• 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 Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.3
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• pp.177-181
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• 2005

Yellowish uranium compounds were enriched at the interface between a Ca-bentonite block and a waste glass, containing about $20\%$ uranium oxide, in contact with the block due to the dissolution of uranium by a synthetic granitic groundwater in Ar atmosphere. The uranium compound formed for 6 years leach time was identified as a beta-uranophane $[Ca(UO_2)_2(SiO_{3}OH)_{2}5H_{2}O]$ using XRD, IR and mass spectrometer. The solubility of the beta-uranophane was measured to be about $10^{-6}\;mole/L$ in de-mineralized water at $80^{\circ}C$.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.1
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• pp.1-7
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• 2007

Electrorefining experiments were successfully carried out in LiCl-KCl eutectic molten salt with a graphite cathode. It was found that the formation of Uranium-Graphite intercalation compound(U-GIC) helped the self-scraping mechanism of the uranium dendrite and the efficiency of the electrorefiner increased due to an elimination of the stripping step. The contaminations of the uranium deposit by rare earth elements was negligible while about 300 ppm of carbon was observed. The carbon contamination is believed to be eliminated by further purification by yttrium reaction. The morphology characteristics of the recovered U deposit was compared to that of steel cathode. These are only qualitative preliminary experimental results, but we believe that further research on this type of activity change the direction of the electrorefining research on spent nuclear fuel.

• 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 $U_4O_9$.

• Nuclear Engineering and Technology
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• v.15 no.2
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• pp.117-122
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• 1983

The chelating agent with $\beta$-diketo funtional group, 1-phenyl-3-methyl-4-acyl-pyrazolone-5-one, has been used in separating and extracting radionuclides in a waste solution. The derivatives of this pyrazolone compound, prepared by different acyl groups, were synthesized and examined to figure out the extracting ability for Uranium (VI) and Zirconium (IV). The product prepared with succinic anhydride, called succinyl pyrazolone, showed excellent extraction for uranium (VI) in a chloroform solvent system. This result indicates that acyl pyrazolones having carboxylic acid group as a functional group forming $\beta$-diketo functionality are very selective for uranium (VI) and generally other metal ions with high valency.

• Applied Chemistry for Engineering
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• v.8 no.1
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• pp.99-107
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• 1997

This study Provides the treatment methods of liquid wastes from the dissolution/purification process of nuclear fuel raw material and the fabrication process of nuclear fuel powder. One of the treatment methods is to process liquid waste from uranium raw material dissolution/purification process. This waste, of the strong acid, can be reused to dissolve the fine ADU particles in filtrate which is ADU waste of pH 8.0 converted from AUC waste after recovery of uranium. To dissolve the fine ADU particles, ADU filtrate was pretreated to pH 4.0 with the dissolution/purification waste, and then mixed with the lime to pH 9.2 and aged for 30 minutes. From this processing, uranium content of the filtrate was decreased to below 3ppm. The waste from fuel powder fabrication is emulsified solution dispersed with fine oil droplets. This emulsion was destroyed effectively by adding and mixing the nitric acid with rapid heating at the same time. After this processing, $Na_2U_2O_7$ compound is produced by addition of NaOH. Optimum condition of this processing was shown at pH 11.5, and uranium content of the filtrate was analyzed to 5ppm. To remove the trace of uranium in the filtrate, lime should be added. Otherwise, 4N nitric acid was used to destroy the emulsion directly, and then lime was added to this waste. Uranium content of the treated filtrate was below 1 ppm. In addition to these wastes, the trace of uranium in filtrate after recovery of uranium from the AUC waste which is produced during PWR power preparation, is treated with NaOH to takeup fluorine(F) in the waste because fluorine is valuable and toxic material. In the finally treated waste, uranium was not detected.

• 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.

