• Journal of the Korean Ceramic Society
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• v.29 no.8
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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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• v.41 no.2
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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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• v.54 no.5
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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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• v.37 no.3
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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.

• Journal of Energy Engineering
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• v.15 no.1 s.45
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• pp.8-13
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• 2006

The oxidation behavior of $UO_2$, pellet was studied using an XRD and a thermogravimetric analyzer in the temperature range from 573 to 873 K and in the density range from 94.64 to 99.10% of theoretical density in air. It was found from the XRD study that $UO_2$ was completely converted to $U_3O_8$ in this experimental temperature range. The formation of $U_3O_8$ displays sigmoidal reaction kinetics. The oxidation rate was reduced with density. Induction time for the oxidation of $UO_2$ was delayed with density because of open pore formed in surface of $UO_2$ pellet. The activation energy for oxidation of $UO_2$ was determined to be 89.54 kJ/mol and 34.40 kJ/mol in the temperature range from 573 to 723 K and from 723 to 873 K, respectively.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.309.1-309.1
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• 2004

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.75-81
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• 1990

Experiments to predict the thermal contact conductance between the fuel pellet and cladding have been performed, which is important to determine the temperature distibution within the fuel rod. UO$_2$and Zircaloy-2 are used in these experiments. The measuring apparatus is composed of a presser which controls the contact pressure, a thermometer with 5.5 sheathed thermocouples, a vacuum pump, pellet and cladding rods, and two heating devices, etc. The thermal contact conductances were measured with varying the contact pressure and surface roughnesses of UO$_2$and Zircaloy-2 bars. The results show that an increase in the contact pressure and a decrease of surface roughness resulted in increase of the thermal contact conductance. Finally, a fitting correlation has been established and compared with widely-used correlations.

• Nuclear Engineering and Technology
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• v.51 no.1
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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.

• Resources Recycling
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• v.11 no.4
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• pp.3-10
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• 2002

The characteristics of dry and wet milled powder prepared by 1 cycle OREOX (oxidation and reduction of oxide fuels) treatment were investigated using the simulated spent fuel pellet. Sintered pellets simulating spent nuclear fuel burned in reactor were fabricated from $UO_2$ powder using as a starting material in fabrication of nuclear fuel. The 1 cycle OREOX-treated powder was prepared by only one path of oxidation md reduction of the simulated pellet. Powder having average particle size of less than 1 $\mu\textrm{m}$ could be easily obtained by dry milling, but not be achieved by wet milling. And, specific surface area of dry milled pow-der was higher than that of wet milled powder. Dry milled powder formed loose agglomerate, while wet milled powder showed the shape of irregular and angular particles. Dry milled powder provided higher green density, resulting in higher sintered density of higher than 95% TD and average grain size of larger than 8 $\mu\textrm{m}$ satisfying the standard specification of sintered pellets.

• Nuclear Engineering and Technology
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• v.41 no.8
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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.