• Journal of the Korea Society of Computer and Information
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• v.16 no.10
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• pp.83-92
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• 2011

The aim of this study is to evaluate whether 3D nuclear chromatin texture features are significant in recognizing the progression of cervical cancer. In particular, we assessed that our method could detect subtle differences in the chromatin pattern of seemingly normal cells on specimens with malignancy. We extracted nuclear texture features based on 3D GLCM(Gray Level Co occurrence Matrix) and 3D Wavelet transform from 100 cell volume data for each group (Normal, LSIL and HSIL). To evaluate the feasibility of 3D chromatin texture analysis, we compared the correct classification rate for each of the classifiers using them. In addition to this, we compared the correct classification rates for the classifiers using the proposed 3D nuclear texture features and the 2D nuclear texture features which were extracted in the same way. The results showed that the classifier using the 3D nuclear texture features provided better results. This means our method could improve the accuracy and reproducibility of quantification of cervical cell.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.1972-1981
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• 2022

A typical ferritic/martensitic (F/M) steel sheet was subjected to pulsed laser surface remelting (LSR) and corrosion test in lead-bismuth eutectic (LBE) at 550 ℃. There present two modification zones with distinct microstructures in the LSRed specimen: (1) remelted zone (RZ) consisting of both bulk δ-ferrite grains and martensitic plates and (2) heat-affected zone (HAZ) below the RZ, mainly composed of martensitic plates and high-density precipitates. Martensitic transformation occurs in both the RZ and the HAZ with the Kurdjumov-Sachs and Nishiyama-Wassermann orientation relationships followed concurrently, resulting in scattered orientations and specific misorientation characteristics. Hardnesses of the RZ and the HAZ are 364 ± 7 HV and 451 ± 15 HV, respectively, considerably higher than that of the matrix (267 ± 3 HV). In oxygen-saturated and oxygen-depleted LBE, thicknesses of oxide layers developed on both the as-received and the LSRed specimens increase with prolonging corrosion time (oxide layers always thinner under the oxygen-depleted condition). The corrosion resistance of the LSRed F/M steel in oxygen-saturated LBE is improved, which can be attributed to the grain-refinement accelerated formation of dense Fe-Cr spinel. In oxygen-depleted LBE, the growth of oxide layers is very low with both types of specimens showing similar corrosion resistance.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4809-4818
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• 2022

The decommissioning project of a nuclear facility is a large-scale process that is expected to take about 15 years or longer. The range of risks to be considered is large and complex, then, it is expected that various risks will arise in decision-making by area during the project. Therefore, in this study, the risk family derived from the Decommissioning Risk Management (DRiMa) project was reconstructed into a decommissioning project risk profile suitable for the Kori Unit 1. Two criteria of uncertainty and importance are considered in order to prioritize the selected 26 risks of decommissioning project. The uncertainty is scored according to the relevant laws and decommissioning plan preparation guidelines, and the project importance is scored according to the degree to which it primarily affects the triple constraints of the project. The results of risks are divided into high, medium, and low. Among them, 10 risks are identified as medium level and 16 risks are identified as low level. 10 risks, which are medium levels, are classified in five categories: End state of decommissioning project, Management of waste and materials, Decommissioning strategy and technology, Legal and regulatory framework, and Safety. This study is a preliminary assessment of the risk of the decommissioning project that could be considered in the preparation stage. Therefore, we expect that the project risks considered in this study can be used as an initial data for reevaluation by reflecting the detail project progress in future studies.

• Journal of Life Science
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• v.33 no.11
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• pp.851-858
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• 2023

Unique cartilage matrix-associated protein (UCMA) is an extrahepatic vitamin K-dependent protein rich in γ-carboxylated (Gla) residues. UCMA has been recognized for its ability to promote osteoblast differentiation and enhance bone formation; however, its impact on osteoblasts under hyperglycemic stress remains unknown. In this paper, we investigated the effect of UCMA on MC3T3-E1 osteoblastic cells under hyperglycemic conditions. After exposure to high glucose, the MC3T3-E1 cells were treated with recombinant UCMA proteins. CellROX and MitoSOX staining showed that the production of reactive oxygen species (ROS), which initially increased under high-glucose conditions in MC3T3-E1 cells, decreased after UCMA treatment. Additionally, quantitative polymerase chain reaction revealed increased expression of antioxidant genes, nuclear factor erythroid 2-related factor 2 and superoxide dismutase 1, in the MC3T3-E1 cells exposed to both high glucose and UCMA. UCMA treatment downregulated the expression of heme oxygenase-1, which reduced its translocation from the cytosol to the nucleus. Moreover, the expression of dynamin-related protein 1, a mitochondrial fission marker, was upregulated, and AKT signaling was inhibited after UCMA treatment. Overall, UCMA appears to mitigate ROS production, increase antioxidant gene expression, impact mitochondrial dynamics, and modulate AKT signaling in osteoblasts exposed to high-glucose conditions. This study advances our understanding of the cellular mechanism of UCMA and suggests its potential use as a novel therapeutic agent for bone complications related to metabolic disorders.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.259-264
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• 1993

