• 전력전자학회:학술대회논문집
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• 전력전자학회 2004년도 전력전자학술대회 논문집(2)
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• pp.479-483
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• 2004

Fuel Cell (FC) has slow response characteristic for load variation. During a load step, the inverter cannot pull more power from the fuel cell than is currently available so supplemental power must be provide by some sort of energy storage elements. In this paper, hi-directional do-dc converter for FC generation system is proposed to improve load response characteristic. The hi-directional converter interfaces the low voltage battery to the inverter dc link of FC generation system. The converter is based on a active full bridge in the primary side and on a half bridge in the secondary of a high frequency isolation transformer. The complete operating principles and simulation results in presented.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2002년도 가을 학술발표회 논문집
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• pp.131-138
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• 2002

The fuel handling machine handles the core assembly in refueling period of the reactor, it is necessary to predict the motion and structural integrity of it. The kinetic analysis of the fuel handling machine was carried out for the refueling motion. The reaction forces at the joints of machine were calculated with IDEAS code considering the weight of the machine and the loading force of the core assembly, Also, the structural analysis for the machine modeled by lumped-mass and beam elements was performed by using ANSYS code. The stresses and deformations were calculated for the equivalent static force based on the kinetic analysis and the seismic loads. The calculated displacements and stresses are quite low compared with allowable limits.

• Nuclear Engineering and Technology
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• 제55권2호
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• pp.391-407
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• 2023

This article provides a survey of wet (aqueous) methods for recovery, separation, and purification of uranium from fission products in carbonate solutions during the reprocessing of spent nuclear fuel and methods for removal of radionuclides from alkaline radioactive waste. The main methods such as selective direct precipitation, ion exchange, and solvent extraction are considered. These methods were compared and evaluated for reprocessing of spent nuclear fuel in carbonate media according to novel alternative non-acidic methods and for treatment processes of alkaline radioactive waste.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.1013-1021
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• 2024

A series of experiments were performed to produce a fuel source for a molten salt reactor (MSR) through pyroprocessing technology. A simulated LiCl-KCl-UCl₃-NdCl₃ salt system was prepared, and the U element was fully recovered using a liquid cadmium cathode (LCC) by applying a constant current. As a result, the salt was purified with an UCl₃ concentration lower than 100 ppm. Subsequently, the U/RE ingot was prepared by melting U and RE metals in Y₂O₃ crucible at 1473 K as a surrogate for RE-rich ingot product from pyroprocessing. The produced ingot was sliced and used as a working electrode in LiCl-KCl-LaCl₃ salt. Only RE elements were then anodically dissolved by applying potential at - 1.7 V versus Ag/AgCl reference electrode. The RE-removed ingot product was used to produce UCl₃ via the reaction with NH4Cl in a sealed reactor.

• 한국수소및신에너지학회논문집
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• 제29권1호
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• pp.64-70
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• 2018

In this study, the kg-class Ni-Al alloy fabrication process at low temperature was developed from the physical mixture of Ni and Al powders. The AlCl3 as an activator was used to reduce the temperature of alloy synthesis below the melting temperature of Ni and Al elements (<$500^{\circ}C$). Mixed phase of Ni3Al intermetallic and Ni-Al solid-solution were identified in the XRD pattern analysis. Furthermore, from the analysis of SEM and particle size analyzer, we found that the particle size of synthesized alloy powders was not changed compared to the initial size of Ni particle after the formation of alloy powder at $500^{\circ}C$. In the creep test, the anode (which was fabricated by the prepared Ni-Al alloy powders in this study) displayed the enhanced creep resistance compared to the conventional anode.

• Journal of Advanced Marine Engineering and Technology
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• 제38권9호
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• pp.1057-1063
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• 2014

국제사회가 관심을 가지고 있는 이산화탄소 배출 감소를 위해 많은 연구 기관들이 연료 절감 방법을 연구하여 결과를 제시하고 있다. 비용 절감이 최종 결론이지만 아직까진, 어떻게 제시된 방법이 작동되는가, 어떠한 효과가 있는가, 어떠한 크기의 배에 가장 이상적이며 효과가 클까 등등 많은 의문점들이 제기되고 있다. 선주와 운영자들에게는 연료비가 가장 큰 비용 중 하나이다. 그리고 연료 절감은 선주나 운영자들에게는 비용 절감과 동시에 이산화탄소를 줄이는 가장 좋은 방법이다. 본 논문은 추진 효율을 높여 연료를 절감하는 방법 및 투자되는 비용과 일 년 동안 운항 후 얻어지는 연료비 절감을 통하여 투자금 회수 기간을 대략적으로 계산을 할 수 있도록 하고, 각 방법 간의 미치는 영향에 대하여 소개를 하고자 한다.

• Nuclear Engineering and Technology
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• 제46권2호
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• pp.147-158
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• 2014

In order to investigate how the microstructure of fuel/matrix-interaction (FMI) layers change during irradiation, different U-7Mo dispersion fuel plates have been irradiated to high fission density and then characterized using scanning electron microscopy (SEM). Specifially, samples from irradiated U-7Mo dispersion fuel elements with pure Al, Al-2Si and AA4043 (~4.5 wt.%Si) matrices were SEM characterized using polished samples and samples that were prepared with a focused ion beam (FIB). Features not observable for the polished samples could be captured in SEM images taken of the FIB samples. For the Al matrix sample, a relatively large FMI layer develops, with enrichment of Xe at the FMI layer/Al matrix interface and evidence of debonding. Overall, a significant penetration of Si from the FMI layer into the U-7Mo fuel was observed for samples with Si in the Al matrix, which resulted in a change of the size (larger) and shape (round) of the fission gas bubbles. Additionally, solid fission product phases were observed to nucleate and grow within these bubbles. These changes in the localized regions of the microstructure of the U-7Mo may contribute to changes observed in the macroscopic swelling of fuel plates with Al-Si matrices.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2010년도 추계 학술논문요약집
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• pp.223-224
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• 2010

• 한국대기환경학회지
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• 제34권4호
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• pp.534-541
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• 2018

Concentrations of total particulate matter (TPM), $PM_{10}$ and $PM_{2.5}$ were measured at three different sites based on each different fuel type (solid, liquid and gas) used in thermal power plants operating in Yeosu and Gwangyang National Industrial Complexes during 2017. The highest concentrations of TPM, $PM_{10}$, and $PM_{2.5}$ were observed at the solid fuel facility, and these values were $3.356mg/Sm^3$, $2.342mg/Sm^3$ and $1.834mg/Sm^3$, respectively. The ratio of $PM_{2.5}$ to TPM was the highest value of 54.6% in solid fuel case, and the lowest was 35.7% found in liquid fuel case. As a result of analyzing 9 kinds of metal compound with respect to each particle size, the metal concentration of TPM is higher than those of $PM_{10}$ and $PM_{2.5}$ in all fuel types. Total concentrations of metal elements in TPM by fuel difference are $1.2702mg/Sm^3$ in solid fuel, 0.0603 mg/Sm3 in liquid fuel, and $0.0733mg/Sm^3$ in gas fuel, respectively. Relatively higher total metal concentration in gas fuel than in liquid fuel was found; and this could be higher Cr and Al concentrations in use of gas fuel. As a result of estimating the emission factors of each facility, in case of solid fuel, TPM emissions per electricity production were found to be 0.7080 kt/PJ, followed by liquid fuel and gas fuel. $PM_{10}$ and $PM_{2.5}$ emissions per hour of electricity production were similar to those of TPM.

• 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.