• Nuclear Engineering and Technology
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• 제45권7호
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• pp.839-846
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• 2013

This paper describes a finite element model of the microstructure of dispersion type nuclear fuels, which can be used to determine the effective thermal conductivity of the fuels during irradiation. The model simulates a representative region of the fuel as a prism shaped unit cell made of brick elements. The elements within the unit cell are assigned material properties of either the fuel or the matrix depending on position, in such a way as to represent randomly distributed fuel particles with a size distribution similar to that of the as manufactured fuel. By applying an appropriate heat flux across the unit cell it is possible to determine the effective thermal conductivity of the unit cell as a function of the volume fraction of the fuel particles. The presence of a fuel/matrix interaction layer is simulated by the addition of a third set of material properties that are assigned to the finite elements that surround each fuel particle. In this way the effective thermal conductivity of the material may also be determined as a function of the volume fraction of the interaction layer. Work is on going to add fission gas bubbles in the fuel as a fourth phase to the model.

• 항공우주시스템공학회지
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• 제13권3호
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• pp.9-14
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• 2019

현대에 사용하는 다양한 종류의 액체 및 고체연료는 각각 장단점을 가지고 있으며, 이에 따라 많은 연구자들은 각 연료의 단점을 극복하고 장점만을 취하고자 새로운 형태의 연료를 연구하였다. 본 연구는 액체 에탄올을 고형화 하는 공정을 개발하고, 제조된 연료의 기초 물성치 및 연소특성을 관찰하는데 그 목적이 있다. 고형 에탄올은 아가로스 하이드로젤을 제조하고 이를 에탄올에 침전시키는 방법으로 제조하였다. 실험 조건으로 제조된 고형 에탄올 연료의 정성적/정량적 특성을 관찰하였으며, 이를 통해 제조된 연료의 유효성 및 고형 에탄올 연료의 실제 활용 가능성을 고찰하였다.

• 한국세라믹학회지
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• 제40권5호
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• pp.410-414
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• 2003

Characteristics of $UO_{2+x}$ powders oxidized at different temperatures were examined. Pellets were fabricated by adding these oxidation powders and their properties were also investigated. Particle size of the $UO_{2+x}$ powders decreased with increasing oxidation temperature while surface area increased. Only the powders oxidized at 35$0^{\circ}C$ enhanced the strength of green pellet. However, 35$0^{\circ}C$ oxidized powders added pellet had many surface defects. The difference of shrinkage rate between the oxidized and UO$_2$ powders was thought to be the cause of them.

• 에너지공학
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• 제24권4호
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• pp.166-174
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• 2015

본 연구에서는 물성 변화에 따라 자동차 성능 및 환경에 미치는 영향을 살펴보기 위하여 휘발유 대비 물성이 상이한 연료를 선정하여 연소특성을 살펴보았다. 실험은 각 연료에 맞는 점화시기 및 공연비제어를 수행하여 물성변화에 따른 최적의 점화시기와 희박연소에 에서의 연소특성을 살펴보았으며, 그에 따른 배출가스도 평가하였다. 실험을 위해 단기통 엔진을 사용하였으며, 휘발유 물성변화를 위해 "석유 및 석유대체사업법"에 고시된 품질기준을 벗어나는 가짜연료를 선정하여 실험을 수행하였다. 그 결과 선정된 연료의 경우 옥탄가와 증류성상, 증기압에서 차이를 보였으며, 불안정 연소 및 다량의 유해배출가스를 유발함을 알 수 있었다.

• 한국압력기기공학회 논문집
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• 제8권2호
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• pp.7-13
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• 2012

A spacer grid which is one of the most important structural components in a pressurized water reactor fuel is an interconnected array of slotted grid straps, welded at the intersections to form an egg-crate structure. The spacer grid is required to not only protect fuel rods stably but also have sufficient lateral crush strength for the sake of enabling shut-down of the nuclear reactor during abnormal operating environments. Then, the lateral crush strength of the spacer grid is closely related with welding quality of the spacer grid. Previous research on the crush strength analysis of the spacer grid had been performed using only parent material properties. In this study, to investigate the effect on the crush strength of the spacer grid when used mechanical properties in weld zone instead of parent material properties, crush strength analysis considering mechanical properties in weld zone obtained from the instrumented indentation technique was performed and compared the results with the previous research.

• 대한금속재료학회지
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• 제47권4호
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• pp.209-216
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• 2009

It was evaluated that the microstructural and mechanical properties of Ta-bearing 9Cr-0.5Mo-2W ferritic/martensitic experimental steels. All the experimental steels showed the tempered martensitic microstructures, and $M_{23}C_6$ carbides, whose sizes were ranged from 200 to 300 nm, were easily observed at both boundaries of the prior austenite grain and the martensite lath. In addition, a relatively large Nb-rich MX carbonitrides were intermittently detected at the prior austenite grain boundaries, whereas a lot of Vrich MX carbonitrides, whose mean diameter was less than 50 nm, were observed randomly at both boundaries. Ta was mainly incorporated into the V-rich MX carbonitrides rather than the Nb-rich ones and their content was spanned from 5 to 20 at.%. Ta contents within the MX precipitates also increased as the content of Ta increased. Because the Ta addition into the steels would be attributed to the precipitation strengthening, solid solution strengthening and lath width reduction, it was shown that the mechanical properties, including hardness, tensile strength and creep rate of the 9%Cr-0.5Mo-2W steels were improved by the increase of Ta content. Especially, 9Cr-0.5Mo-2W-0.3V-0.05Nb-0.14Ta steel was revealed to be relatively excellent in the application for the SFR fuel cladding.

• 대한기계학회논문집B
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• 제40권11호
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• pp.697-704
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• 2016

본 논문에서는 역청탄인 Glencore 탄을 염산과 질산수용액을 이용하여 산 처리하고 원탄과 산 처리 된 석탄의 물리, 화학적 비교분석과 직접탄소 이용 연료전지(Direct Carbon Fuel Cell, DCFC)의 성능 비교 분석을 수행하였다. 석탄의 물성들을 분석하기 위해 열중량 분석과 가스 흡착법, X선 광전자 분광법을 수행하였다. 열중량 분석을 통해 연료의 열적 반응성이 증가하였음을 알 수 있었고, 가스 흡착법 결과로 기공의 평균지름은 변화가 없었지만 표면적은 감소함을 알 수 있었다. X선 광전자 분광법에서는 $HNO_3$ 처리의 경우 가장 높은 산소/탄소 비율을 보였고, 이를 통해 다양한 표면 산소작용기가 증가한 것을 확인하였다. 연료의 표면 물성과 전기화학 성능을 비교한 결과, 표면의 산소 성분의 변화가 DCFC의 성능 향상에 가장 큰 영향을 미침을 알 수 있었다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.905-906
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.515-516
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• 2005

• Nuclear Engineering and Technology
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• 제39권5호
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• pp.617-626
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• 2007

A centrifugal atomization process for uranium fuel was developed in order to fabricate high uranium density dispersion fuel for advanced research reactors. Spherical powders of $U_3Si$ and U-Mo were successfully fabricated and dispersed in aluminum matrices. Thermal and mechanical properties of dispersion fuel meat were characterized. Irradiation tests at the research reactor HANARO confirm the excellent performance of high uranium density dispersion fuel.