• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1999.05a
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• pp.225-228
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• 1999

MCFC의 장수명화를 위해 기존의 Li$_2$CO$_3$-K$_2$CO$_3$계 전해질을 Li$_2$CO$_3$-$Na_2$CO$_3$계 용융탄산염으로 대체함으로써, 전지 수명을 단축시키는 NiO의 용출을 억제하고자하는 연구가 진행중에 있다. 이러한 대체 Li$_2$CO$_3$-$Na_2$CO$_3$ 전해질은 실제 전지에서 사용되고 있는 분리판 재료인 스테인레스강의 안정성에도 기존의 Li$_2$CO$_3$-K$_2$CO$_3$ 혼합염과는 다른 경향을 보이는 것으로 알려져 있다.(중략)

• Proceedings of the Korean Institute of Surface Engineering Conference
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• 2013.05a
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• pp.116-116
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• 2013

(Ni-22Cr-10Al-1Y)와 ($ZrO_2-25CeO_2-2.5Y_2O_3$)로 구성되는 금속/세라믹 복합코팅을 대기용사(ASP; air plasma spay)으로 철 기판위에 1:3, 2:2, 3:1의 무게비로 혼합하여 제조하였다. 용사된 코팅은 금속이영지역과 세라믹잉여지역으로 구별되고, 용사중에 NiCrAlY중의 Al이 선택적으로 산화되어 Al2O3가 계면에 존재하였다. 복합코팅은 $NaCl-Na_2SO_4$ 용융염에서 $800{\sim}900^{\circ}C$, 50시간 동안 부식실험을 실시하였다. 부식생성물은 NiO, $Cr_2O_3$, ${\alpha}-Al_2O_3$가 생성되는데, 부식이 진행되면서 용해되었다. 용융염 부식이 진행되는 동안에 Cr, Al이 외방확산하였고, Na, Cl, S는 내부로 확산되었다. 시간 및 온도뿐만 아니라 금속의 양이 증가할수록 코팅의 내식성은 저하되었다.

• Journal of the Korean Electrochemical Society
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• v.18 no.3
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• pp.121-129
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• 2015

The platinum anode for the electrolytic reduction process is generally surrounded by a nonporous ceramic shroud with an open bottom to offer a path for $O_2$ gas produced on the anode surface and prevent the corrosion of the electrolytic reducer. However, the $O^{2-}$ ions generated from the cathode are transported only in a limited fashion through the open bottom of the anode shroud because the nonporous shroud hinders the transport of the $O^{2-}$ ions to the anode surface, which leads to a decrease in the current density and an increase in the operation time of the process. In the present study, we demonstrate the electrolytic reduction of 1 kg-uranium oxide ($UO_2$) using the porous shroud to investigate its long-term stability. The $UO_2$ with the size of 1~4mm and the density of $10.30{\sim}10.41g/cm^3$ was used for the cathode. The platinum and 5-layer STS mesh were used for the anode and its shroud, respectively. After the termination of the electrolytic reduction run in 1.5 wt.% $Li_2O-LiCl$ molten salt, it was revealed that the U metal was successfully converted from the $UO_2$ and the anode and its shroud were used without any significant damage.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.16-25
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• 2016

In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconiumalloy fuel claddingmaterials are rapidlyheateddue to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF) design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI) is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in $1,200-1,500^{\circ}C$ steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstratedcorrosionresistance.Asthese composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Moalloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are discussed in this document. In addition to assisting plants in meeting Light Water Reactor (LWR) challenges, accident-tolerant Mo-based cladding technologies are expected to be applicable for use in high-temperature helium and molten salt reactor designs, as well as nonnuclear high temperature applications.