• Journal of Digital Convergence
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• v.13 no.4
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• pp.303-310
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• 2015

Game technologies are now applied in various engineering areas such as the simulation of surgical operation or the implementation of a cyber model house. One of the essential and important technology in these applications is cutting of the 3D polygon model in real time. Real-time cutting technology is an essential technology needed to implement the simulation of a building demolition or a car assembly for training or educational purpose. Previous cutting method using the conventional BSP-Tree structure has some limitations in that they divide the whole world including the 3D model and its environment, only into two parts with respect to an infinite plane. In this paper, we show a technique cutting the 3D model in a finite extent in order to solve this problem. Specifically, we restricted the cut surface in a finite rectangular area and constructed the mesh for the divided surface. To show the usefulness of our partial cutting technique, an example of the dismantling process simulation of a nuclear reactor polygon model was illustrated.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.37 no.9
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• pp.855-862
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• 2013

A series of 1/5 scale-down reactor flow distribution tests had been conducted to determine the hydraulic characteristics of an APR+ (Advanced Power Reactor Plus), which were used as the input data for an open core thermal margin analysis code. In this study, to examine the applicability of computational fluid dynamics with the porous model to the analysis of APR+ internal flow, simulations were conducted using the commercial multi-purpose computational fluid dynamics software ANSYS CFX V.14. It was concluded that the porous domain approach for some reactor internal structures could adequately predict the flow characteristics inside a reactor in a qualitative manner. If sufficient computational resources are available, the predicted core inlet flow distribution is expected to be more accurate by considering the real geometry of the internal structures, especially upstream of the core inlet.

• Journal of the Institute of Electronics Engineers of Korea SC
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• v.42 no.2 s.302
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• pp.33-38
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• 2005

In this paper, the sensing unit of a non-invasive blood glucose sensor for home users, using a magneto-resonance absorption method, have been designed and manufactured. The sensor is capable of non-invasively determining blood glucose levels through measuring the 1H spin-lattice relaxation time in human body, The comparison of initial models, with different dimensions and shapes, for the sensing unit has led us to select the materials of the final model, which has adequate size and weight for home use. Through the design optimization using the FEM model, the dimension of final model has been determined to satisfy the required strength and uniformity of the magnetic field in the detecting area.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.34 no.6
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• pp.385-392
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• 2021

Interest in the seismic performance of nuclear facilities under strong earthquakes has increased because their nonlinear response is important. In this paper, we proposed appropriate parameters for the nonlinear finite element analysis of a concrete material model, for a reinforced concrete (RC) shear wall in nuclear facilities: maximum tensile strength, dilation angle, and damage parameter. The study of the effects of the important parameters, on the nonlinear behavior and shear failure mode of the RC shear wall having low aspect ratio, was conducted using ABAQUS finite element analysis program. Based on the study results the nonlinear response of a nuclear reactor containment building (RCB) subjected to a strong earthquake was evaluated using nonlinear time-history analysis.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.356-361
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• 1996

물질전달과 열전달의 유사성을 이용하는 나프탈렌 승화법을 핵연료집합체 모델에 적용하여 봉다발에서의 국소 열전달 계수의 분포를 측정하였다. 실험 모델은 가압경수형 원자로에서 나타나는 부수로 즉, 벽면 부수로와 모서리 부수로 및 내부 부수로로 구성되는 3$\times$3 봉다발이다. 봉다발에서의 국소 열전달 계수 값은 부수로의 형상과 인접한 봉 및 벽면의 영향이 크게 작용하는 것으로 측정되었다. 내부 부수로에 둘러져 있는 봉에서의 국소 열전달계수값은 봉과 봉 사이에서는 부수로 중심 방향보다 낮았고, 평균열전달계수는 Dittus-Boelter의 상관식보다 약간 낮은 값을 보였다. 벽면 부수로에 인접한 봉에서의 열전달계수는 벽면의 영향으로 내부 부수로에 있는 봉보다 상대적으로 낮았으며, 모서리 부수로의 봉에서는 벽면의 영향이 증대되어 더욱 낮게 나타났다.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.27 no.4
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• pp.220-227
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• 2015

Complex thermal hydraulic characteristics exist inside the reactor because the reactor internals consist of fuel assembly, internal structures and so on. In this study, to examine the effect of Reynolds-Averaged Navier-Stokes (RANS)-based two-equation turbulence models in the analysis of flow distribution inside a 1/5 scaled-down APR+, simulation was performed using the commercial computational fluid dynamics software, ANSYS CFX R.13 and the predicted results were compared with the measured data. It was concluded that reactor internal flow pattern was locally different depending on the turbulence models. In addition, the prediction accuracy of k-${\varepsilon}$ model was superior to that of other two-equation turbulence models and this model predicted the relatively uniform distribution of core inlet flow rate.

