• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1756-1763
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• 2020

The paper studies spark plasma sintering (SPS) of industrially used UO₂-based fuel containing integral fuel burnable absorber (IFBA) of neutrons Gd₂O₃. Densification dynamics of pristine UO₂ powder and the one added with 2 and 8 wt% of Gd₂O₃ under ultrasonication in liquid has been studied under SPS conditions at 1050, 1250, and 1450 ℃. Effect of sintering temperature on phase composition as well as on O/U stoichiometry has been investigated for UO₂ SPS ceramics. Sintering of uranium dioxide added with Gd₂O₃ yields solid solution (U,Gd)O₂, which is isostructural to UO₂. SEM with EDX and metallography were implemented to analyze the microstructure of the obtained UO₂ ceramics and composite UO₂-Gd₂O₃ one, particularly, open porosity, defects, and Gd₂O₃ distribution were studied. Microhardness, compressive strength and density were shown to reduce after addition of Gd₂O₃. Obtained results prove the hypothesis on formation of stable pores in the system of UO₂-Gd₂O₃ due to Kirkendall effect that reduces sintering efficiency. The paper expands fundamental knowledge on pros and cons of fuel fabrication with IFBA using SPS technology.

• Nuclear Engineering and Technology
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• v.19 no.1
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• pp.22-33
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• 1987

The confirmed integrity of nuclear fuel cladding materials is an important object during steady state and transient operations at nuclear power plant. In this context, the clad material yielding behavior is especially important because of pellet-clad gap expansion. During the steep power excursion, the in-pile irradiation behavior differences between uranium-dioxide fuel pellet and zircaloy clad induce the contact pressure between them. If this pressure reaches the zircaloy clad yield pressure, the zircaloy clad will be plastically deformed. After the reactor power resumed to normal state, this plastic permanent expansion of clad tube give rise to the pellet-clad gap expansion. In this paper, the simple mandrel expansion test method which utilizes thermal expansion difference between copper mandrel and zircaloy tube was adopted to simulate this phenomenon. That is, copper mandrel which has approximately three times of thermal expansion coefficient of zircaloy-4 (PWR fuel cladding material) were used in this experiment at the temperature range from 400C to 700C. The measured plastic expansion of zircaloy outer radius and derived mathematical relations give the yield pressure, yield stress of zircaloy-4 clad at the various clad wall temperatures, the activation energy of zircaloy tube yielding, and pellet-clad gap expansion. The obtained results are in good agreement with previous experimental results. The mathematical analysis and simple test method prove to be a reliable and simple technique to assess the yielding behavior and gap expansion measurement between zircaloy-4 tube and uranium-dioxide fuel pellet under biaxial stress conditions.

• Nuclear Engineering and Technology
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• v.15 no.3
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• pp.197-206
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• 1983

The slightly hyperstoichiometric uranium dioxide, i.e. U $O_{2.005}$ and U $O_{2.01}$ within a range of the requirement for the use of a nuclear fuel, were sintered directly in an atmosphere of $CO_2$/CO mixture without any succeeding reduction process. The kinetics of sintering in the late stage were investigated for various O/U ratios. A sintering diagram, which show the relation of Temperature-Time-Density-Grain size, was established for each O/U ratio. Only by controlling the oxygen partial pressure in the sintering atmosphere, U $O_2$ pellet could be sintered very easily at low temperature 1050$^{\circ}$～120$0^{\circ}C$ with a density above 95% T.D. and average grain size above 7${\mu}{\textrm}{m}$. It was found that the rate of grain growth follows D=(Kt)$^{1}$4/ in the late stage of sintering. And the activation energies for grain growth in the final sintering stage were found to be 75, 64 and 62kca1/mo1 for U $O_{2.005}$, U $O_{2.01}$ and U $O_{2.10}$, respectively. Although no significant differences are obtained between the activation energies for different O/U ratios, the sinterability is enhanced considerably with increasing the oxygen partial pressure in the sintering atmosphere.tmosphere.

