• Nuclear Engineering and Technology
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• 제40권3호
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• pp.163-174
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• 2008

As part of the spent fuel treatment program at the Idaho National Laboratory, a vacuum distillation process is being employed for the recovery of actinide products following an electrorefining process. Separation of the actinide products from a molten salt electrolyte and cadmium is achieved by a batch operation called cathode processing. A cathode processor has been designed and developed to efficiently remove the process chemicals and consolidate the actinide products for further processing. This paper describes the fundamentals of cathode processing, the evolution of the equipment design, the operation and efficiency of the equipment, and recent developments at the cathode processor. In addition, challenges encountered during the processing of irradiated spent nuclear fuel in the cathode processor will be discussed.

• Nuclear Engineering and Technology
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• 제46권6호
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• pp.863-868
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• 2014

The evaluation of thermal properties of actinide oxide fuels is a problem of high importance for the development of new generation reactors. In the present study, an expression obtained for n-dimensional Debye functions is used to derive a simple analytical expression for the specific heat capacity of nuclear fuels. To test the validity and reliability of this expression, the analytical expression is applied to $UO_2$, $NpO_2$, $ThO_2$, and $PuO_2$. It is seen that the formula was in agreement with the experimental and theoretical results reported in the literature.

• Nuclear Engineering and Technology
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• 제31권5호
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• pp.498-505
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• 1999

Trivalent actinide-lanthanide group separation is difficult to perform on an industrial scale, because of the many drawbacks of the available chemical process. In this paper, picolinamide(C$_{8}$H$_{17}$) is synthesized and characterized, and extraction yields of Am-241, Eu-152 and Nd are determined in batch extraction experiments. In particular, the influence of the solvent is described. The extraction yields of Am-241, Eu-152 and Nd depended on the LiNO$_3$ concentration, the picolinamide(C$_{8}$H$_{17}$) concentration and the acidity. A favorable picolinamide(C$_{8}$H$_{17}$) concentration was found to be about 2M. The appropriate nitric acid concentration and LiNO$_3$ concentration were confirmed to be about 0.125M and 3M, respectively. The separation factor of Am and Eu was about 9.9 at optimum conditions. The picolinamide(C$_{8}$H$_{17}$) is a very promising extractant for the actinide(III)-lanthanides(III) separation.aration.aration.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2005년도 Proceedings of The 6th korea-china joint workshop on nuclear waste management
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• pp.86-91
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• 2005

A metal product obtained from an electrolytic reduction process, possesses less volume and radioactivity than those of the unprocessed spent oxide fuels. The chemical composition of the metal product varies according to the process condition. In this work, a basic study was performed to evaluate the chemical forms of the spent oxide fuel components in an electrolytic reduction process with the operation conditions. One of the most important operation conditions is the cell potential applied for the reduction cell. It is expected that $PU_{2}O_3$ is difficult to reduce even though the cell potential is negative enough to reduce the lithium oxide when the activity of $Li_{2}O$ exceeds 0.003. The reduction of actinide oxides via the reduction of $Li_{2}O$ is assumed to have a greater reduction yield than a direct reduction of the actinide oxides.

• 한국표면공학회지
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• 제56권1호
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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.

• 한국광물학회지
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• 제15권1호
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• pp.78-84
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• 2002

New pyrochlore-type phases($A_2$$B_2$$O_{7}$) were synthesized in the systems: CaO-C$eO_2$-T$iO_2$, CaO-$UO_2$(T$hO_2$)-Z$rO_2$, CaO-$UO_2$(T$hO_2$)-$Gd_2$$O_3$-T$iO_2$-Z$rO_2$, 및 CaO-T$hO_2$-S$nO_2$. The starting materials were pressed with the pressure of 200~400 MPa and sintered at 1500~ 155$0^{\circ}C$ for 4~8 hours in air and at 1300~ 135$0^{\circ}C$ for 5 ~50 hours under oxygen atmosphere. The products were characterized using XRD, SEM/EDS and TEM. In the bulk compositions of CaCe$Ti_2$$O_{7}$, CaTh$Zr_2$$O_{7}$,($Ca_{0.5}$ Gd$Th_{0.5}$)(ZrTi)$O_{7}$) ($Ca_{0.5}$Gd$Th_{0.5}$)(ZrTi)$O_{7}$, ($Ca_{0.5}$G$dU_{0.5}$)(ZrTi)$O_{7}$ and CaTh$Sn_2$$O_{7}$ , pyrochlore was the major phase, together with other oxide phase $of_2$$O_{7}$ fluorite structure. In the samples with target compositions CaU$Zr_2$$O_2$및 $Ca_{0.5}$ G$dU_{0.5}$)$Zr_2$T$iO_{7}$ pyrochlore was not identified, but a fluorite-structured phase was detected. The formation factor as the stable phase depended on crystal chemical characteristics of the actinide and lanthanide elements of the system concerned.

• 방사성폐기물학회지
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• 제20권1호
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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.

• 방사성폐기물학회지
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• 제18권3호
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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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• 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.

• 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.