• Analytical Science and Technology
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• v.17 no.4
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• pp.302-307
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• 2004

Determination of actinides in the environmental sample requires separation of each element. This procedure is tedious and time consuming. And also, the detection limits of some nuclides using alpha spectrometry are rather higher. To overcome the lower detection limit and complicated separation procedure, a simple analytical technique for the determination of actinide isotopes in the environmental samples was developed and applied to IAEA and NIST reference sediment samples. For the separation of actinides from matrix, anion exchange resin and TRU-spec extraction chromatography resin were used and chemical yields were obtained using natural uranium, thorium, $^{242}Pu$ and $^{243}Am$ tracers. For overcoming the higher detection limits of U and Th in alpha spectrometry, neutron activation analysis was applied. Using combined method, the detection limit was increased about 10 times. The activity values of each isotope were consistent with the reference values reported by IAEA and NIST.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.103-103
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• 2018

The liquid cathode processing is necessary to separate cadmium from the actinide elements in the pyroprocessing since the actinide deposits are dissolved or precipitated in a liquid cathode. Distillation process was employed for the cathode processing owing to the compactness. It is very important to avoid a splattering of cadmium during evaporation due to the high vapor pressure. Several methods have been proposed to lower the splattering of cadmium during distillation. A multi-layer porous round cover was proposed to avoid a cadmium splattering in our previous study. In this study, distillation behavior of $Cd-ZrO_2$ and Cd - Bi systems were investigated to examine a multi-layer porous round cover for the development of the cadmium splatter shield of distillation crucible. It was designed that the cadmium vapor can be released through the holes of the shield, whereas liquid drops can be collected in the multiple hemisphere. The cover was made with three stainless steel round plates with a diameter of 33.50 mm. The distance between the hemispheres and the diameter of the holes are 10 and 1 mm, respectively. Bismuth or zirconium oxide powder was used as a surrogate for the actinide elements. About 40 grams of Cd was distilled at a reduced pressure for two hours at various temperatures. The mixture of the cadmium and the surrogate was distilled at 470, 570 and $620^{\circ}C$ in the crucible with the cover. Most of the bismuth or zirconia remained in the crucible after distillation at 470 and $570^{\circ}C$ for two hours. It was considered that the crucible cover hindered the splattering of the liquid cadmium from the distillation crucible. A considerable amount of the surrogate material reduced after distillation at $620^{\circ}C$ due to the splattering of the liquid cadmium. The low temperature is favorable to avoid a liquid cadmium splattering during distillation. However, the optimum temperature for the cadmium distillation should be decided further, since the evaporation rate decreases with a decreasing temperature.

• 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 Nuclear Society Conference
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• 2001.05a
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• pp.80-80
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• 2001

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.590-591
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• 2018

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.3
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• pp.373-382
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• 2020

This study proposes a method of separating uranium (U) and minor actinides from rare earth (RE) elements in the LiCl-KCl salt system. Several RE metals were used to reduce UCl₃ and MgCl₂ from the eutectic LiCl-KCl salt systems. Five experiments were performed on drawdown U and plutonium (Pu) surrogate elements from RECl₃-enriched LiCl-KCl salt systems at 773 K. Via the introduction of RE metals into the salt system, it was observed that the UCl₃ concentration can be lowered below 100 ppm. In addition, UCl₃ was reduced into a powdery form that easily settled at the bottom and was successfully collected by a salt distillation operation. When the RE metals come into contact with a metallic structure, a galvanic interaction occurs dominantly, seemingly accelerating the U recovery reaction. These results elucidate the development of an effective and simple process that selectively removes actinides from electrorefining salt, thus contributing to the minimization of the influx of actinides into the nuclear fuel waste stream.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1893-1908
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• 2021

In light of the importance of helium production in influencing the behaviour of fast reactor fuels, in this work we present a burn-up module with the objective to calculate the production of helium in both in-pile and out-of-pile conditions tracking the evolution of 23 alpha-decaying actinides. This burn-up module relies on average microscopic cross-section look-up tables generated via SERPENT high-fidelity calculations and involves the solution of the system of Bateman equations for the selected set of actinide nuclides. The results of the burn-up module are verified in terms of evolution of actinide and helium concentrations by comparing them with the high-fidelity ones from SERPENT, considering two representative test cases of (U,Pu)O₂ fuel in fast reactor conditions. In addition, a code-to-code comparison is made with the independent state-of-the-art module TUBRNP (implemented in the TRANSURANUS fuel performance code) for the same test cases. The herein presented burn-up module is available in the SCIANTIX code, designed for coupling with fuel performance codes.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.1
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• pp.33-41
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• 2022

Liquid Bi pool is a candidate electrode for an electrometallurgical process in the molten LiCl-KCl eutectic to treat the spent nuclear fuels from nuclear power plants. The electrochemical behavior of Bi³⁺ ions and the electrode reaction on liquid Bi pool were investigated with the cyclic voltammetry in an environment with or without BiCl₃ in the molten LiCl-KCl eutectic. Experimental results showed that two redox reactions of Bi³⁺ on inert W electrode and the shift of cathodic peak potentials of Li⁺ and Bi³⁺ on liquid Bi pool electrode in molten LiCl-KCl eutectic. It is confirmed that the redox reaction of lithium with respect to the liquid Bi pool electrode would occur in a wide range of potentials in molten LiCl-KCl eutectic. The obtained data will be used to design the electrometallurgical process for treating actinide and lanthanide from the spent nuclear fuels and to understand the electrochemical reactions of actinide and lanthanide at liquid Bi pool electrode in the molten LiCl-KCl eutectic.

• Journal of the Korean institute of surface engineering
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• v.56 no.1
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• pp.84-93
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• 2023

The spent nuclear fuel is burned during the planned cycle in the plant and then generates elements such as actinide series, fission products, and plutonium with a long half-life. An 'interim storage' step is needed to manage the high radioactivity and heat emitted by nuclides until permanent-disposal. In the case of Korea, there is no space to dispose of high-level radioactive waste after use, so there is a need for a period of time using interim storage. Therefore, the intensity of neutrons and gamma-ray must be determined to ensure the integrity of spent nuclear fuel during interim storage. In particular, the most important thing in spent nuclear fuel is burnup evaluation, estimation of the source term of neutrons and gamma-ray is regarded as a reference measurement of the burnup evaluation. In this study, an analysis of spent nuclear fuel was conducted by setting up a virtual fuel burnup case based on CE16×16 fuel to check the total amount and spectrum of neutron, gamma radiation produced. The correlation between BU (burnup), IE (enrichment), and CT (cooling time) will be identified through spent nuclear fuel burnup calculation. In addition, the composition of nuclide inventory, actinide and fission products can be identified.

• Proceedings of the Optical Society of Korea Conference
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• 2002.07a
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• pp.236-237
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• 2002

원자력 발전소에서 사용되는 연료의 효율을 증가시키기 위해서는 반응로에서 연소된 연료에 잔존하는 미량의 란탄족 (lanthanide)과 악티늄족 (actinide) 원소를 측정하고, 동위원소 성분비를 분석하는 작업이 필요하다. 이러한 미량원소 측정에는 흡수분광학 (absorption spectroscopy) 방법이 주로 사용되고 있다. 최근 들어 단일 종 모드 (single-longitudinal mode) 반도체 레이저가 개발되면서 미량원소 분석을 위한 분광 기술이 급속히 발전하고 있다. 반도체 레이저는 동작시킬 때 전력소모가 적고, 설치 공간이 작다는 점 때문에 현장 적용이 용이하므로 미량원소 측정 연구의 광원으로 많이 활용되고 있다. (중략)