• Nuclear Engineering and Technology
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• v.31 no.2
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• pp.133-138
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• 1999

Research on the etching reaction of UO$_2$ with CF$_4$/O$_2$gas mixture plasma is carried out. The reaction rates are investigated as a function of CF$_4$/O$_2$ ratio, plasma power, and substrate temperature. It is found that there exists an optimum CF$_4$/O$_2$ ratio around 4:1 at all temperatures up to 37$0^{\circ}C$ and surface analysis using XPS X-ray Photoelectron Spectroscopy) confirms the result. Peak rate at the optimum gas composition increases with increasing temperature. Highest rate obtained in this study leaches 1050 monolayers/min. at 37$0^{\circ}C$ under r. f. power of 150 W, which is equivalent to about 0.5${\mu}{\textrm}{m}$/min. The rate also increases with increasing r. f. power, thus, higher power and higher substrate temperature will undoubtedly raise the etching reaction rate much further. This reaction seems to be an activated process, whose activation energy will be derived in the following experiments.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.551-556
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• 1998

The correlation of isotope composition of Zr with the turnup and some heavy isotopes in PWR uranium dioxide fuel has been investigated. The total and partial ($^{235}$ U) burnup were determined by $^{148Nd}$ and by U and Pu mass spectrometric method, respectively. After separating Zr from the fuel samples, its isotope composition was measured by mass spectrometry. In addition, the quantities of the U and Pu in the spent fuel were determined by isotope di lution mass spectrometric method using $^{233}$ U and $^{242}$ Pu as spikes. The content of some heavy isotopes, $^{235}$ U, $^{239}$ Pu and $^{241}$ Pu, and the Pu Contribution to total turnup were expressed by the correlation with Zr isotope ratios, $^{91}$ Zr/$^{96}$ Zr and $^{93}$ Zr/$^{96}$ Zr The correlations by isotope compositions measured were compared wi th those calculated from ORIGEN2 code.

• Journal of the Korean Ceramic Society
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• v.42 no.5 s.276
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• pp.346-351
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• 2005

The linear thermal expansions of $UO_2$ and $(U_{1-y}Ce_y)O_2$ pellet were measured from room temperature to $1400^{\circ}C$ as a function of Ce contents (0, 7.63, 14.84, and $21.68 mol\%$) by using the TMA(Thermo-Mechanical Analysis) method. From the measured data, the linear thermal expansion rate, the coefficient of linear thermal expansion and density variation with temperature were calculated, and the best-fitted temperature-dependent equations were recommended. It was shown that the rate and coefficient of $(U_{1-y}Ce_y)O_2$ thermal expansion increased and the density decreased with increasing Ce contents.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.4
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• pp.324-329
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• 2010

Fuel rods using in nuclear power plants consist of uranium dioxide pellets enclosed in zirconium alloy(zircaloy) tubes. It is vitally important for the pellet surface to remain free from pits, cracks and chipping defects after it is loaded into the tubes to prevent local hot spots during reactor operation. This paper investigates the feasibility study for detecting surface flaws of pellets contained within nuclear fuel rod through X-ray tomography simulation. Reconstructed images used by parallel and fan-beam filtered back projection method were presented and confirmed the accessibility between simulation data and MPS(missing pellet surface) image data.

• Nuclear Engineering and Technology
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• v.43 no.3
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• pp.309-316
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• 2011

Molybdenum is the most abundant fission product since its fission yield is equivalent to that of xenon, and it has a very special role in the chemistry of nuclear fuel because it influences the oxygen potential of $UO_2$ fuel. In this study, the distribution of molybdenum in spent $UO_2$ fuel specimens with 33.3, 41.0 and 57.6 GWd/tU burnup was measured by a LA-ICP-MS system and the reproducibility of the measured data was obtained. The Mo distribution was almost constant along the radius of a fuel except an increase at the periphery of the fuel. It showed a drop in reproducibility with relatively high deviation of measured values for the highest burnup fuel. To explain this, the state of molybdenum in a $UO_2$ matrix and its effect on the oxidation behavior of $UO_2$ were investigated. The low reproducibility was explained by the segregation of molybdenum, and the inhibition of oxidation by the molybdenum was also observed.

• 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.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4052-4059
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• 2021

The microstructure of the fuel pellet plays an essential role in fission gas buildup and release and is critical for the safe and continued operation of nuclear power stations. Structural analysis of uranium dioxide (UO₂)-molybdenum (Mo) composite fuel pellets prepared at a range of sintering temperatures from 1300 to 1800 ℃ was performed. Mo micro and nanoparticles were used in making the composite pellets. A systematic investigation into the influence of processing parameters during Spark Plasma Sintering (SPS) of the pellets on the microstructure, texture, grain size, and grain boundary characters of UO₂-Mo is presented. UO₂-Mo composite show significant differences in the fraction of general boundaries and also special/coincident site lattice (CSL) boundaries. EBSD orientation maps demonstrated that <111> texturing was observed in the pellets fabricated at 1500 ℃. The experimental investigations suggest that UO₂-Mo composite pellets have favorable microstructural features compared to the UO₂ pellet.