• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.298-298
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• 2002

• Resources Recycling
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• v.22 no.1
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• pp.3-10
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• 2013

This general paper covers three parts of the uranium research and development. Part one covers scope of research and development of uranium ore and future prospect, supply and demand of uranium in the world market, deposit, grade and properties of Korean uranium ore and the second part covers status of previous study and supply target for yellow cake, technology of leaching, separation and preparation, procedure of the recovery of U / V from Korean black shale ore. Final part concludes the summary of the present discussion.

• 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.

• 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 $80^{\circ}C$.

• 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.

• Proceedings of the Korea Information Processing Society Conference
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• 2012.04a
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• pp.828-831
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• 2012

한국원자력연구원내에는 연구실험 목적용으로 사용된 후 용도 폐기된 각종 우라늄 화합물이 수요가 증대 되어온 바, 이러한 불용 우라늄 화합물의 저장관리에 대한 체계적인 시스템의 구축이 필요하게 되었다. 이에 한국원자력연구원에서는 불용 핵물질 원격 물리적 방호시스템, UReMon(Uranium Residues Monitoring System)을 개발하였는데 이는 방사성 물질인 불용 우라늄 화합물의 물리적 방호와 관리 및 도난 방지의 목적을 지닌다. UReMon은 기존 모니터링 서비스에서 자주 사용되던 RFID나 바코드가 가지는 기술적 문제로 인한 위치확인, 도난, 훼손 등의 실태 파악에 소요되는 많은 시간과 경비를 줄이기 위하여 USN 센서와 Zigbee를 이용하여 한국원자력연구원에 기 구축되어 있는 USN기반 화재 예방시스템(KAERI-uFIPI)과의 연계를 통해 불용 핵물질의 모니터링, 위치 추적 및 재고관리의 효율성을 높인다. UReMon은 연구원 내 물리적 방호 시스템, 핵 물질 및 RI 관리, 출입통제 시스템 등에도 효율적으로 적용 가능하며, 향후 이에 대한 적용성 평가를 수행할 예정이다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.491-496
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• 1996

감손우라늄 폐기물은 칩의 형태로 발생하며 이들은 열적으로 불안정하여 운반 및 저장에 주의를 요하게 된다. 본 연구에서는 감손우라늄 폐기물의 안정한 처리를 위해 공기조절식 산화 장치를 개발하고 장치의 운전에 필요한 기초 자료를 얻기 위해 산화실 험을 수행하였다. 저장 및 처분시 가장 안정한 화합물인 U$_3$O$_{8}$ 으로 변환되는 산화온도는 약 3$25^{\circ}C$ 이상이며 산화속도는 다음과 같다.

• Journal of the Korean Ceramic Society
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• v.39 no.2
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• pp.204-209
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• 2002

In this study, $UO_4{\cdot}2NH_4F$, the precipitates which has low solubility, was obtained by chemical precipitation method to recover and reuse the trace uranium from the liquid waste producing in AUC process and for this compound it was characterized by means of chemical analysis, TG-DTA, XRD and FT-IR analyses. This compound was analyzed as $UO_4{\cdot}2NH_4F$ and shape of this precipitate was hexagonal type, having the size of 2∼3 ${\mu}m$. Also, the intermediates were obtained as $UO_4F,\;UO_4,\;UO_3,\;and\;U_3O_8$ by the thermal decomposition over the temperature of 220, 310, 515 and 640$^{\circ}C$, respectively. It is concluded that under the condition of a constant heating rate of 5$^{\circ}C$/min in air atmosphere range of between room temperature and 800$^{\circ}C$, thermal decomposition reaction mechanism of $UO_4{\cdot}2NH_4F$ is as follow; $UO_4{\cdot}2NH_4F{\rightarrow}UO_4F{\rightarrow}UO_4{\rightarrow}UO_3{\rightarrow}U_3O_8$.

• Economic and Environmental Geology
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• v.41 no.6
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• pp.719-729
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• 2008

In this study, the speciation, adsorption, and transport of uranium in groundwater environments were simulated using geochemical models. The retarded transport of uranium by adsortption was effectively simulated using 3-D groundwater flow and reactive transport models. The results showed that most uranium was adsorbed(up to 99.5%) in a neutral pH(5.5$pCO_2(10^{-3.6}atm)$ condition. Under the higher $pCO_2(10^{-2.5}atm)$ condition, however, the pH range where most uranium was absorbed was narrow from 6 to 7. Under very low $pCO_2(10^{-4.5}atm)$ condition, uranium was mostly absorbed in the relatively wide pH range between 5.5 and 8.5. In the model including anion complexes, the uranium adsorption decreased by fluoride complex below the pH of 6. The results of this study showed that uranium transport is strongly affected by hydrochemical conditions such as pH, $pCO_2$, and the kinds and concentrations of anions($Cl^-$, ${SO_4}^{2-}$, $F^-$). Therefore, geochemical models should be used as an important tool to predict the environmental impacts of uranium and other hazardous compounds in many site investigations.

• 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.