• 한국세라믹학회지
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• 제29권8호
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• pp.609-616
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• 1992

This study has been investigated on the milling of Aluminium Distearate (ADS) powder and characteristics of the ADS-doped UO2 pellets. As-received ADS powder of the agglomerated particles has not shown any milling effect because of heat generated during planetary milling. But the use of coolant to effectively remove heat generated during milling has been found an effective way in breaking up the agglomerates of ADS powder. The green density of the UO2 pellet decreases with the amount of ADS powder doped. Therefore, in order to get the sintered density of 95% pellet decreases with the amount of ADS powder doped. Therefore, in order to get the sintered density of 95% theoretical density, the 200 ppm ADS-doped UO2 pellet has to be pressed under higher compacting pressure of 3500~4000 kgf/$\textrm{cm}^2$ compared with the ADS-undoped UO2 pellet pressed under around 3000 kgf/$\textrm{cm}^2$. The ADS-dpoed UO2 pellet with even relatively low sintered density of 10.27 g/㎤ exhibits open porosity of 1% while open porosity of the ADS-undoped UO2 pellet is reduced to around 1% only after its sintered density increases to 10.43g/㎤. It is, therefore, concluded that doping of ADS powder significantly contributes to the decrease in open porosity of the UO2 pellet. The dilatometry of the ADS doped UO2 pellet shows the sintering rate curve with the bimodal mode, which could be attributed to a phase reaction between UO2 and ADS. The X-ray diffraction analysis indicates that there occurs not any new phase formed but the shift of the peaks. It would be expected that a phase reaction resulting in solid solution would happen in the temperature range of 130$0^{\circ}C$ to 150$0^{\circ}C$ between UO2 and ADS.

• Nuclear Engineering and Technology
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• 제41권2호
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• pp.155-162
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• 2009

Since the 90's, EDF and AREVA-NP have irradiated, up to very high burnups, lead assemblies housing $M5^{(R)}$ cladded fuels. Post-irradiation examination of high burnup $UO_2$ pellets show an increase in the fission-gas release rate, an increase in fuel swelling, and formation of fission-gas bubbles throughout the pellets. Xenon abundances were quantified, and phenomena leading to this bubble formation were identified. All examinations provided valuable data on the complex state of the fuel during irradiation. They show the good behavior of these fuels, exhibiting various microstructures at very high burnups, none of which is likely to lead to problems during irradiation.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1541-1548
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• 2022

The thermal conductivity of uranium oxide (UO₂) containing pores and grain boundaries is investigated using continuum-level simulations based on the finite-difference method in two and three dimensions. Steady-state heat conduction is solved on microstructures generated from the phase-field model of the porous polycrystal to calculate the effective thermal conductivity of the domain. The effects of porosity, pore size, and grain size on the effective thermal conductivity of UO₂ are quantified. Using simulation results, a new empirical model is developed to predict the effective thermal conductivity of porous polycrystalline UO₂ fuel as a function of porosity and grain size.

• Nuclear Engineering and Technology
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• 제37권3호
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• pp.287-292
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• 2005

[ $\pi$ ]The effects of silica-based additives have been investigated to improve the creep property of a $UO_2$ pellet. The additive composition, $50wt\%SiO_2-47wt{\%}CaO-3wt\%Cr_2O_3$ (SCC), was selected according to the dihedral angle and the distribution of the second phase. It was observed that the creep rate of the $0.07 wt\%$ SCC-added $UO_2$ was slower than that of the pure $UO_2$. However, the creep rate of the $0.22 wt\%$ SCC-added $UO_2$ was about 3_48 times faster than that of the pure $UO_2$, depending on the applied stress in the lower stress range. In the case of the $0.35 wt\%$ SCC-added $UO_2$, the creep rate decreased in comparison with that of the $0.22 wt\%$ SCC-added $UO_2$. The observed enhancement in the creep rate might depend on a balance between the positive role of the viscous intergranular phase and the negative roles of the additives and the grain growth.

