• Proceedings of the Korean Nuclear Society Conference
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• 1995.05b
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• pp.853-858
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• 1995

$^{152}$EU, $^{241}$Am 및 $^{241}$Np를 선정하여 추출제 di-(2-ethylhexyl)phosphoric acid (DEHPA)로 이들 세가지 핵종의 상호 분리공정에 대해 연구하였다. 추출공정에서 질산농도는 0.IM이고 추출 제DEHPA/n-dodecane 농도가 IM 일때, $^{152}$Eu 및 $^{241}$Am, $^{241}$Np 핵종 모두 99.9% 이상 추출되었다. $^{241}$Am 핵종의 역추출을 위한 제 1 역추출공정에서 역추출제인 0.05M di-ethylenetriamine pentaacetic acid (DTPA)와 1.5M 젖산 (lactic acid) 혼합용액의 pH가 3.6 일때 $^{152}$Eu과 $^{241}$Am의 상호분리비와 역추출율이 가장 좋았다. $^{152}$Eu의 역추출을 위한 제 2 역추출단계에서 역추출제인 질산용액의 최적농도는 6M이며, 이때 $^{152}$Eu은 99%가 역추출되었다.$^{241}$Np 핵종의 역추출을 위한 제 3 역추출공정에서 역추출제는 0.5M 옥살산이며, 이때 $^{241}$Np은 92%가 역추출되었다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.1
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• pp.1-7
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• 2005

Anion exchange chromatography and HDEHP extraction chromatography using DTPA-lactic acid as an eluent were applied in series for the separation of $^{241}$Am and $^{244}$Cm in radwaste samples. The separated elements were determined by electrodeposition at the sodium hydrogen sulfate-sodium sulfate buffer solution followed by alpha-spectrometry. The recovery yields of $^{241}$Am and $^{244}$Cm were 85.2$\pm$$15.3\%$, respectively, from the synthetic solution of spent nuclear fuel sample. The amounts of 241Am and 2440m determined in radwaste sample solutions of condensate bottoms were at the range of 1.5-1.9 Bq/g and -1.7 Bq/g, respectively.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.06a
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• pp.248-254
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• 2005

This paper presents a quantitative method of sequential separation of $^{90}Sr,\;^{241}Am$ and Pu nuclides with an anion exchange resin and a Sr-Spec resin. The Pu isotopes were purified with an anion exchange resin. The americium and strontium fractions were separated from the matrix elements with an oxalate co-precipitation method. Americium fraction was separated from the strontium fraction with iron co-precipitation method and purified from lanthanides with anion exchange resin. Strontium-90 was purified from other hindrance elements with the Sr-Spec resin after oxalate co-precipitation. The measurement of Pu and Am isotopes was carried out by an ${\alpha}$-spectrometer. Strontium-90 was measured by a liquid scintillation counter. The radiochemical procedure of $^{90}Sr,\;^{241}Am$ and Pu nuclides investigated in this study has been validated by application to IAEA-Reference soils.

• Analytical Science and Technology
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• v.18 no.6
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• pp.469-474
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• 2005

This paper presents a quantitative method of sequential separation of $^{90}Sr$, $^{241}Am$ and Pu radionuclides with an anion exchange resin and a Sr-Spec resin. The Pu isotopes were purified with an anion exchange resin. The americium and strontium fractions were separated from the matrix elements with an oxalate co-precipitation method. Americium fraction was separated from the strontium fraction with iron co-precipitation method and purified from lanthanides with anion exchange resin. Strontium-90 was purified from other hindrance elements with the Sr-Spec resin after oxalate co-precipitation. The measurement of Pu and Am isotopes was carried out by an ${\alpha}$-spectrometer. Strontium-90 was measured by a liquid scintillation counter. The radiochemical procedure of $^{90}Sr$, $^{241}Am$ and Pu radionuclides investigated in this study has been validated by application to IAEA-Reference soils.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.423-428
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• 1996

