• Journal of the Korean Ceramic Society
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• v.44 no.2 s.297
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• pp.124-131
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• 2007

In this study, first the ADU gel particle, an intermediate for final $UO_2$ kernel of a HTGR nuclear fuel, was prepared from sol-gel method using the broth solution which was made by mixing of the uranyl nitrate, poly vinyl alcohol and tetra-hydrofurfuryl alcohol. The prepared dried-ADU gel particles were converted to the $UO_2\;via\;UO_3$ from thermal treatment with the 4% $H_2$ atmosphere. The sizes of the spherical liquid droplets appeared $1900{\sim}2100{\mu}m$, and the harmony between the flow rate of the broth solution and the frequency and the amplitude of a vibrating system are important factors for the spherical ADU gel particles via the mono size spherical droplets. From the XRD and FT-IR analyses, the prepared ADU gel particles were judged to be a $UO_3{\cdot}xNH_3{\cdot}yH_2O$ form, and the most important factor during the thermal treatment of the dried-ADU gel particle must be avoided a rapidly heating rate in the range of $180{\sim}400^{\circ}C$, and the heating rate should be kept below $5^{\circ}C/min$.

• Nuclear Engineering and Technology
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• v.27 no.1
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• pp.67-73
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• 1995

To investigate the storage behavior of the defective LWR spent fuel, air-oxidation tests on non-irradiated and irradiated U $O_2$ were performed. The weight gains of non-irradiated U $O_2$ specimens are characterized by the S-shape curves at 250-40$0^{\circ}C$ temperature range. One hundred percent conversion of U $O_2$ to U$_3$ $O_{8}$ corresponds with about 4％ weight increase. The activation energies are 110 kJ/mol above 35$0^{\circ}C$ and 153 kJ/mol below 35$0^{\circ}C$. The irradiated U $O_2$ specimens with about 35 GWD/MTU burnup were oxidized at 300-40$0^{\circ}C$ in air. They show a rapid increase of weight gain at the initial stage and a slow increase at the later stage when compared with non-irradiated U $O_2$. The activation energy under these conditions is 95 kJ/mol. Burnup and aging effects of irradiated U $O_2$ were also investigated at 35$0^{\circ}C$ in air environment, but the specimens appears insensitive to these variables.s.

• Bulletin of the Korean Chemical Society
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• v.6 no.4
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• pp.225-228
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• 1985

The absorptions of Fe(Ⅲ), Tb(Ⅲ), Tl(Ⅰ), Ce(Ⅲ), Th(Ⅳ), and $UO_2^{2+}$ ions into the Dowex 50W-X2, 100-200 mesh resin were investigated by spectrophotometry to understand the abnormal strong absorption behavior of cations to cation exchange resins in concentrated HClO4. The distribution coefficients increase in the order : Tl(Ⅰ) < Fe(Ⅲ) < Tb(Ⅲ)∼Ce(Ⅲ) < $UO_2^{2+}$< Th(Ⅳ) and the order is interpreted in terms of the ratio of charge-to-ionic radius. The mole ratios of increment of $ClO_4^-$ ion absorption to metal ion absorption showed the same order as the distribution coefficients, which indicates that the electrostatic association between $ClO_4^-$ ion and metal ion plays a major role in the strong absorption.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1300-1309
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• 2023

Simulated debris was synthesized using UO₂, Zr, and stainless steel and a heat treatment method under inert or oxidizing conditions. The primary U solid phase of the debris synthesized at 1473 K under inert conditions was UO₂, whereas a (U, Zr)O₂ solid solution formed at 1873 K. Under oxidizing conditions, a mixture of U₃O₈ and (Fe, Cr)UO₄ phases formed at 1473 K, whereas a (U, Zr)O_2+x solid solution formed at 1873 K. The leaching behavior of the fission products from the simulated debris was evaluated using two methods: the irradiation method, for which fission products were produced via neutron irradiation, and the doping method, for which trace amounts of non-radioactive elements were doped into the debris. The dissolution behavior of U depended on the properties of the debris and aqueous solution for immersion. Cs, Sr, and Ba leached out regardless of the primary solid phases. The leaching of high-valence Eu and Ru ions was suppressed, possibly owing to their solid-solution reaction with or incorporation into the uranium compounds of the simulated debris.

• Analytical Science and Technology
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• v.16 no.2
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• pp.117-124
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• 2003

Oxygen to metal ratio has been measured by wet and dry chemical analysis to study the properties of sintered $UO_2$ pellets and $U_3O_8$ in the lithium reduction process of spent pressurized water reactor fuels. Uranium dioxide pellets simulated for the spent PWR fuels with burnup values of 20,000~60,000 MWd/MtU were prepared by mixing $UO_2$ powder and oxides of fission product elements, pelleting the powder mixture and sintering it at $1,700^{\circ}C$ under a hydrogen atmosphere. For wet chemical analysis, the simulated spent fuels were dissolved with mixed acid (10 M HCl : 8 M $HNO_3$, 2.5 : 1, v/v) using acid digestion bomb technique. The total amount of uranium and fission products added in the simulated spent fuels were measured using inductively coupled plasma atomic emission spectrometry. Weight change of the simulated fuel during its oxydation was measured by thermogravimetry and then the O/M ratio result was compared to that obtained by wet chemical analysis. Influence of $Mo_{0.4}-Ru_{0.4}-Rh_{0.1}-Pd_{0.1}$, quaternary alloy, on the determination of O/M ratio was investigated.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.217-222
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• 1997

핵연료 물질인 금속 우라늄과 이산화 우라늄의 플라즈마 기체에 의한 에칭 연구가 수행되었다. 연구에 사용된 플라즈마 기제는 CF$_4$와O$_2$의 혼합기체이며 CF$_4$/O$_2$의 혼합비. 시편 표면의 온도, R.F power, 그리고 압력에 따른 에칭율을 측정하였다. L-metal의 경우는 R.F power를 50W로 고정하고 아주 낮은 $O_2$의 성분비와 반응시간에 따른 에칭정도를 질량결손으로 계산하였다. $UO_2$의 에칭에 있어서는 CF$_4$/O$_2$의 비가 4:1에서 가장 높은 에칭율을 보였으며 그 에칭율은 최대 1000 monolayers/min 이었으며 U-metal의 경우 그 에칭율은 $UO_2$와 비교하여 10배 가량 낮은 것으로 나타났다.

• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1711-1716
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• 2017

Uranium tetrafluoride ($UF_4$) is the most used nuclear material for producing metallic uranium by reduction with Ca or Mg. Metallic uranium is a raw material for the manufacture of uranium silicide, $U_3Si_2$, which is the most suitable uranium compound for use as nuclear fuel for research reactors. By contrast, ammonium uranyl carbonate is a traditional uranium compound used for manufacturing uranium dioxide $UO_2$ fuel for nuclear power reactors or $U_3O_8-Al$ dispersion fuel for nuclear research reactors. This work describes a procedure for recovering uranium and ammonium fluoride ($NH_4F$) from a liquid residue generated during the production routine of ammonium uranyl carbonate, ending with $UF_4$ as a final product. The residue, consisting of a solution containing high concentrations of ammonium ($NH_4^+$), fluoride ($F^-$), and carbonate ($CO_3^{2-}$), has significant concentrations of uranium as $UO_2^{2+}$. From this residue, the proposed procedure consists of precipitating ammonium peroxide fluorouranate (APOFU) and $NH_4F$, while recovering the major part of uranium. Further, the remaining solution is concentrated by heating, and ammonium bifluoride ($NH_4HF_2$) is precipitated. As a final step, $NH_4HF_2$ is added to $UO_2$, inducing fluoridation and decomposition, resulting in $UF_4$ with adequate properties for metallic uranium manufacture.

• Journal of the Korean Electrochemical Society
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• v.3 no.3
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• pp.141-145
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• 2000

Anodically Polymerized conducting Polypyrrole film electrode was employed to Pick up uranyl ion with the type of Gr/ppy, xylenol orange modified electrode. To have Porous and oriented ppy film, NBR was applied as precoating agent. The rate constant of polymerization was $3.22\times10^{-3}s^{-1}$ which was 1.6 times smaller value than bare graphite surface. The deposited amount of uranyl iou on $1.70Ccm^{-2}$ of ppy was $1.55\times10^{-4}g$. The matrix effect in artificial seawater was $6.8\%$. The polymer film electrode has a diffusion controlled process in conduction, but the modified Gr/ppy, $X.O^{4-}UO^+$ type was influenced on the ion doping and electronic conduction of film itself owing to increasing of impedance. The capacitance of electrical double layer was respectively enhanced to 56 and 130 times in Gr/ppy, $X.O.^{4-}$ and Gr/ppy, $X.O^{4-}UO^+$ than Grippy type electrode.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.3
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• pp.189-197
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• 2007

The solubility of U(VI) hydrolysis products was determined by using a laser-induced breakdown detection (LIBD) technique. The experiments were carried out at uranium concentrations in range from $2{\times}10^{-4}\;M\;to\;4{\times}10^{-6}\;M$, pH values between 3.8 and 7.0, the constant ionic strength of 0.1 M $NaClO_4$ and the temperature of $25.0{\pm}0.1^{\circ}C$. The solubility product of U(VI) hydrolysis products was calculated from LIBD results by using the hydrolysis constants selected in NEA-TDB. The solubility product extrapolated to zero ionic strength, ${\log}K^{\circ}_{sp}=-22.85{\pm}0.23$ was calculated by using a specific ion interaction theory (SIT). The spectral features of ionic species in uranium solutions were investigated by using a conventional UV-visible absorption spectrophotometer and a fluorophotometer, respectively, $(UO_2)_2(OH)_2^{2+}\;and\;(UO_2)_3(OH)_5^+$ were dominant species at uranium concentration of $2{\times}10^{-4}\;M$.

• Applied Chemistry for Engineering
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• v.8 no.3
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• pp.430-437
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• 1997

Uranyl-peroxide compound was prepared by the reaction of excess hydrogen peroxide solution and trace uranium in filtrate from nuclear fuel conversion plant. The $CO_3{^{2-}}$ in filtrate was removed first by heating more than $98^{\circ}C$, because uranyl-peroxide compound could not be precipitated by $CO_3{^{2-}}$ remaining in filtrate. The optimum condition for uranyl-peroxide compound was ageing for 1 hr after controling the pH with $NH_3$ gas and adding the excess $H_2O_2$ of 10ml/lit.-filtrate. Uranium concentration in the filtrate was appeared to 3 ppm after the precipitation of uranyl-peroxide compound, and the chemical composition of this compound was analyzed to $UO_4{\cdot}2NH_4F$ with FT-IR, X-ray diffractometry, TG and chemical analysis. Also, this fine particle, about $1{\sim}2{\mu}m$, could be grown up to $4{\mu}m$ at pH 9.5 and $60^{\circ}C$. The separation efficiency of precipitate from mother liquor was increased with increase of pH and reaction temperature. Otherwise, the crystal form of this particle showed octahedral by SEM and XRD, and $U_3O_8$ powder was obtained by thermal decomposition at $650^{\circ}C$ in air atmosphere.