• Nuclear Engineering and Technology
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• 제37권5호
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• pp.433-446
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• 2005

The development of accelerator science and technology has been accommodating ever increasing demand from scientific community of the beam energy and intensity of proton beams. The use of high-powered proton beams has extended from the traditional application of nuclear and high-energy physics to other applications, including spallation neutron source replacing nuclear reactor, nuclear actinide transmutation, energy amplification reactors. This article attempts to review development of proton accelerator, both linear and circular, and issues related to the proton beam energy, intensity as well as its output power. For related accelerator physics and technical review, one should refer to the recent article in the Reviews of Modem Physics [1]

• Nuclear Engineering and Technology
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• 제43권4호
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• pp.329-334
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• 2011

Two conceptual flowsheets were developed for recycling used nuclear fuel. One flowsheet was developed for recycling used oxide nuclear fuel from light water reactors while the other was developed for recycling used metal fuel from fast spectrum reactors. Both flowsheets were developed from a set of design principles including efficient actinide recovery, nonproliferation, waste minimization and commercial viability. Process chemistry is discussed for each unit operation in the flowsheet.

• Nuclear Engineering and Technology
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• 제34권5호
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• pp.517-531
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• 2002

The speciation and solubility of Am, Np, Pu and U have been analyzed by means of the geochemical code MUGREM, under the chemical conditions of domestic deep groundwater, in order to support the preliminary safety assessment for a Korean HLW disposal concept. Under the conditions of groundwaters studied, the stable solid phase is AmOHC $O_3$(s) or Am(OH)$_3$(s), soddyite((U $O_2$)$_2$ $SiO_2$.2$H_2O$) or N $a_2$ $U_2$ $O_{7}$ (c), Np(OH)$_4$(am), and Pu(OH)$_4$(am) for Am, U, Np, and Pu, respectively. The dominating aqueous species are as follows: the complexes of Am(III), Am(OH)$_2$$^{+}$ and Am(C $O_3$)$_2$$^{[-10]}$ , the complexes of U(VI), U $O_2$(OH)$_3$$^{[-10]}$ and U $O_2$(C $O_3$)$_3$$^{4-}$, the complexes of Np(IV), Np(OH)$_4$(aq) and Np(OH)$_3$C $O_3$, and the complexes of Pu(IV), Pu(OH)$_4$(aq) and Pu(OH)$_3$C $O_3$$^{[-10]}$ . The calculated solubilities exist between 1.9E-10 and 1.3E-9 mol/L for Am, between 5.6E-6 and 1.2E-4 mol/L for U, between 3.1E-9 and 1.3E-8 mol/L for Np, and between 6.6E-10 and 2.4E-10 mol/L for Pu, depending on groundwater conditions. The present solubilities of each actinide agree well with the results of other studies obtained under similar conditions.s.

• Nuclear Engineering and Technology
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• 제42권1호
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• pp.79-88
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• 2010

Spent $UO_2$ nuclear fuel discharged from a nuclear power plant (NPP) contains fission products, U, Pu, and other actinides. Due to neutron capture by $^{238}U$ in the rim region and a temperature gradient between the center and the rim of a fuel pellet, a considerable increase in the concentration of fission products, Pu, and other actinides are expected in the pellet periphery of high burnup fuel. The characterization of the radial profiles of the various isotopic concentrations is our main concern. For an analysis, spent nuclear fuels originating from the Yeonggwang-2 pressurized water reactor (PWR) were chosen as the test specimens. In this work, the distributions of some actinide isotopes were measured from center to rim of the spent fuel specimens by a radiation shielded laser ablation inductively coupled plasma mass spectrometer (LA-ICP-MS) system. Sampling was performed along the diameter of the specimen by reducing the sampling intervals from 500 ${\mu}m$ in the center to 100 ${\mu}m$ in the pellet periphery region. It was observed that the isotopic concentration ratios for minor actinides in the center of the specimen remain almost constant and increase near the pellet periphery due to the rim effect apart from the $^{236}U$ to $^{235}U$ ratio, which remains approximately constant. In addition, the distributions of local burnup were derived from the measured isotope ratios by applying the relationship between burnup and isotopic ratio for plutonium and minor actinides calculated by the ORIGEN2 code.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2019년도 춘계학술논문요약집
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• pp.69-69
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• 2019

The distillation of liquid cathode is necessary to separate cadmium from the actinide elements in the pyroprocessing since the actinide deposits are dissolved or precipitated in a liquid cathode. 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. One of the important methods is an installation of crucible cover on the distillation crucible. A multi-layer porous round cover was proposed to avoid a cadmium splattering in our previous study. In this study, the effect of crucible cover on the cadmium distillation was examined to develop a splatter shield. Various surrogates were used for the actinides in the cadmium. The surrogates such as bismuth, zirconia, and tungsten don't evaporate at the operational temperature of the Cd distiller due to their low vapor pressures. The distillation experiments were carried out in a crucible equipped with cover and in a crucible without cover. 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 heated at $470{\sim}620^{\circ}C$. Most of the bismuth remained in the crucible equipped with cover after distillation under $580^{\circ}C$ for two hours, whereas small amount of bismuth decreased in the crucible without cover above $580^{\circ}C$. The liquid bismuth escaped with liquid cadmium drop from the crucible without cover. It seems that the crucible cover played a role to prevent the splash of the liquid cadmium drop. The effect of the cover was not clear for the tungsten or zirconia surrogate since the surrogates remained as a solid powder at the experimental temperature. From the results of this work, it can be concluded that the crucible cover can be used to minimize the deposit loss by prevention of escape of liquid drop from the crucible during distillation of liquid cathode.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.257-262
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• 2019