The feasibility study on burning Am-241, 243 and Cm-244 nuclides in a conventional PWR (Pressurized Water Reactor) was carried out by using the TRIFON code that was developed by the Institute of Theoretical and Experimental Physics in Russia in 1992. TRIFON code uses updated ABBN Russian nuclear cross section library. The reference reactor is the Korea nuclear power plant unit 8 (YGN 2). The burning effect of Am-241, 243 and Cm-244 nuclides was studied with UO$_2$(3.5 w/o)fuel assembly and MOX (4.44 w/o) fuel assembly. The loaded mass ratio of Am-241, 243 and Cm-244 nuclides was obtained from the mass ratio of Am-241, 243 and Cm-244 nuclides in 10 year cooling spent fuel with average discharge burnup of 33 GWD/MTU. The effective transmutation rates of Am-241, 243 and Cm-244 nuclides in UO$_2$ fuel assembly were found to be higher than those in MOX fuel assembly. The result from TRIFON code was compared to that from CASMO-3/NEM-3D code system. For more reliable calculation of transmutation for MA(Minor Actinides) more sophisticated decay chain scheme of MA should be investigated and nuclear cross section library of MA should be considerably improved.

• Applied Chemistry for Engineering
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• v.7 no.1
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• pp.153-161
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• 1996

Several main nuclides($^{241}Am$, $^{152}Eu$ and $^{237}Np$) in radioactive waste solution were selected and examined to mutual separation with di-(2-ethylhexyl) phosphoric acid by solvent extraction technique. $^{237}Np$ was extracted more than 99.9% adding the $H_2O_2$ that was a good reductant for the oxidation state control of $^{237}Np$. $^{241}Am$, $^{152}Eu$ and $^{237}Np$ could be fairly well separated one another during the different sequence stripping stages, but about 7~9.6% of the other nuclides were still remained for the $^{241}Am$ stripping solution. This result shows that the product of $^{152}Eu$ and $^{237}Np$ was good, but $^{241}Am$ may be needed to further purification process. It was also discussed on the cause of the third phase formation phenomenon that was found in the solvent regeneration.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.11a
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• pp.407-408
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• 2009

In this work, a quantitative method of the sequential separation of $^{90}Pu,\;^{90}Sr$ and $^{241}Am$ nuclides was developed in environmental soil samples using by an anion exchange resin connected with a Sr Spec resin. The sequential methods of Pu, $^{90}Sr$ and $^{241}Am$ nuclides can reduce analysis cost and time. With the sequential methodupdated in this study, the activity concentrations of $^{239,240}Pu,\;^{90}Sr$ and $^{241}Am$ in the IAEA reference materials were close to the reference values reported by the IAEA.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.278-278
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• 1998

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.383.1-383
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• 2002

• Nuclear Engineering and Technology
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• v.3 no.3
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• pp.155-160
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• 1971

Average and effective energies for 239Pu-Be, 241Am-Li and 241Am-F neutron sources have been calculated from a number of published data for the neutron spectra and for the dose equivalent as a function of neutron energies by a numerical method. Also a calculation of the dose equivalent conversion factors, i. e., the first collision dose equivalent and the surface (or multicollision) dose equivalent that equals the product of surface-absorbed dose and a corresponding quality factor, per unit fluence of neutrons from these sources has been carried out in the same way as before. The results are as follows : 1. for average energies 4.07$\pm$0.33, 0.42 and 1.41 MeV； 2. for effective energies based on the concept of the first collision process in the human body 4.45$\pm$0.344, 0.51 and 1.47 MeV; 3. for effective energies based on the concept of the multi-collision process in the human body 4.50$\pm$0.36, 0.50 and 1.45 MeV； 4. for fluence-first collision dose equivalent conversion factors (2.74$\pm$0.07)10$^{-8}$ , 1.58$\times$ 10$^{-8}$ and 2.34$\times$10$^{-8}$ rems/(n/$\textrm{cm}^2$); and 5. for fluence-surface dose equivalent conversion factors (3.55$\pm$0.09)10$^{-8}$ , 2.19$\times$10$^{-8}$ and 2.82$\times$10$^{-8}$ rems/(n/$\textrm{cm}^2$) : respectively.