• Journal of Energy Engineering
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• v.25 no.2
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• pp.1-20
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• 2016

A condensation heat transfer model is very important for the safety analysis of nuclear power plants. Especially, condensation under the presence of noncondensable gases (NCGs) is an important issue in nuclear safety because the presence of even a small quantity of NCGs in the vapor largely reduces the condensation rate. In this study, the condensation heat transfer model of the severe accident analysis code MELCOR 1.8.6 has been assessed using a set of condensation experiments performed under the thermal-hydraulic conditions similar to those inside a containment during design-basis accidents or severe accidents. Experiment conditions are categorized into 4 types according to the shape of the condensation surface: vertical flat plates, outer surface of vertical pipes, inner surface of vertical pipes, the inner surface of horizontal pipes. The results of the calculations show that the MELCOR code generally under-predicts the condensation heat transfer except the condensation on inner surface of vertical pipes.

• Computational Structural Engineering
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• v.5 no.3
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• pp.105-112
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• 1992

The design of reactor internals requires the accurate vibration characteristics of each component for subsequent dynamic structural response analysis. For Korean standard nuclear power plant some modifications on the Control Element Assembly shroud from the reference design have been made. Since the shroud is complex in geometry having an array of vertical round tubes and webs in a square grid pattern, and being tied down by preloaded tie rods into position, it is planned to perform a vibration measurement program consisting of both experimental and analytical modal studies upon that component. To determine the proper test conditions, the pre-test analysis has been performed using the general purpose structural analysis program ANSYS. Also the effects of the number of master degrees of freedom, holes in the web and tie-rod preload on the natural frequencies are examined prior to the pre-test analysis. After decision of appropriate finite element model, frequency analysis and harmonic analysis are performed and ideas for the test conditions such as the number of measurement points, their locations, measurement frequency range and the excitation force level are determined.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.25 no.5
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• pp.269-278
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• 2013

In this study, in order to examine the applicability of computational fluid dynamics with the porous model to the analysis of APR+ (Advanced Power Reactor Plus) internal flow, simulation was conducted with the commercial multi-purpose computational fluid dynamics software, ANSYS CFX V.14. In addition, among the various reactor internals, the effect of flow skirt geometry on reactor internal flow was investigated. It was concluded that the porous model for some reactor internal structures could adequately predict the hydraulic characteristics inside the reactor in a qualitative manner. If sufficient computation resource is available, the predicted core inlet flow distribution is expected to be more accurate, by considering the real geometry of the internal structures, especially located in the upstream of the core inlet. Finally, depending on the shape of the flow skirt, the flow distribution was somewhat different locally. The standard deviation of the mass flow rate (${\sigma}$) for the original shape of flow skirt was smaller, than that for the modified shape of flow skirt. This means that the original shape of the flow skirt may give a more uniform distribution of mass flow rate at the core inlet plane, which may be more desirable for the core cooling.

• Computational Structural Engineering
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• v.5 no.4
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• pp.93-102
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• 1992

The design of reactor internals requires the accurate vibration characteristics of each component for subsequent dynamic structural response analyses. For Korean standard nuclear power plant some modifications on the Control Element Assembly shroud from the reference design have been made, Since the shroud is complex in geometry having an array of vertical round tubes and webs in a square grid pattern, and being tied down by preloaded tie rods into position, it is planned to perform a vibration measurement program consisting of both experimental and analytical modal studies upon that component. The shroud modal testing was performed on the low frequency global survey to measure the first several modes. The analysis using the finite element model was also performed for the as-tested conditions. The natural frequencies and mode shapes from both test and analysis have been acquired and compared to be in good agreement. It is concluded that finite element model generated is good enough to be used in the design for the dynamic response analysis under various loading conditions.