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Abstract  

For the determination of 210Po in water samples, two alternative procedures (a) DDTC solvent extraction and (b) extraction chromatography using Sr Resin were selected and then validated in terms of trueness, repeatability and reproducibility with a tap water spiked with a known amount of 210Po. In this work the optimization conditions for the auto-deposition of Po for source preparation were also studied.

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Abstract  

This work investigates the sorption of americium [Am(III)] onto kaolinite and the influence of humic acid (HA) as a function of pH (3–11). It has been studied by batch experiments (V/m = 250:1 mL/g, C Am(III) = 1 × 10−5 mol/L, C HA = 50 mg/L). Results showed that the Am(III) sorption onto the kaolinite in the absence of HA was typical, showing increases with pH and a distinct adsorption edge at pH 3–5. However in the presence of HA, Am sorption to kaolinite was significantly affected. HA was shown to enhance Am sorption in the acidic pH range (pH 3–4) due to the formation of additional binding sites for Am coming from HA adsorbed onto kaolinite surface, but reduce Am sorption in the intermediate and high pH above 6 due to the formation of aqueous Am-humate complexes. The results on the ternary interaction of kaolinite–Am–HA are compared with those on the binary system of kaolinite–HA and kaolinite–Am and adsorption mechanism with pH are discussed. Effect of different molecular weight of HA, with three HA fractions separated by ultrafiltration techniques, on the Am sorption to kaolinite were also studied. The results showed that the enhancement of the sorption of Am onto kaolinite at the acidic pH conditions (pH 3–4) was higher with HA fractions of higher molecular weight. Also, the Am sorption over a pH range from 6 to 10 decreased with decreasing molecular weight of HA.

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Abstract  

Instrumental neutron activation analysis was used to measure the concentrations of about 27 elements associated with airborne PM 10 samples that were collected from a roadside sampling station at a moderately polluted urban area of Taejon city, Korea. The magnitude of their concentrations was clearly distinguished and spanned over four orders. If compared in terms of enrichment factors, it was found that certain elements (e.g., As, Br, Cl, Sb, Se, and Zn) are enriched in PM 10 samples of the study site. The factor analysis indicated three factors with statistical significance, which may exert dominant controls on regulating the metal concentration levels in the study area.

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Abstract  

A simple and rapid separation procedure was systemized for the determination of 99Tc, 90Sr, 94Nb, 55Fe and 59,63Ni in low and intermediate level radioactive wastes. The integrated procedure involves precipitation, anion exchange and extraction chromatography for the separation and purification of individual radionuclide from sample matrix elements and from other radionuclides. After separating Re (as a surrogate of 99Tc) on an anion change resin column, Sr, Nb, Fe and Ni were sequentially separated as follows; Sr was separated as Sr (Ca-oxalate) co-precipitates from Nb, Fe and Ni followed by purification using Sr-Spec extraction chromatographic resin. Nb was separated from Fe and Ni by anion exchange chromatography. Fe was separated from Ni by anion exchange chromatography. Ni was separated as Ni-dimethylglyoxime precipitates after the removal of 134,137Cs and 110mAg by Cs-phosphotungstate and AgCl precipitation, respectively. Finally, the radionuclide sources were prepared by precipitation for their radioactivity measurements. The reliability of the procedure was evaluated by measuring the recovery of chemical carriers added to a synthetic radioactive waste solution.

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Reaction Kinetics, Mechanisms and Catalysis
Authors: Saet Byul Kim, Mi Ran Lee, Eun Duck Park, Sang Min Lee, HyoKyu Lee, Ki Hyun Park, and Myung-June Park

Abstract

A kinetic model of the homogeneous conversion of d-xylose in high temperature water (HTW) was developed. Experimental testing evaluated the effects of operating conditions on xylose conversion and furfural selectivity, with furfural yields of up to 60% observed without the use of acid catalysts. The reaction order for the decomposition of d-xylose was assumed to be above two, while the conversion of d-xylose to furfural and the degradation of furfural were first order reactions. Estimated kinetic parameters were within the range of values reported in the literature. The activation energy of furfural production showed that the ionization rate was high enough for HTW to replace acid catalysts. Simulated results from this model were in good agreement with experimental data, allowing the model to aid reactor design for the maximization of productivity.

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Abstract  

In order to evaluate radionuclide inventories as an essential item for the permanent disposal of spent fuel storage racks, chemical conditions for a sample pretreatment of a spent fuel storage rack were studied. Especially, the surface microstructure and the radionuclide distributions for the spent fuel storage rack were investigated by using a SEM–EDX and γ-spectrometer for minimizing the matrix effect which could affect a chemical separation process of some β-emitting radionuclides. The samples were pretreated with a mixed solution of 5 M HCl and 2 M HNO3 by an ultrasonic surface leaching method. Some radionuclides in the raw racks showed the radioactivity of 102–103 Bq for about 10 g of sample weight. From the sample pretreatment, it was confirmed that almost all radionuclides in the rack were completely extracted from the rack when the dissolved thickness of the rack became a maximum 15 μm by the ultrasonic surface leaching method. The established pretreatment method was applied for all spent fuel storage rack generated from Korean NPPs to determine the scaling factor. The radioactivities of 60Co and 137Cs radionuclides in the pretreated solutions were in the range of 4.9E−1~1.5E+2 and 1.2E−1~9.0E+0 Bq/g, respectively.

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Abstract  

This work studied a way to reclaim uranium from contaminated UO2 oxide scraps as a sinterable UO2 powder for UO2 fuel pellet fabrication, which included a dissolution of the uranium oxide scraps in a carbonate solution with hydrogen peroxide and a UO4 precipitation step. Dissolution characteristics of reduced and oxidized uranium oxides were evaluated in a carbonate solution with hydrogen peroxide, and the UO4 precipitation were confirmed by acidification of uranyl peroxo–carbonate complex solution. An agglomerated UO4 powder obtained by the dissolution and precipitation of uranium in the carbonate solution could not be pulverized into fine UO2 powder by the OREOX process, because of submicron-sized individual UO4 particles forming the agglomerated UO4 precipitate. The UO2 powder prepared from the UO4 precipitate could meet the UO2 powder specifications for UO2 fuel pellet fabrication by a series of steps such as dehydration of UO4 precipitate, reduction, and milling. The sinterability of the reclaimed UO2 powder for fuel pellet fabrication was improved by adding virgin UO2 powder in the reclaimed UO2 powder. A process to reclaim the contaminated uranium scraps as UO2 fuel powder using a carbonate solution was finally suggested.

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