• Journal of the Korean Society for Precision Engineering
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• v.19 no.2
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• pp.65-71
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• 2002

Turbine blade is subject to cyclic bending force by steam pressure. Stress analysis by fractography is already established technology as means far seeking cause of fracture and has been widely employed. In the X-ray frctography, plastic deformation and residual stress near the fracture surface can be determined and information of internal structure of material can be obtained. Therefore, to find a fracture mechanism of torsion-mounted blade in nuclear power plant, based on the information from the fracture surface obtained by fatigue test, the correlation of X-ray parameter and fracture mechanics parameter was determined and then the stress intensity factor to actual broken turbine blade was predicted.

• Journal of the Korean Society for Precision Engineering
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• v.17 no.4
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• pp.85-96
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• 2000

An Eggcrate assembly is an important part to hold and support 16,000 tubes containing hot and contaminated water in the steam generator of nuclear power plant. As a great number of tubes should be inserted into the eggcrate assembly, the dimensions of each eggcrate grid are one of the critical factors to determine the availability of tube insertion. in this paper. we propose a machine vision algorithm for measuring the inner-circle diameter of each eggcrate grid whose shape is not exact quadrangular. The overall procedure of the algorithm is composed of camera calibration, eggcrate image preprocessing, grid height adjustment, and inner-circle diameter estimation. The algorithm is tested on real specimens and the results show that the algorithm works fairly well.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 1995.10a
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• pp.125-128
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• 1995

The remote sleeve repair-welding technology using the pulsed Nd:YAG laser for increasing the lifetime of the steam generator tube in the nuclear power plant has been developed. The laser welding has many advantages on deep penetration depth and narrow heat affect zone(HAZ). The inconel 600 tube and inconel 690 sleeve used high temperature and high pressure service have been welded as round lap welds. It is found that the relation between the connection width and welding parameters. It is found that the shear strength in proportion to the connection width by conducting tensile-shear tests.

• 한국연소학회:학술대회논문집
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• 2005.10a
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• pp.93-96
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• 2005

High temperature furnace such as Steam power plant and incinerator contribute considerable part of NOx generation and face urgent demand of De-NOx system. Reducing agents are necessary to use De-NOx system. In this study mixing caused by direct injection of reducing agent solution spray into flue gas duct was measured. Carbonated water was used as tracer and simulated agent because ammonia as a reducing agent is not proper to experiment. Mixing and evaporation must occur simultaneously and quickly enough to achieve desirable efficiency. To achieve that, the angle of attack of static mixer and the location is simulated and $CO_2$ concentration is measured.

• Proceedings of the Korean Institute of Intelligent Systems Conference
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• 1997.10a
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• pp.441-444
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• 1997

증기 발생기 수위 제어기의 성능 향상은 발전소의 정기 횟수를 줄여 발전소 신뢰도 및 가동률을 향상시키고 또한 기타 여러 부품의 수명에도 영향을 주어 경제적으로 보다 효율적인 발전소 운영에 기여한다. 이러한 수위 제어의 발전을 위해서 본 연구에서는 E. Irvingd의 모델을 사용하였다. E. Irving이 모델이 단순화한 관계로 단점을 가지고는 있으나 프로그램화가 편리하고, 또한 증기 발생기의 특성을 잘 표현하기 때문에 이용하였다. 먼저 시스템의 출력, 즉 증기 발생기의 수위를 안정화시키기 위하여 퍼지 제어기를 Case by Case로 선정하여 제어를 하였으며, 그 다음으로 시스템의 두 입력, 증기량과 퍼지 제어기에서 선택되어진 급수 유량, 그리고 전 단계의 출력인 증기 발생기의 수위를 입력으로 하는 신경 회로망을 이용하여 시스템을 규명하였다.

• 유체기계공업학회:학술대회논문집
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• 2006.08a
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• pp.527-530
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• 2006

In the present work, characteristics of the flow in the cage of a steam turbine bypass control valve for thermal power plant are investigated. Experimental measurement for wall static pressure has been carried out to validate numerical solutions. And, the flowfield is analyzed by solving steady three-dimensional Reynolds-averaged Navier-Stokes equations. Shear stress transport (SST) model is used as turbulence closure. The effects of the flow area between stages of the cage on the pressure drop are also found.

• Proceedings of the KSME Conference
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• 2003.04a
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• pp.629-636
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• 2003

The purpose of this study is localization of safety relief valves for Nuclear Service through technical development with overall design, fabrication, inspection, capacity certification test and functional qualification test of safety relief valves in accordance with ASME Section III and KEPIC Code. Safety relief valve is the important equipment used to protect the pressure vessel, the steam generator and the other pressure facility from overpressure by discharging the operating medium when the pressure of system is reaching the design pressure of the system. But we're depending on technology of the other country up to the present time. Because we don‘ have our own technologies, we have been spent the great time and money on installing and repairing safety relief valve at nuclear power plant. Therefore we have to achieve the development of safety relief valves for Nuclear Service with our own technologies.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2001.04a
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• pp.778-783
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• 2001

Turbine blade is subject to cyclic bending force by steam pressure. Stress analysis by fractography is already established technology as means for seeking cause of fracture and has been widely employed. In the X-ray fractography, plastic deformation and residual stress near the fracture surface can by determined and information of internal structure of material can be obtained. Therefore, to find a fracture mechanism of torsion-mounted blade in nuclear power plant, based on the information from the fracture surface obtained by fatigue test, the correlation of X-ray parameter and fracture mechanics parameter was determined and then the load applied to actual broken turbine blade was predicted.

• Nuclear Engineering and Technology
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• v.47 no.2
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• pp.148-156
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• 2015

Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs) that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C) strategies for a large 1,000 MWe iPWR is described. Reactor system modeling-which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum-is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

• Proceedings of the KSME Conference
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• 2000.11a
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• pp.730-735
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• 2000

The finite element analysis by the computer program SYSWELD in consideration of phase transformation was carried out to simulate the multi-pass welding process of SA106 Gr. C which is used for the main steam pipe in nuclear power plant. All the numerical results such as temperatures, the size of heat affected zone and the residual stresses were compared to the experimental results.