• Analytical Science and Technology
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• v.23 no.1
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• pp.45-53
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• 2010

For the determination of chlorine in uranium compound, analytical methods by using a steam distillation and a pyrohydrolysis have been developed. The steam distillation apparatus was composed of steam generator, distilling flask and condenser etc. The samples were prepared with an aliquot of LiCl standard solution and a simulated spent nuclear fuel. A sample aliquot was mixed with a solution containing 0.2 M ferrous ammonium sulfate-0.5 M sulfamic acid 3 mL, phosphoric acid 6 mL and sulfuric acid 15 mL. The chloride was then distilled by steam at the temperature of $140^{\circ}C$ until a volume of $90{\pm}5\;mL$ is collected. The pyrohydrolysis equipment was composed of air introduction system, water supply, quartz reaction tube, combustion tube furnace, combustion boat and absorption vessel. The chloride was separated from powdered sample which is added with $U_3O_8$ accelerator, by pyrohydrolysis at the temperature of $950^{\circ}C$ for 1 hour in a quartz tube with a stream of air of 1 mL/min supplied from the water reservoir at $80^{\circ}C$. The chlorides collected in each absorption solution by two methods was diluted to 100 mL and measured with ion chromatography to determine the recovery yield. For the ion chromatographic determination of chlorine in molten salt retained in a metal ingot, the chlorine was separated by means of pyrohydrolysis after air and dry oxidation, and grinding for the sample.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.3
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• pp.199-206
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• 2013

Electrowinning process in pyroprocessing recovers U (uranium) and TRU (Trans Uranium) elements simultaneously from spent fuels using a liquid cadmium cathode (LCC). When the solubility limit of U deposits over 2.35wt% in Cd, U dendrites were formed on the LCC surface during the electrodeposition at $500^{\circ}C$. Due to the high surface area of dendritic U, the deposits were not submerged into the liquid cadmium pool but grow out of the LCC crucible. Since the U dendrites act as a solid cathode, it prevents the co-deposition of U and TRUs. In this study, the electrodeposition of U onto a LCC was carried out at 440 and $500^{\circ}C$ to compare the morphology and component of U deposits. The U deposits at $440^{\circ}C$ have a specific shape and were stacked regularly at the center of the LCC pool, while the U dendrites (i.e., ${\alpha}$-phase) at $500^{\circ}C$ were grow out of the LCC crucible. Through the microscopic observation and XRD analysis, the electrodeposits at $440^{\circ}C$, which have a round shape, were identified as an intermetallic compound such as $UCd_{11}$. It can be concluded that the LCC electrowinning operation at $440^{\circ}C$ achieves the co-recovery of U and TRU without the formation of U dendrites.

• Applied Chemistry for Engineering
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• v.7 no.6
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• pp.1164-1173
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• 1996

Treating methods and characteristics of waste from a nuclear fuel powder conversion plant were studied. To recovery or treat a trace uranium in liquid waste, the ammonium uranyl carbonate(AUC) filtrate must be heated for $CO_2$ expelling, essentially. Uranium content of final treated waste solution from fuel powder processes for a heavy water reactor(HWR) could be lowered to 1 ppm by the lime treatment after the ammonium di-uranate(ADU) precipitation by simple heating. Otherwise, in case of the waste from fuel powder processes for a pressurized light water reactor(PWR), it is result in 0.8 ppm as a form of uranium peroxide such as $UO_4{\cdot}2NH_4F$ compounds. Optimum condition was found at $101^{\circ}C$ by the simple heating method in case of HWR powder process waste. And in case of PWR powder process waste, optimum condition could be obtained by precipitating with adding hydrogen peroxide and adjusting at pH 9.5 with ammonia gas at $60^{\circ}C$ after heating the waste In order to expelling $CO_2$. As the characteristics of recovered uranium compounds, median particle size of ADU was increased with pH increasing in case of HWP waste. Also, in case of uranium proxide compound recovered from PWR waste, the property of $U_3O_8$ power obtained after thermal treatment in air atmosphere was similar to that of the powder prepared from AUC conversion plant.