• Korean Journal of Materials Research
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• v.10 no.2
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• pp.166-170
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• 2000

Nuclear fuel $UO_{2+x}$ power was produced from concentrated $UF_6$ by the DCP(Dry Conversion Process). The characterstics of $UO_{2+x}$ powder, prepared with respect to steam flowing conditions and temperature variations in a rotary kiln reactor, have been investigated with a uranium analyzer, water vapor measurement, and SEM. Fluorine content of the powder could be reduced to 8ppm. The moisture content was found to be optimized.

• Journal of the Korean Chemical Society
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• v.29 no.6
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• pp.615-622
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• 1985

The chemical species formed by uranium and vanadium and their equilibria have been investigated in the various concentrations of oxalic and acetic acids by the ion exchange chromatography and UV-Vis spectrophotometry. Uranyl and vanadyl ions seem to be form the complex as $UO_2(C_2O_4)_2=$, $UO_2(C_2O_4)_3^{4-}$ and $VO_2(C_2O_4)-2^{3-}$ respectively in the concentration range of 0.005∼0.05M oxalic acid concentration. In the case of acetic acid the equilibria of $UO_2^{2+}+3Ac^-=UO_2(Ac)_3^-$ and $VO_2^++2Ac^-=VO_2(Ac)_2^-$ were existed individually according to the increase of acetic acid concentration.

• Journal of Powder Materials
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• v.16 no.2
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• pp.115-121
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• 2009

The effects of thermal treatment conditions on ADU (ammonium diuranate) prepared by SOL-GEL method, so-called GSP (Gel supported precipitation) process, were investigated for $UO_2$ kernel preparation. In this study, ADU compound particles were calcined to $UO_3$ particles in air and Ar atmospheres, and these $UO_3$ particles were reduced and sintered in 4%-$H_2$/Ar. During the thermal calcining treatment in air, ADU compound was slightly decomposed, and then converted to $UO_3$ phases at $500^{\circ}C$. At $600^{\circ}C$, the $U_3O_8$ phase appeared together with $UO_3$. After sintering of theses particles, the uranium oxide phases were reduced to a stoichiometric $UO_2$. As a result of the calcining treatment in Ar, more reduced-form of uranium oxide was observed than that treated in air atmosphere by XRD analysis. The final phases of these particles were estimated as a mixture of $U_3O_7$ and $U_4O_9$.

• Korean Journal of Materials Research
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• v.30 no.10
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• pp.533-541
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• 2020

UO₂ kernels, a key component of fuel elements for high temperature gas cooled reactors, have usually been prepared by sol-gel methods. Sol-gel processes have a number of advantages, such as simple processes and facilities, and higher sphericity and density. In this study, to produce 900 ㎛-sized UO₂ particles using an external gelation process, contact length extension of the NH₃ gas of the broth droplets pass and the improvement of the gelation device capable of spraying 14 M-NH₄OH solution are used to form 3,000 ㎛-sized liquid droplets. To produce high-sphericity and high-density UO₂ particles, HMTA, which promotes the gelation reaction in the uranium broth solution, is added to diffuse ammonium ions from the outside of the gelation solution during the aging process and generate ammonium ions from the inside of the ADU gel particles. Sufficient gelation inside of ADU gel particles is achieved, and the density of the UO₂ spheres that undergo the subsequent treatment is 10.78 g/㎤; the sphericity is analyzed and found to be 0.948, indicating good experimental results.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2206-2214
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• 2023

The existing wet reconversion processes for the recovery of scraps generated in manufacturing of nuclear fuel are complex and require several unit operation steps. In this study, it is attempted to simplify the recovery process of high-quality fuel-grade UO₂ powder. A novel wet reconversion process for uranyl nitrate hexahydrate solution is suggested by using a newly developed pulsed fluidized bed reactor, and the resultant chemical characteristics are evaluated for the intermediate ammonium uranate hydrate product and subsequently converted UO₂ powder, as well as the compliance with nuclear fuel specifications and advantages over existing wet processes. The UO₂ powder obtained by the suggested process improved fuel pellet properties compared to those derived from the existing wet conversion processes. Powder performance tests revealed that the produced UO₂ powder satisfies all specifications required for fuel pellets, including the sintered density, increase in re-sintered density, and grain size. Therefore, the processes described herein can aid realizing a simplified manufacturing process for nuclear-grade UO₂ powders that can be used for nuclear power generation.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1099-1109
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• 2020

The performance of gadolinium burnable absorber (GdBA) for reactivity control in UO₂ and (U, Th)O₂ fuels and its impact on spent fuel characteristics was performed. Five fuel assemblies: one without GdBA fuel rod and four each containing 16, 24, 34 and 44 GdBA fuel rods in both fuels were investigated. Reactivity swing in all the FAs with GdBA rods in UO₂ fuel was higher than their counterparts with similar GdBA fuel rods in (U, Th)O₂ fuel. The excess reactivity in all FAs with (U, Th)O₂ fuel was higher than UO₂ fuel. At the end of single discharge burn-up (~ 49.64 GWd/tHM), the excess reactivity of (U, Th) O₂ fuel remained positive (16,000 pcm) while UO₂ fuel shows a negative value (-6,000 pcm), which suggest a longer discharge burn-up in (U, Th)O₂ fuel. The concentration of plutonium isotopes and minor actinides were significantly higher in UO₂ fuel than in (U, Th)O₂ fuel except for ²³⁶Np. However, the concentration of non-actinides (gadolinium and iodine isotopes) except for ¹³⁵Xe were respectively smaller in (U, Th)O₂ fuel than in UO₂ fuel but may be two times higher in (U, Th)O₂ fuel due to its potential longer discharge burn-up.