Ion chromatographic method has been applied for burnup measurement of irradiated nuclear fuel by dynamic system using 1-octanesulfonate as a cation exchanger and $\alpha$-hydroxyisobutyric acid as an eluant. A number of elution techniques were evaluated for the optimum separation of plutonium, uranium and neodymium. These elements were individually separated and collected by gradient elution between 0.05 M and 0.40 M of $\alpha$-hydroxyisobutyric acid in a single column, and finally determined by isotope dilution mass spectrometry. The burnup data from this method were compared with those from conventional anion exchange method. The results showed a good agreement within 3.5 % of difference between two methods.

• Nuclear Engineering and Technology
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• v.47 no.5
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• pp.617-623
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• 2015

$ZrO_2$-based composites reinforced with 6.5 vol.% of carbon foam, carbon fiber, and graphite were fabricated using spark plasma sintering, and characterized using scanning electron microscopy and X-ray diffractometry. Their thermal properties were also investigated. The microstructures of the reinforced composites showed that carbon fiber fully reacted with $ZrO_2$, whereas carbon foam and graphite did not. The carbothermal reaction of carbon fiber had a negative effect on the thermal properties of the reinforced $ZrO_2$ composites because of the formation of zirconium oxycarbide. Meanwhile, the addition of carbon foam had a positive effect, increasing the thermal conductivity from 2.86 to $3.38Wm^{-1}K^{-1}$ at $1,100^{\circ}C$. These findings suggest that the homogenous distribution and chemical stability of reinforcement material affect the thermal properties of $ZrO_2$-based composites.

• 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.

• Journal of the Korean Ceramic Society
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• v.50 no.6
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• pp.378-383
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• 2013

Continuous fiber-reinforced ceramic matrix composites (CFCCs) have a complex distribution of porosity, consisting of interfiber micro pores and interbundle/interply macro pores. Owing to the complex geometry of the pores and fiber architecture, it is difficult to obtain representative microstructural features throughout the specimen volume with conventional, destructive ceramographic approaches. In this study, we introduce X-ray computed microtomography (X-ray ${\mu}CT$) to nondestructively analyze the microstructures of disk shaped and tubular $SiC_f$/SiC composites fabricated by the chemical vapor infiltration (CVI) method. The disk specimen made by stacking plain-woven SiC fabrics exhibited periodic, large fluctuation of porosity in the stacking direction but much less variation of porosity perpendicular to the fabric planes. The X-ray ${\mu}CT$ evaluation of the microstructure was also effectively utilized to improve the fabrication process of the triple-layered tubular SiC composite.

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2387-2393
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• 2020

In this study, a gradient material for shielding neutrons and gamma rays was developed, which consists of epoxy resin, boron carbide (B₄C), lead (Pb) and a little graphene oxide. It aims light weight and compact, which will be applied on the transportable nuclear reactor. The material is made up of sixteen layers, and the thickness and components of each layer were designed by genetic algorithm (GA) combined with Monte Carlo N Particle Transport (MCNP). In the experiment, the viscosities of the epoxy at different temperatures were tested, and the settlement regularity of Pb particles and B₄C particles in the epoxy was simulated by matlab software. The material was manufactured at 25 ℃, the Pb C and O elements of which were also tested, and the result was compared with the outcome of the simulation. Finally, the material's shielding performance was simulated by MCNP and compared with the uniformity material's. The result shows that the shielding performance of gradient material is more effective than that of the uniformity material, and the difference is most noticeable when the materials are 30 cm thick.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.317-326
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• 2005

The burning-out of excess plutonium from the reprocessing of spent nuclear fuel and from the dismantlement of nuclear weapons is recently emphasized due to the difficulties in securing the final repository for the spent fuel and the necessity to consume the ex-weapons plutonium. An irradiation test in the Halden reactor was launched by the OECD Halden Reactor Project (HRP) to investigate the in-pile behavior of plutonium-embedded fuel as a form of mixed oxide (MOX) and of inert matrix fuel (IMF). The first cycle of irradiation was successfully accomplished with good integrity of test fuel rods and without any undesirable fault of instrumentations. The test results revealed that the MOX fuel is more stable under irradiation environments than IMF. In addition, MOX fuel shows lower thermal resistance due to its better thermal conductivity than IMF. The on-line measured in-pile performance data of attrition milled MOX fuel are used in the analysis of the in-pile performance of the fuel with the fuel performance code, COSMOS. The COSMOS code has been developed for the analysis of MOX fuel as well as $UO_2$ fuel up to high burnup and showed good capability to analyze the in-reactor behavior of MOX fuel even with different instrumentation.