The control of microstructure is critical for the porous fuel particles used for infiltrating actinide nuclides. This study concerns the effect of heating processes on properties and microstructure of the fuel particles. The uniform gel precursor beads were synthesized by a microfluidic sol-gel process and then the porous $CeO_2$ microspheres, as a surrogate for the ceramic nuclear fuel particles, were obtained by heating treatment of the gel precursors. The fabricated $CeO_2$ microspheres have a narrow size distribution and good sphericity due to the feature of microfluidics. The effects of heating processes parameters, such as heating mode and peak temperatures on the properties of microspheres were studied in detail. An optimized heating mode and the peak temperature of $650^{\circ}C$ were selected to produce porous $CeO_2$ microspheres. The optimized heating mode can avoid the appearance of broken or crack microspheres in the heating process, and as-prepared porous microspheres were of suitable pore size distribution and pore volume for loading minor actinide (MA) solution by an infiltration method that is used for fabrication of MA-bearing nuclear fuel beads. After the infiltration process, $1000^{\circ}C$ was selected as the final temperature to improve the compressive strength of microspheres.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2018년도 춘계학술논문요약집
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• pp.57-57
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• 2018

The electrorefining process is generally composed of two recovery steps in pyroprocessing - the deposit of uranium onto a solid cathode and the recovery of the remaining uranium and TRU elements simultaneously by a liquid cadmium cathode. The liquid cathode processing is necessary to separate cadmium from the actinide elements since the actinide deposits are dissolved or precipitated in a liquid cathode. Distillation process was employed for the cathode processing. It is very important to avoid a splattering of cadmium during evaporation due to the high vapor pressure. In this study, a multi-layer porous round cover was proposed and examined to develop a splatter shield for the Cd distillation crucible. Cadmium vapor can be released through the holes of the shield, whereas liquid drops can be collected in the multiple hemisphere. The collected drops flow on the round surface of the cover and flow down into the crucible. The crucible cover was fabricated and tested in the Cd distiller. 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. About 40 grams of Cd and about 4 grams of Bi was distilled at a reduced pressure for two hours at $470^{\circ}C$. After the Cd distillation experiment, cadmium was not detected and more than 90 % of Bi remained in the ICP-OES analysis. Therefore the crucible cover can be a candidate for the splatter shield of the Cd distillation crucible. Further development of the crucible cover is necessary for the decision of the optimum cover geometry and the operating conditions of the Cd distiller.

• 방사성폐기물학회지
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• 제16권3호
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• pp.369-376
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• 2018

The ability to recover the nearly limitless supply of uranium contained within the world's oceans would provide supply security to uranium based fuel cycles. Therefore, in addition to U.S. national laboratories conducting R&D on a system capable of harvesting seawater uranium, a number of collaborative university partners have developed alternative technologies to complement the national laboratory scheme. This works summarizes the systems analysis of such novel uranium recovery technologies along with their potential impacts on seawater uranium recovery. While implementation of some recent developments can reduce the cost of seawater uranium by up to 30%, other researchers have sought to address a weakness while maintaining cost competitiveness.

• 한국자기공명학회논문지
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• 제5권1호
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• pp.13-18
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• 2001

Natural alpha radiation produced a stable defect center to certain minerals. Electron Paramagnetic Resonance(EPR) spectroscopy is a powerful tool f3r quantifying this defect center. EPR method has been applied to trace alpha-radiation effect around the uranium ore deposit. The results show that EPR technique can be used to measure rapidly and nondestructively the defect center produced by natural alpha radiation. In general, a good correlation was achieved between defect center concentration and actinide elements(U, Th). These results imply that the concentration of defect center is dependent on the alpha radiation dose over long time scale.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(3)
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• pp.373-378
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• 1996

알칼리 및 알칼리토금속(Cs,Rb,Ba,Sr), 전이감속원소(Zr,Fe,Mo,Ni,Pd,Rh), 란탄족 (La,Y,Nd,Ce,Eu.) 및 MA(Np,Am)등 17개 원소로 구성된 질산매질의 모의 방사성용액에서 옥살산에 의한 란탄족과 MA(Minor Actinide)의 공침전 연구를 수행하였다. 옥살산농도 0.5M에서 질산농도의 영향과 아스코빅산 첨가에 따른 원소들의 침전율이 조사되었다. 각 원소들의 침전율은 질산농도에 따라 약간 감소하였으나 란탄족과 MA는 99%이상 공침전되었다. 아스코빅산이 첨가되는 경우 Pd이 금속으로 환원침전되고 Mo.Fe,Ni.Ba의 경우는 침전율이 10∼20% 감소하는 것으로 나타났으나 기타원소들에 대해서는 영향이 나타나지 않았다. Pd의 환원침전은 질산농도 1.0M미만에서 일어났으며. 아스코빅산 농도가 0.01M∼0.02M 부근에서 최대로 나타났다. 하이드라진이 아스코빅산과 같이 첨가될 때 Pd의 환원침전을 억제하는 역할을 하였다.