrate and solid phase manganese, was located at 3-5 cm below the sediment surface. Phosphate flux obtained by benthic chambers was 28.S mmol/$m^2yr$. On the other hand, estimates of uranium and silicate fluxes were orders of magnitude greater than those based on pore water profiles. Flux estimates on the basis of pore water concentration is believed to have greater reliability than those obtained from benthic chamber data.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.2
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• pp.101-116
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• 2017

The interaction of U(VI) (hexavalent uranium) species with natural organic matter (NOM) in KURT (KAERI Underground Research Tunnel) groundwater is investigated using a laser spectroscopic technique. The luminescence spectra of the NOM are observed in the ultraviolet and blue wavelength regions by irradiating a laser beam at 266 nm in groundwater. The luminescence spectra of U(VI) species in groundwater containing uranium concentrations of $0.034-0.788mg{\cdot}L^{-1}$ are measured in the green-colored wavelength region. The luminescence characteristics (peak wavelengths and lifetime) of U(VI) in the groundwater agree well with those of $Ca_2UO_2(CO_3)_3(aq)$ in a standard solution prepared in a laboratory. The luminescence intensities of U(VI) in the groundwater are weaker than those of $Ca_2UO_2(CO_3)_3(aq)$ in the standard solution at the same uranium concentrations. The luminescence intensities of $Ca_2UO_2(CO_3)_3(aq)$ in the standard solution mixed with the groundwater are also weaker than those of $Ca_2UO_2(CO_3)_3(aq)$ in the standard solution at the same uranium concentrations. These results can be ascribed to calcium-U(VI)-carbonate species interacting with NOM and forming non-radiative U(VI) complexes in groundwater.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.72-80
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• 2004

Electrolytic dissolution study was carried out to evaluate the applicability of electrochemical decontamination process using a neutral salt electrolyte as a decontamination technology for the recycle or self disposal with authorization of large amount of metallic wastes contaminated with uranium compounds generated by dismantling a retired uranium conversion plant using SUS-304 and Inconel-600 specimen as the main materials of internal system components of the plant. The effects of type of neutral salt as an electrolyte, current density, and concentration of electrolyte on the dissolution of the materials were evaluated. 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 plant were peformed in $Na_2SO_4$ and $NaNO_3$ solution. It was verified that the electrochemical decontamination of the dismantled metallic wastes was quite successful in $Na_2SO_4$ and $NaNO_3$ neutral salt electrolyte by reducing $\beta$ radioactivities below the level of self disposal with authorization within 10 minutes regardless of the type of contaminants and the degree of contamination.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.11a
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• pp.343-344
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• 2009

• Journal of Radiation Protection and Research
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• v.37 no.4
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• pp.213-218
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• 2012

Alpha spectrometry has been typically used for determination of the uranium isotopes in soil. For a number of uranium analysis in soil samples, rapid sample digestion with limited quantities of mixed acid containing HF will give a contribution for effective management of uranium analysis. Microwave digestion system is evaluated for rapid sample digestion using reference uranium soil (IAEA-375 soil). For completion of 0.5 g of soil digestion by microwave, 3 ml of HF in a 10 ml of mixed acid is minimum requirement volume for completed soil digestion for 80 minutes. Microwave digestion is timely effective techniques for uranium measurement using alpha spectrometry compared to the other methods (open vessel digestion, closed vessel digestion) due to rapid sample digestion. In addition, it can be reduced the occurrence of hazardous substances by minimizing the amount of HF.

• Analytical Science and Technology
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• v.8 no.3
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• pp.285-292
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• 1995

Uranium has variable oxidation states(${UO_2}^{+2}$, $UO^{+2}$, $U^{+4}$, $U^{+3}$) and 1-(2-Pyridylazo)-2-naphthol forms a very stable chelate with Uranium(${UO_2}^{+2}$). The determination method of Uranium(${UO_2}^{+2}$) in 0.1M Borate buffer(pH 7.10) has been investigated by adsorptive stripping voltammetry. The optimum conditions were PAN concentration of $5{\times}10^{-7}M$, accumulation potential of 0.00V(vs. Ag/AgCl) and accumulation time of 120sec. The calibration curve was linear over the range of $5{\sim}60{\mu}g/L$ and the various metal ions did not interfere with the determination Uranium(${UO_2}^{+2}$) except Cu(II) and Co(II).