• Journal of Nuclear Fuel Cycle and Waste Technology
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• v.1 no.1
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• pp.65-73
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• 2013

Temperature-dependent hydrolysis behaviors of aqueous U(VI) species were investigated with time-resolved laser fluorescence spectroscopy (TRLFS) in the temperature range from 15 to $75^{\circ}C$. The formation of four different U(VI) hydrolysis species was measured at pHs from 1 to 7. The predominant presence of $UO{_2}^{2+}$, $(UO_2)_2(OH){_2}^{2+}$, $(UO_2)_3(OH){_5}^+$, and $(UO_2)_3(OH){_7}^-$ species were identified based on the spectroscopic properties such as fluorescence wavelengths and fluorescence lifetimes. With an increasing temperature, a remarkable decrement in the fluorescence lifetime for all U(VI) hydrolysis species was observed, representing the dynamic quenching behavior. Furthermore, the increase in the fluorescence intensity of the further hydrolyzed U(VI) species was clearly observed at an elevated temperature, showing stronger hydrolysis reactions with increasing temperatures. The formation constants of the U(VI) hydrolysis species were calculated to be $log\;K{^0}_{2,2}=-4.0{\pm}0.6$ for $(UO_2)_2(OH){_2}^{2+}$, $log\;K{^0}_{3,5}=-15.0{\pm}0.3$ for $(UO_2)_3(OH){_5}^+$, and $log\;K{^0}_{3,7}=-27.7{\pm}0.7$ for $(UO_2)_3(OH){_7}^-$ at $25^{\circ}C$ and I = 0 M. The specific ion interaction theory (SIT) was applied for the extrapolation of the formation constants to infinitely diluted solution. The results of temperature-dependent hydrolysis behavior in terms of the U(VI) fluorescence were compared and validated with those obtained using computational methods (DQUANT and constant enthalpy equation). Both results matched well with each other. The reaction enthalpies and entropies that are vital for the computational methods were determined by a combination of the van't Hoff equation and the Gibbs free energy equation. The temperature-dependent hydrolysis reaction of the U(VI) species indicates the transition of a major U(VI) species by means of geothermal gradient and decay heat from the radioactive isotopes, representing the necessity of deeper consideration in the safety assessment of geologic repository.

• Nuclear Engineering and Technology
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• v.31 no.2
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• pp.182-191
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• 1999

The sorption of UO$_{2}$$^{2+}$ onto goethite and kaolinite under various experimental conditions was successfully interpreted using surface complexation modeling (SCM). The SCM approach used in this work is the triple-layer model (TLM) in which weakly bonded ions are modeled as outer-sphere (ion-pair) complexes and strongly bonded ions as inner-sphere (surface coordination) complexes. The change of ionic strength did not affect the U(VI) sorption onto goethite, thus the formation of inner-sphere surface complexes, (FeO)$_2$UO$_2$ and (FeO)$_2$(UO$_2$)$_3$OH$_{5}$ was assumed to simulate the effects of ionic strength and goethite concentration. On the other hand, the U(VI) sorption onto kaolinite showed ionic strength dependence, thus the formation of AlO-UO$_{2}$$^{2+}$(outer-sphere complex) and SiO(UO$_2$)$_3$OH$_{5}$ (inner-sphere complex) was assumed to simulate the experimental data. In the presence of carbonates, the sorption of U(VI) onto kaolinite decreased in the weakly alkaline pH range. This was well simulated assuming the formation of a outer-sphere surface complex, A1OH$^{2+}$- (UO$_2$)$_2$CO$_3$OH$_3$. Since SCM approach uses thermodynamic data such as surface complexation constants, it is more predictive than empirical modeling approach in which conditional values such as partition coefficient are used. used.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.343-353
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• 2017

Processing and equipment were tailored for engineering scale fabrication of $UO_2$ porous pellets, a feed material for the electrolytic reduction process in the PRIDE (PyRoprocessing Integrated DEmonstration) facility at KAERI (Korea Atomic Energy Research Institute). The starting materials, $UO_2$ powder and pre-milled surrogate oxide powders, were proportioned to simulate the chemical composition of spent fuel (so-called Simfuel). The Simfuel powders were homogenized by mixing, compacted into a pellet shape, and finally heat treated using a tumbling mixer, rotary press, and sintering furnace. After sintering at $1450^{\circ}C$ for 24 h in $4%\;H_2-Ar$, the average bulk density of the $UO_2$ Simfuel pellets was $6.89g{\cdot}cm^{-3}$, which meets the standard of the following electrolytic reduction process. In addition, the results of a microstructural analysis demonstrated that the sintered Simfuel $UO_2$ porous pellets accurately simulate the properties of spent fuel in terms of the formation of second phases. These results provide essential information for the massive fabrication of $UO_2$ porous pellets for engineering scale pyroprocessing research.

• Nuclear Engineering and Technology
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• v.14 no.3
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• pp.138-147
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• 1982

A theoretical basis for analysis of plutonium-hearing fuel in a thermal nuclear power reactor has been established. The analysis of UO$_2$-PuO$_2$ fueled, light water moderated uniform lattice experiments has been performed. A unit cell program, KARATE, which is based on the theoretical models of GAM and THERMOS with some modifications, has been developed to generate a few-group cross-sections. These cross-sections are subsequently used in the diffusion theory code, KIDD, to compare the calculated values of the effective multiplication factor with the measured. The average value of the effective multiplication factor for 41 selected critical experiments is estimated to be 0.9997 with standard deviation of 0.43%. This illustrates the fact that KARATE/KIDD system can be effectively used for the analysis of uniform lattices of UO$_2$-PuO$_2$fuels.