• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.310.2-310.2
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.340.2-340.2
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• 2003

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.244-250
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• 1996

고리 2호기에서 2주기 동안 연소된 1개 KOFA 연료봉에 대한 조사후 검사결과, 핵분열기체 방출량 및 소결체 밀도가 연료봉 설계코드의 예측범위내에 있음을 확인하였으며, 소결체의 미세구조 및 연료봉내의 축방향 분포 검사를 통해 $UO_2$ 소결체가 아무 이상이 없이 안정적으로 연소되었음을 확인하였다. 단지 1개 연료봉에 대한 조사후 검사만으로는 KOFA 핵연료 $UO_2$ 소결체의 노내 거동을 검증하였다고는 할수 없기 때문에 연소된 핵연료에 대한 지속적인 조사후 검사가 필요한 것으로 사료된다. 특히 한국형원자로의 핵연료인 영광 3호기 핵연료에 대해 조사후 검사를 수행하고, 또한 일부 시험연료봉을 고연소도까지 연소시킨후 조사후 검사를 수행하면, 핵연료의 성능 검증뿐만 아니라 국내 고유의 핵연료 성능자료를 생산하게됨으로써, 앞으로 국내 고유의 고연소도핵연료개발 및 연료봉성능분석코드 개발에 활용할 수 있다.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.9 no.4
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• pp.207-217
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• 2011

In this study the complex formation reactions between uranium(VI) and 2,6-dihydroxybenzoate (DHB) as a model ligand of humic acid were investigated by using UV-Vis spectrophotometry and time-resolved laser-induced fluorescence spectroscopy (TRLFS). The analysis of the spectrophotometric data, i.e., absorbance changes at the characteristic charge-transfer bands of the U(VI)-DHB complex, indicates that both 1:1 and 1:2 (U(VI):DHB) complexes occur as a result of dual equilibria and their distribution varies in a pH-dependent manner. The stepwise stability constants determined (log $K_1$ and log $K_2$) are $12.4{\pm}0.1$ and $11.4{\pm}0.1$. Further, the TRLFS study shows that DHB plays a role as a fluorescence quencher of U(VI) species. The presence of both a dynamic and static quenching process was identified for all U(VI) species examined, i.e., ${UO_2}^{2+}$, $(UO_2)_2{(OH)_2}^{2+}$, and $(UO_2)_3{(OH)_5}^+$. The fluorescence intensity and lifetimes of each species were measured from the time-resolved spectra at various ligand concentrations, and then analyzed based on Stern-Volmer equations. The static quenching constants (log $K_s$) obtained are $4.2{\pm}0.1$, $4.3{\pm}0.1$, and $4.34{\pm}0.08$ for ${UO_2}^{2+}$, $(UO_2)_2{(OH)_2}^{2+}$, and $(UO_2)_3{(OH)_5}^+$, respectively. The results of Stern-Volmer analysis suggest that both mono- and bi-dentate U(VI)-DHB complexes serve as groundstate complexes inducing static quenching.

• Journal of the Korean Ceramic Society
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• v.32 no.4
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• pp.471-481
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• 1995

Unirradiated UO2 pellets were pulverized by oxidation in air at 40$0^{\circ}C$, and the oxidized powders were reduced in H2 and CO atmospheres at $600^{\circ}C$. During the oxidation of UO2 at 40$0^{\circ}C$, intergranular cracks which caused the spallation were mainly developed by the volume contraction due to the formation of intermediate phase (U4O9 or U3O7). As oxidation proceeded, U3O8 finally formed. As the oxidation/reduction cycles were repeated, the powder surface became coarser, specific surface area was increased and average particle size was decreased. The sintered densities of the powder were increased by the oxidation/reduction cycle due to the characteristic changes of the powder.

• Journal of the Korean Ceramic Society
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• v.33 no.6
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• pp.631-636
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• 1996

The powder characteristics of UO2 via AUC prepared by precipitation from a UN with AC soiution produced from nuclear fuel powder conversion plant and that of the existing facility were compared. Mean particle size of AUC powder was decreased and agglomerates were much occured in case of using the AC solution that that of the gases but other properties such as particle size distribution and shape of particle are thought to be similarly. In compaction of UO2 powder the breaking pressur of agglomerated UO2 powder and the sintered density of final UO2 pellet from AC solution were measured 1.45$\times$108 N/m2 and 10.52 g/cc, These values could be used in nuclear fuel powder fabrication process.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.259-267
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• 2018

Focused ion beam-scanning electron microscope and electron backscattered diffraction examinations were conducted in the center of a $73\;GWd/t_U\;UO_2$ fuel. They showed the formation of subdomains within the initial grains. The local crystal orientations in these domains were close to that of the original grain. Most of the fission gas bubbles were located on the boundaries. Their shapes were far from spherical and far from lenticular. No interlinked bubble network was found. These observations shed light on previous unexplained observations. They plead for a revision of the classical description of fission gas release mechanisms for the center of high burn-up $UO_2$. Yet, complementary detailed observations are needed to better understand the mechanisms involved.

• Journal of Powder Materials
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• v.14 no.1 s.60
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• pp.8-12
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• 2007

The effect of a CAM (Continuous Attrition Mill) treatment on the sinterability of ex-ADU $UO_2$ powder was investigated. As the cycles of a CAM increased, the apparent density, specific surface area and O/U of the milled powder increased, but there particle sizes decreased. However the sintered density of the $UO_2$ pellet decreased as the cycles of the CAM increased. It is considered that the decrease of the sintered density is due to the formation of $U_3O_8$, which was produced by a CAM mechanism.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.1002-1008
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• 2016

An understanding of the coarsening process of the large fission gas pores in the high burn-up structure (HBS) of irradiated $UO_2$ fuel is very necessary for analyzing the safety and reliability of fuel rods in a reactor. A numerical model for the description of pore coarsening in the HBS based on the Ostwald ripening mechanism, which has successfully explained the coarsening process of precipitates in solids is developed. In this model, the fission gas atoms are treated as the special precipitates in the irradiated $UO_2$ fuel matrix. The calculated results indicate that the significant pore coarsening and mean pore density decrease in the HBS occur upon surpassing a local burn-up of 100 GWd/tM. The capability of this model is successfully validated against irradiation experiments of $UO_2$ fuel, in which the average pore radius, pore density, and porosity are directly measured as functions of local burn-up. Comparisons with experimental data show that, when the local burn-up exceeds 100 GWd/tM, the calculated results agree well with the measured data.

• Nuclear Engineering and Technology
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• v.43 no.3
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• pp.309-316
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• 2011

Molybdenum is the most abundant fission product since its fission yield is equivalent to that of xenon, and it has a very special role in the chemistry of nuclear fuel because it influences the oxygen potential of $UO_2$ fuel. In this study, the distribution of molybdenum in spent $UO_2$ fuel specimens with 33.3, 41.0 and 57.6 GWd/tU burnup was measured by a LA-ICP-MS system and the reproducibility of the measured data was obtained. The Mo distribution was almost constant along the radius of a fuel except an increase at the periphery of the fuel. It showed a drop in reproducibility with relatively high deviation of measured values for the highest burnup fuel. To explain this, the state of molybdenum in a $UO_2$ matrix and its effect on the oxidation behavior of $UO_2$ were investigated. The low reproducibility was explained by the segregation of molybdenum, and the inhibition of oxidation by the molybdenum was also observed.