Kyuseok Song

Determination of Pu and U isotopes in safeguard swipe sample with extraction chromatographic techniques.

A simple and reliable separation method of plutonium and uranium was developed using extraction chromatographic techniques. Using the separation method of plutonium and uranium developed in this study, recoveries of plutonium and uranium were compared with those by commercial separation methods of Pu and U. The redox reactions of hydrogen peroxide with plutonium in a nitric acid medium were investigated by UV-vis-NIR absorption spectroscopic techniques. The separation method of Pu and U isotopes with the UTEVA resin updated in this study could be effective to measure quantitatively nano- and picogram amounts of uranium and plutonium in swipe samples using isotope dilution thermal ionization mass spectrometry (ID-TIMS).

Rapid sequential determination of Pu, 90Sr and 241Am nuclides in environmental samples using an anion exchange and Sr-Spec resins.

This paper presents a quantitative and rapid method of sequential separation of Pu, (90)Sr and (241)Am nuclides in environmental soil samples with an anion exchange resin and Sr Spec resin. After the sample solution was passed through an anion exchange column connected to a Sr Spec column, Pu isotopes were purified from the anion exchange column. Strontium-90 was separated from other interfering elements by the Sr Spec column. Americium-241 was purified from lanthanides by the anion exchange resin after oxalate co-precipitation. Measurement of Pu and Am isotopes was carried out using an α-spectrometer. Strontium-90 was measured by a low-level liquid scintillation counter. The radiochemical procedure of Pu, (90)Sr and (241)Am nuclides investigated in this study validated by application to IAEA reference materials and environmental soil samples.

Oxidation State Shift of Uranium during U(III) Synthesis with Cd(II) and Bi(III) in LiCl–KCl Melt

Synthesis of U 3+ with Cd 2+ or Bi 3+ was investigated in a LiCl―KCl eutectic melt at 500°C by using cyclic voltammetry (CV) and ultraviolet-visible (UV-vis) spectroscopy. CV results showed that the equilibrium potentials of the species are larger in sequence of U 0/3+ < Cd 0/2+ < U 3+/4+ < Bi 0+/3+ . UV-vis spectroscopy results exhibited that U 3+ was directly produced in the eutectic melt containing Cd 2+ . In BiCl 3 , U metal was oxidized to U 4+ by Bi 3+ , and then U 4+ was reduced to U 3+ by additional U. From these results, the reaction mechanisms taking place during the synthesis are proposed.

The measurement of 129I for the cement and the paraffin solidified low and intermediate level wastes (LILWs), spent resin or evaporated bottom from the pressurized water reactor (PWR) nuclear power plants.

In this paper a relatively simple and low cost analysis procedure to apply to a routine analysis of (129)I in low and intermediate level radioactive wastes (LILWs), cement and paraffin solidified evaporated bottom and spent resin, which are produced from nuclear power plants (NPPs), pressurized water reactors (PWR), is presented. The (129)I is separated from other nuclides in LILWs using an anion exchange adsorption and solvent extraction by controlling the oxidation and reduction state and is then precipitated as silver iodide for counting the beta activity with a low background gas proportional counter (GPC). The counting efficiency of GPC was varied from 4% to 8% and it was reversely proportional to the weight of AgI by a self absorption of the beta activity. Compared to a higher pH, the chemical recovery of iodide as AgI was lowered at pH 4. It was found that the chemical recovery of iodide for the cement powder showed a lower trend by increasing the cement powder weight, but it was not affected for the paraffin sample. In this experiment, the overall chemical recovery yield of the cement and paraffin solidified LILW samples and the average weight of them were 67+/-3% and 5.43+/-0.53 g, 70+/-7% and 10.40+/-1.60 g, respectively. And the minimum detectable activity (MDA) of (129)I for the cement and paraffin solidified LILW samples was calculated as 0.070 and 0.036 Bq/g, respectively. Among the analyzed cement solidified LILW samples, (129)I activity concentration of four samples was slightly higher than the MDA and their ranges were 0.076-0.114 Bq/g. Also of the analyzed paraffin solidified LILW samples, five samples contained a little higher (129)I activity concentration than the MDA and their ranges were 0.036-0.107 Bq/g.

