• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.318-318
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.338.1-338
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.289-294
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• 1997

우라늄 염화물의 생성을 위한 U$_3$O$_{8}$ 분말의 염화반응에 대한 열역학적 해석을 수행하여 실험조건을 도출하였고 이를 실험적으로 확인하였다. U$_3$O$_{8}$ 분말의 탄소, 염소와의 염화반응은 비가역 발열반응으로서 아르곤 기체분위기에서는 옥시염화물의 생성이 억제된 우라늄 염화물들을 생성할 수 있었다. 염화반응 적정온도 범위는 863～1065k로서 이 온도범위에서의 반응종결시간은 U$_3$O$_{8}$ 분말 60g을 기준으로 약 4시간 이내였고, 반응온도가 높을수록 반응속도가 빠르게 나타났다. 우라늄 염화물의 비휘발 회수율은 생성된 우라늄 염화물 총량의 30% 이내로 낮게 나타났다.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.223.1-223.1
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• 2003

• Journal of the Korean Crystal Growth and Crystal Technology
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• v.17 no.2
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• pp.75-81
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• 2007

Densification and grain growth have been investigated in 5 wt% $U_3O_8$ seeded $UO_2$ and compared with those of the common $UO_2$ pellet. $UO_2$ compacts and $U_3O_8$ seeded $UO_2$ compacts were sintered at $1300{\sim}1700^{\circ}C$ for 0 h to 4 h. Density and grain size of the sintered pellets were measured by the water immersion method and the image analyzer, respectively. The seeded pellet has a slightly lower density during the intermediate sintering stage. However, the difference of density between two pellets decreases to about 0.5%TD with increasing the sintering temperature. The grain size of the two pellets is similar until $1600^{\circ}C$ but that of the seeded pellet rapidly increases with increasing the sintering temperature.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.2
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• pp.193-207
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• 2022

Thermodynamically, TRUO_x, REO_x, and SrO_x can be chlorinated using ammonium chloride (NH₄Cl) as a chlorinating agent, whereas uranium oxides (U₃O₈ and UO₂) remain in the oxide form. In the preliminary experiments of this study, U₃O₈ and CeO₂ are reacted separately with NH₄Cl at 623 K in a sealed reactor. CeO₂ is highly reactive with NH₄Cl and becomes chlorinated into CeCl₃. The chlorination yield ranges from 96% to 100%. By contrast, U₃O₈ remains as UO₂ even after chlorination. We produced U/REO_x- and U/SrO_x-simulated fuels to understand the chlorination characteristics of the oxide compounds. Each simulated fuel is chlorinated with NH₄Cl, and the products are dissolved in LiCl-KCl salt to separate the oxide compounds from the chloride salt. The oxide compounds precipitate at the bottom. The precipitate and salt phases are sampled and analyzed via X-ray diffraction, scanning electron microscope-energy dispersive spectroscopy, and inductively coupled plasma-optical emission spectroscopy. The analysis results indicate that REO_x and SrO_x can be easily chlorinated from the simulated fuels; however, only a few of U oxide phases is chlorinated, particularly from the U/SrO_x-simulated fuels.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.411-420
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• 2018

This study presents a revised crackling core model for the description of $UO_2$ sphere oxidation in air atmosphere. For close reproduction of the sigmoid behavior exhibited in $UO_2$ to $U_3O_8$ conversion, the fragmentation effect contributing to the increased reactive surface area and the concept of variable grain conversion time were considered in the model development. Under the assumptions of two-step successive reaction of $UO_2{\rightarrow}U_3O_7{\rightarrow}U_3O_8$ and final grain conversion time equivalent to ten times the initial grain conversion time, the revised model showed good agreement with the experimental data measured at 599 - 674 K and a lowest deviation when compared with Nucleation and Growth model and AutoCatalytic Reaction model. The evaluated activation energy at 100% conversion to $U_3O_8$, $57.6kJ{\cdot}mol^{-1}$, was found to be closer to the experimentally extrapolated value than to the value determined in AutoCatalytic Reaction model, $48.6kJ{\cdot}mol^{-1}$.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.06a
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• pp.352-352
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• 2004

산화물 형태의 사용후핵연료를 고온 용융염계에서 금속 형태로 전환하는 전기화학적 금속전환 공정 개발의 일환으로 $U_3O_8$ 분말로 충전된 다공성 마그네시아 용기 및 스테인레스강 고체전극으로 구성된 일체형 음극과 $SnO_2$ 재질의 양극을 사용하여 5kg $U_3O_8$/batch 규모의 mock-up 시험을 수행하였다. 백금 재질의 양극을 사용하였을 때 99% 이상의 금속전환율을 보인 동일한 전하량을 공급하고 실험을 중단한 결과 X-선 회절분석(XRD) 및 열중량 분석(TG)으로부터 스테인레스강 고체전극 부분에서는 거의 금속으로 전환되었으나 다공성 마그네시아 용기 부분에서는 비교적 금속전환율이 낮은 경향을 나타내었다.(중략)

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.215.2-215
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• 1999

• Nuclear Engineering and Technology
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• v.20 no.3
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• pp.165-169
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• 1988

As a part of tandem fuel cycle feasibility study, the residual U and Pu nuclide contents of PWR spent fuels are computed using ORICEN2 code for each Korea Nuclear Unit and batch to investigate the potential of utilizing them as CANDU fuels. The annual and accumulated discharged amounts of U and Pu nuclides are computed for the PWRs from KNU 1 through KNU 10. The results of computation show that the spent fuels having 0.7-0.8 w/o U-235 are dominant and considerable amounts of fissile Pu are produced. The enrichment of U-235 is less than the expected 0.8-0.9 w/o U-235 since the burnups offered by KEPCO are higher than those of other PWRs.