• 에너지공학
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• 제15권1호
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• pp.8-13
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• 2006

열중량 분석기와 XRD를 이용하여 $573\sim873K$의 온도범위와 이론밀도의 $94.64\sim99.10%$범위에서 $UO_2$ 핵연료 소결체의 공기 중 산화실험을 수행하였다. XRD를 이용하여 실험온도 범위에서 $UO_2$는 $U_3O_8$으로 산화되는 것을 확인하였다. 시간에 따른 산화반응은 S자의 형태를 나타내고 있어 핵생성과 성장의 과정을 따르는 것으로 나타났다. 밀도가 증가함에 따라 산화속도는 떨어지고, 산화유도시간은 늘어나는 것으로 나타났다. $UO_2$에서 $U_3O_8$으로의 산화에 대한 활성화 에너지는 $573\sim723K$의 온도범위에서 약 89.54kJ/mol, $723\sim873K$의 온도범위에서는 34.40 kJ/mol로 나타났다.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 요약집 춘계학술발표대회
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• pp.309.1-309.1
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• 2004

• Nuclear Engineering and Technology
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• 제22권1호
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• pp.75-81
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• 1990

핵연료봉내의 온도 분포를 결정하는데 있어서 중요한 핵연료소자와 피복판 사이의 접촉 열전도도를 결정하기 위한 실험을 수행하였다. 이 실험에 사용된 측정장치는 접촉압력을 임의로 변화시켜 줄 수 있는 가압기와 열전대, 진공펌프, 핵연료소자, 봉형태의 피복관, 그리고 두 개의 히터 등으로 구성되어 있다. 접촉 열전도도는 $UO_2$ 소자와 Zircaloy-2 피복관 사이의 접촉 압력과 표면 조도를 변화시키면서 측정하였다. 그 결과 두 물체사이의 접촉압력이 증가함에 따라, 그리고 표면이 매끄러울수록 접촉 열전달계수는 증가하였다. 실험에서 얻은 값을 가지고 상관식을 만들었으며 일반적으로 사용되고 있는 상관식과 비교하였다.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.1-12
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• 2019

One of proposed Accident Tolerant Fuel (ATF) concept is fully ceramic micro-encapsulated fuel (FCMF). FCMF using uranium mononitride (UN) has better safety aspects than $UO_2$ pellet fuel although it might not have a better neutronic performance due to the presence of matrix and high neutron-induced interaction of $^{14}N$. Before implementing UN-FCMF technology in Small-PWR, further research must be taken place to make sure the proposed design of fuel assembly has inherent safety features and maintain the fuel performance. This study focusses on the neutronic analysis of UN-FCMF based fuel assembly using Serpent and SCALE codes. It is shown in the proposed fuel assembly design has inherent safety features with respect to the fuel temperature reactivity coefficient, void reactivity coefficient, and moderator temperature reactivity coefficient. It is noted that the use of FCMF leads to a lower ratio of burnup to $^{235}U$ enrichment ratio compared to the $UO_2/Zr$ fuel.

• 자원리싸이클링
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• 제11권4호
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• pp.3-10
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• 2002

핵연료 원료인 $UO_2$ 분말을 사용해 원자로에서 연소된 사용후 핵연료 소결체를 모의 제조하여 1회 산화ㆍ환원처리하여 분말로 만든 후, 건ㆍ습식 attrition 분쇄에 따른 분말의 특성 및 소결성을 조사하였다. 분쇄에 의한 분말의 평균입자크기는 건식분쇄의 경우에는 1 $mu extrm{m}$ 이하인 미분말이 쉽게 생성되었으나, 습식분쇄에서는 그 이상의 분말만이 생성되었다. 그리고 분쇄분말의 비표면적은 건식분쇄한 경우가 습식분쇄한 경우 보다 높았다. 분말의 미세구조는 건식분쇄에 의해서는 느슨한 응집체가 형성되었으며, 습식분쇄 분말은 압분성이 낮은 불규칙적이고 각진 입자형태를 나타내었다. 건식분쇄에 의해서 압분체 밀도는 크게 증가하며 소결체 요구 조건을 만족하는 이론밀도의 95%이상이 되고 평균 결정립 크기가 8 $\mu\textrm{m}$이상인 소결체를 얻을 수 있었다.

• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1109-1114
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• 2009

The evolution of a high burnup structure (HBS) in a light water reactor (LWR) $UO_2$ fuel was simulated using the Potts model. A simulation system for the Potts model was defined as a two-dimensional triangular lattice, for which the stored energy was calculated from both the irradiation damage of the $UO_2$ matrix and the formation of a grain boundary in the newly recrystallized small HBS grains. In the simulation, the evolution probability of the HBS is calculated by the system energy difference between before and after the Monte Carlo simulation step. The simulated local threshold burnup for the HBS formation was 62 MWd/kgU, consistent with the observed threshold burnup range of 60-80 MWd/kgU. The simulation revealed that the HBS was heterogeneously nucleated on the intergranular bubbles in the proximity of the threshold burnup and then additionally on the intragranular bubbles for a burnup above 86 MWd/kgU. In addition, the simulation carried out under a condition of no bubbles indicated that the bubbles played an important role in lowering the threshold burnup for the HBS formation, thereby enabling the HBS to be observed in the burnup range of conventional high burnup fuels.