Infrared Spectra of Uranium Oxides Measured by ATR-FTIR

The infrared spectra of uranium oxides (UOx) of various O/U ratios (x = 2–2:667 in UOx) were measured by using ATR-FTIR in order to investigate the applicability of ATR technology to the characterization of uranium oxides or nuclear fuel. The samples of uranium oxides (UOx) were prepared by heating UO2 powder in air atmosphere followed by weighing to determine the gains by oxygen. X-ray diffraction (XRD) patterns for the same samples of infrared spectra were measured to get information on their phases on uranium oxides (UOx) with increasing x. The measured characteristic bands from UOx samples could be assigned well by referring to the spectra reported in the literature and comparing with the diffraction patterns measured. The relationship between absorbance and O/U ratio (x) was determined. We discussed about using it for the O/U ratio estimation of uranium oxides of unknown composition. It was confirmed that ATR technology for infrared spectroscopy can be useful for the studies of radioactive materials that are difficult to handle using conventional methods.

BEHAVIORS OF MOLYBDENUM IN UO₂ FUEL MATRIX

Molybdenum is the most abundant fission product since its fission yield is equivalent to that of xenon, and it has a very special role in the chemistry of nuclear fuel because it influences the oxygen potential of UO₂ fuel. In this study, the distribution of molybdenum in spent UO2 fuel specimens with 33.3, 41.0 and 57.6 GWd/tU burnup was measured by a LA-ICP-MS system and the reproducibility of the measured data was obtained. The Mo distribution was almost constant along the radius of a fuel except an increase at the periphery of the fuel. It showed a drop in reproducibility with relatively high deviation of measured values for the highest burnup fuel. To explain this, the state of molybdenum in a UO₂ matrix and its effect on the oxidation behavior of UO₂ were investigated. The low reproducibility was explained by the segregation of molybdenum, and the inhibition of oxidation by the molybdenum was also observed.

Temperature dependence of fluorescence for EuCl3 in LiCl–KCl eutectic melt

The fluorescence of EuCl(3) in LiCl-KCl eutectic melt according to temperature changes was investigated, and the spontaneous partial reduction of Eu(3+) to Eu(2+) at high temperature was confirmed by the fluorescence results. The fluorescence decreases when the temperature increases, and this was examined in detail. The studies of fluorescence provided information regarding the chemical and physical behavior of europium ions in the molten salt according to the temperature changes. It is applicable for monitoring species and concentrations and estimating the approximate chemical structure of the ions in molten salts.

Classification of materials for explosives from prompt gamma spectra by using principal component analysis

Data from the elemental composition-ratios and experimental prompt gamma spectra of samples were used to develop suitable discriminant classes for suspect samples. The collected data and gamma spectra were applied to principal component analysis (PCA) to discriminate explosives from non-explosive materials.

Explosives Detection Using Deuterium-Deuterium Neutron Generator

Gamma spectra of general and explosive materials were obtained by using the developed neutron induced prompt gamma-ray spectrometry system containing a neutron generator, where the neutron source was created from a deuterium-deuterium reaction, and they were examined to establish whether the samples were explosives or not by using the pattern recognition method. When the gamma spectra of real explosives (or general materials) were projected into the already prepared training set for the pattern recognition, they were placed in the categories of explosives (or general materials) properly.

Application of a wet oxidation method for the quantification of 3H and 14C in low-level radwastes

Wet oxidation methods are utilized to separate and quantify (3)H and (14)C radionuclides in inorganic and organic radwastes generated at nuclear facilities. Because (3)H and (14)C are pure beta emitters with half-lives of 12.3 and 5,730 years, respectively, these radionuclides should be chemically separated from other radionuclides present in radwastes for accurate quantification. In particular, a collection technique for (14)C radionuclide in radwastes is needed because it is converted into (14)CO2 gas by an oxidation reaction. To confirm the recoveries of (3)H and (14)C, various standard radioactive sources were used to verify the proposed method. Because the majority of (3)H radionuclides are distributed in tritiated water (HTO), only tritiated water was used as a standard for (3)H radionuclides. Additionally, (14)C-labeled methanol ((14)CH3OH), lauric acid ((14)CH3(CH2)10COOH), sodium bicarbonate (NaH(14)CO3), and toluene (C6H5(14)CH3) were used as (14)C standards. The compounds were oxidized with chemical oxidants and then separated. The individual species were mixed with a scintillation cocktail and counted using a liquid scintillation counter. The recoveries of (14)C and (3)H were 82-97% and 98%, respectively. The wet oxidation method will be applied to RI wastes for clearance.