• 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; $UO_4{\cdot}2NH_4F{\rightarrow}UO_4F{\rightarrow}UO_4{\rightarrow}UO_3{\rightarrow}U_3O_8$.

• Nuclear Engineering and Technology
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• v.37 no.3
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• pp.293-298
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• 2005

We have performed the density functional theory calculations of $UO_2$ using the spin-polarized generalized gradient approximation (SP-GGA) and the SP-GGA+U approach. The SP-GGA+U approach correctly predicts the insulating electronic structure with antiferromagnetic ordering, but the SP-GGA calculations predict metallic behavior. The cohesive properties obtained from the SP-GGA+U calculations are in good agreement with the available experimental results and previous calculations. The spin-polarized local density of states shows that the antiferromagnetic ordering of $UO_2$ is governed by 5f orbitals of uranium ion. Our calculations demonstrate that the strong correlation of U 5f electrons should be taken into account for a reliable description of $UO_2$ physics.

• Nuclear Engineering and Technology
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• v.35 no.6
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• pp.507-514
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• 2003

Recently dry decontamination/surface-cleaning technology using plasma etching has been focused in the nuclear industry. In this study, the applicability of this new dry processing technique are experimentally investigated by examining the etching reaction of $UO_2$, Co, and Mo in r.f. plasma with the etchant gas of $CF_4/O_2$ mixture. $UO_2$ is chosen as a representing material for uranium and TRU (TRans-Uranic) compounds while metallic Co and Mo are selected because they are the principal contaminants in the used metallic nuclear components such as valves and pipes made of stainless steel or inconel. Results show that in all cases maximum etching rate is achieved when the mole fraction of $UO_2\;in\;CF_4/O_2$ mixture gas is $20\%$, regardless of temperature and r.f. power. In case of $UO_2$, the highest etching reaction rate is greater than 1000 monolayers/min. at $370^{\circ}C$ under 150 W r.f. power which is equivalent to $0.4{\mu}m/min$. As for Co, etching reaction begins to take place significantly when the temperature exceeds $350^{\circ}C$. Maximum etching rate achieved at $380^{\circ}C\;is\;0.06{\mu}m/min$. Mo etching reaction takes place vigorously even at relatively low temperature and the reaction rate increases drastically with increasing temperature. Highest etching rate at $380^{\circ}C\;is\;1.9{\mu}m/min$. According to OES (Optical Emission Spectroscopy) and AES (Auger Electron Spectroscopy) analysis, primary reaction seems to be a fluorination reaction, but carbonyl compound formation reaction may assist the dominant reaction, especially in case of Co and Mo. Through this basic study, the feasibility and the applicability of plasma decontamination technique are demonstrated.

• Korean Journal of Computational Design and Engineering
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• v.5 no.1
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• pp.23-32
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• 2000

In this paper, we present a Web-enabled product data sharing system for the support of RPD (Rapid Product Development) process by incorporating STEP (STandard for the Exchange of Product model data) with Web technology such as VRML (Virtual Reality Markup Language), SGML (Structured Generalized Markup Language) and Java. Extreme competition makes product life cycle short by incessantly deprecating current products with a brand-new one, and thus urges enterprises to devise a new product faster than ever. In this environment, an RPD process with effective product data sharing system is essential to outstrip competitors by speeding up the development process. However, the diversity of product data schema and heterogeneous systems make it difficult to exchange the product data. We chose STEP as a neutral product data schema and Web as an independent exchange environment to overcome these problems. While implementing our system, we focused on the support of STEP AP 203 UoF (Units of Functionality) views to efficiently employ STEP data models that are maximally normalized, and therefore very cumbersome to handle. Our functionality-oriented UoF view approach can increase users'appreciation since it facilitates the modular usage of STEP data models. This can also enhance the accuracy of product data. We demonstrate that our view approach is applicable to the configuration control of mechanical assemblies.

• 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.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1236-1249
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• 2021

Although there are still controversial opinions and uncertainty on application of SiC_f/SiC composite cladding as next-generation cladding material for its great oxidation resistance in high temperature steam environment and other outstanding advantages, it cannot deny that SiC_f/SiC cladding is a potential accident tolerant fuel (ATF) cladding with high research priority and still in the engineering design stage for now. However, considering its disadvantages, such as low irradiated thermal conductivity, ductility that barely not exist, further evaluations of its in-pile behaviors are still necessary. Based on the self-developed code we recently updated, relevant thermohydraulic and mechanical models in FROBA-ATF were applied to simulate the cladding behaviors under normal and accident conditions in this paper. Even through steady-state performance analysis revealed that this kind of cladding material could greatly reduce the oxidation thickness, the thermal performance of UO₂-SiC was poor due to its low inpile thermal conductivity and creep rate. Besides, the risk of failure exists when reactor power decreased. With geometry optimization and dopant addition in pellets, the steady-state performance of UO₂-SiC was enhanced and the failure risk was reduced. The thermal and mechanical performance of the improved UO₂-SiC was further evaluated under Loss of coolant accident (LOCA) and Reactivity Initiated Accident (RIA) conditions. Transient results showed that the optimized ATF had better thermal performance, lower cladding hoop stress, and could provide more coping time under accident conditions.

• 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.

• Analytical Science and Technology
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• v.14 no.3
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• pp.203-211
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• 2001

In this paper it was described on the intereference effect of uranium to analyze Cs in $UO_2$ by Electron Probe Micro Analysis(EPMA) and the beam stability of Cs $L_{\alpha}$ X-ray intensity for some Cs compounds. According to the experimental results, the CsI showed the highest $L_{\alpha}$ X-ray intensity among the tested Cs compounds at the experimental condition; 15~30 kV of accelerating voltage and PET, LiF crystal. When 100 nA of beam current was applied to Cs compounds, Cs $L_{\alpha}$ X-ray intensity was continuously decreased with increasing time. The decreasing rate of Cs $L_{\alpha}$ X-ray intensity was directly proportional to the applied beam current and accelerating voltage but inversely proportional to the applied beam size. It was found that uranium interference can be prevented by using Cs $L_{\alpha}$ X-ray wavelength of Lif crytal for Cs analysis in $UO_2$ by EPMA.

• 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.