• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.1
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• pp.25-33
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• 2008

A selective separation of Actirlide(III) by a nPr-BTP/nitrobezene extraction system was studied. The nPr-BTP (2.6-Bis-(5.6-n-propyl-1.2.4-triazin-3-yl)-pyridine) of a environmentally -friendly CHN type was self-synthesized and its compatability with diluent and stability with nitric acid were investigated. At the 0.1M nPr-BTP/nitrobenzene-1M $HNO_3$ and O/A=2, extraction yields of Am used as a representative of Actinide(III) and Eu were about 85% and 8%, respectively, and the other RE elements such as Nd, Ce and Y were extracted less than 3% (separation factor of Am and Eu was about 60). Thus, there was no problems in the selective extraction of Actinide(III) from RE. The stripping yield of Am with 0.05M $HNO_3$ at O/A= 1, however, was about 43% and the maximum stripping yield was 65% at O/A=0.3. It is necessary to develop the stripping system including the stripping agent instead of nitric acid solution.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11a
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• pp.336-336
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• 2005

• Nuclear Engineering and Technology
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• v.15 no.2
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• pp.117-122
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• 1983

The chelating agent with $\beta$-diketo funtional group, 1-phenyl-3-methyl-4-acyl-pyrazolone-5-one, has been used in separating and extracting radionuclides in a waste solution. The derivatives of this pyrazolone compound, prepared by different acyl groups, were synthesized and examined to figure out the extracting ability for Uranium (VI) and Zirconium (IV). The product prepared with succinic anhydride, called succinyl pyrazolone, showed excellent extraction for uranium (VI) in a chloroform solvent system. This result indicates that acyl pyrazolones having carboxylic acid group as a functional group forming $\beta$-diketo functionality are very selective for uranium (VI) and generally other metal ions with high valency.

• Journal of Radiation Protection and Research
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• v.26 no.1
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• pp.1-6
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• 2001

An accurate and rapid analytical technique of uranium isotopes in highly contaminated soil samples was developed and validated by application to the IAEA-Reference samples. For overcoming the demerits of the TBP extraction method, sample materials were decomposited with $HNO_3$ and HF, and uranium isotopes were purified by an anion exchange resin and a TRU Spec resin. With the extraction chromatography method, the hindrance elements were completely removed from the uranium fraction. The chemical yields with the extraction chromatography method were more 10% higher than those with the TBP extraction method. The concentrations of uranium isotopes in soil samples using the extraction chromatography method were consistent with the reference values reported by the IAEA.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.2
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• pp.123-132
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• 2007

This study was performed to evaluate the co- and mutual separation for Am, Cm and RE elements from the simulated multi-component solution equivalent to real HLW level by a Zr-DEHPA(di-(2-ethylhexyl) phosphoric acid containing Zirconium)/$NDD(n-dodecane)-HNO_3$ extraction system. Zr-DEHPA was self-synthesized and the optimal condition of (15g/L Zr-1M DEHPA)/NDD-1M $HNO_3$ was selected taking into consideration of prevention of the third phase, and effects of concentration of DEHPA, nitric acid and impregnant amount of Zr on the co-extraction of Am, Cm and RE. In that condition, the extraction yields were 81% (Am), 85% (Cm), more than 80% (RE elements), 98% (Mo), 85% (Fe), 98% (U), 73% (Np), and less than 5% (other elements) so that the system developed for the co-extraction of Am-Cm/RE was proved to be available. For that, however, U, Np, Mo and Fe was elucidated to have to be removed in advance, and Zr inducing the third phase formation was found to be practically excluded. The co-extracted Am-Cm/RE were sequentially separated in an order of Am-Cm (stripping agent : 0.05 M DTPA-1M Lactic acid of pH 3.6)${\rightarrow}RE$ (stripping agent : 5M $HNO_3$), and then their separation factors were evaluated. At above conditions, Am of 65.4%, Cm of 63.9%, RE (except for Y) of more than 85% were stripped.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.2
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• pp.95-102
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• 2006

Alpha spectrometry was studied for the determination of $^{237}Np$ in spent nuclear fuel samples. The optimum condition for the electrodeposition of $^{237}Np$ was obtained as follows : for $1{\sim}1.5$ hour of deposition time, at the current intensity of $1.2{\sim}1.5$ A and at sodium sulfate electrolyte without organic additive. The deposition yield and its reproducibility on $^{237}Np$ was decreased as the amount of $^{237}Np$ decreased from 4.16 Bq down to 0.0264 Bq(1ng). The recovery yield of $^{237}Np$ determined by alpha spectrometry after separation in synthetic solution was $98.8{\pm}5.1%$(n=4). The contents of $^{237}Np$ in spent nuclear fuel samples were determined and the result showed an agreement within 10% of a difference between the measurement and the calculation.