• Proceedings of the IEEK Conference
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• 2008.06a
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• pp.1195-1196
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• 2008

It is difficult to solve PCB(Printed Circuit Board) Noise problem. Because Electronic circuit system operates very high frequency. Resonance analysis of PCB layout by PI(Power Integrity) Simulation method visualizes distribution of Switching noise between VDD and GND. By using de-cap, we reduce impedance and solve the EMI problems.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.85-89
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• 2004

The inner side between HP stationary blades in #1 turbine of Nuclear Power Plant A is damaged by the FAC(flow assisted corrosion) which is exposed to moisture. For many years the inner side is repaired by welding the damaged part, however, FAC continues to deteriorate the original material of the welded blade ring. In this study, we have two stages to verify the integrity of stationary blade ring in nuclear power plant A. In the stage I, replication of blade ring is performed to survey the microstructure of blade ring. In the stage II, the stress analysis of blade ring is performed to verify the structural safety of blade ring. Throughout the two stages analysis of blade ring, the stationary blade ring had remained undamaged.

• Journal of Mechanical Science and Technology
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• v.19 no.11
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• pp.1988-1997
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• 2005

This study is concerned with structural integrity evaluations for the interference-fit flywheels in reactor coolant pumps (RCPs) of nuclear power plants. Stresses in the flywheel due to the shrinkage loads and centrifugal loads at the RCP normal operation speed, design overspeed and joint-release speed are obtained using the finite element method (FEM), where release of the deformation-controlled stresses as a result of structural interactions during rotation is considered. Fracture mechanics evaluations for a series of cracks assumed to exist in the flywheel are conducted, considering ductile (fatigue) and non-ductile fracture, and stress intensity factors are obtained for the cracks using the finite element alternating method (FEAM). From analysis results, it is found that fatigue crack growth rates calculated are negligible for smaller cracks. Meanwhile, the material resistance to non-ductile fracture in terms of the critical stress intensity factor (K$_{IC}$) and the nil-ductility transition reference temperature (RT$_{NDT}$) are governing factors for larger cracks.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.14 no.7
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• pp.3121-3126
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• 2013

The integrity evaluation of pipes in nuclear power plant are essential for the safety of reactor vessel, and integrity must be assured when flaws are found. Accurate stress intensity analyses and crack growth rate data of surface-cracked components are needed for reliable prediction of their fatigue life and fracture strengths. Fatigue design and life assessment are the essential technologies to design the structures such as pipe, industrial plant equipment and so on. The effect of crack spacing on stress intensity factor K values was studied using three-dimensional finite element method (FEM). For the case of cylinder under internal pressure, a significant increase in K values observed at the deepest point of the surface crack. Also, this paper describes the fatigue analysis for cracked structures submitted to bending loads.

• Journal of Power System Engineering
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• v.10 no.4
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• pp.71-77
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• 2006

This study introduces the seismic qualification of safety related equipments for nuclear power plants to verify the possibility of resonance in regard to the operating speed and the structural integrity due to external piping nozzle loads as well as seismic dynamic loads using El-Centro earthquake, which was occurred in the 1940's previously. As a first step, it is necessary to investigate the natural frequency of the vertical mixed flow pump in order to determine whether static or dynamic equipment comparing with seismic cut-off frequency, 33hz. Also the normal mode analysis was carried out with the introduction of seismic redesign straint at the middle of vertical pump to increase the natural frequency. In terms of structural integrity, the application of static analysis with normal, upset and faulted nozzle loads event was presented for the comparison of material allowable stress. Also the dynamic analysis was performed to show the design adequacy through the application to the case of El-Centro earthquake.

• The KSFM Journal of Fluid Machinery
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• v.3 no.4 s.9
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• pp.7-14
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• 2000

This paper provides integrity evaluation and root cause analysis for defects observed at volute tongue, or cutwater, of the operating centrifugal pump in power plant. The cause of the cracks are analyzed and reviewed from the viewpoint of the operation and maintenance of the pumps, and the sample obtained from the cracked volute tongue of the pump are examined. At first, in-situ hardness test and microstructure examination were performed to understand the cause of cracking at volute tongue. The evaluation of structural integrity and the possibility of the crack propagation is also evaluated. Cracks were typical intergranular cracking and propagated along with prior austenite grain boundary. At easing volute tongue, the hardness was higher than ASTM requirement and a large amount of intergranular Cr carbide was precipitated. These were due to high C content in material. P content was also higher than ASTM requirement. Therefore, Cr carbide precipitation and P segregation at grain boundary, caused by higher C and P content in material, resulted in intergranular cracking of casing volute tongue. This procedure for integrity evaluation and root cause analysis is used to guide, and support the pump designer and manufacturer's material selection and process design to avoid a costly, unplanned outage of plant.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2014.10a
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• pp.839-841
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• 2014

The Nuclear Steam Supply System(NSSS) is designed to provide an integrated approach that includes areas of monitoring relevant to the integrity of the NSSS. LPMS is designed to function as an alarm system by providing sensor channel alarms for the associated subsystems. LPMS is equipped to provide analysis tools for new alarm events, historical events and for historical periodically stored channel data (e.g. waveforms) for most channels. This paper is intended to introduce the diagnosis principle and abnormal symptom of loose parts monitoring system as a monitoring tool in Nuclear Steam Supply System. And also, we are going to introduce signal analysis program in order to perform the actual diagnosis in power plants.

• Nuclear Engineering and Technology
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• v.34 no.2
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• pp.105-116
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• 2002

In special industrial fields such 3s nuclear power plants and chemical plants, it is often necessary to repair system components without plant shutdown or drainage of system having many piping structures which may have hazardous or expensive fluid. A temporary ice plugging method for blocking internal flow is considered as a useful method in that case. According to the pipe freezing guideline of the nuclear power plant, the length of a freezing jacket must be longer than twice of the pipe diameter. However, for applying the ice plugging to short pipes which do not have enough freezing length because of geometrical configuration, it is inevitable to use shorter jacket less than twice of the pipe diameter. In this study, the integrity evaluation for short pipes in the nuclear power plant Is conducted by an experiment and the finite element analysis. From the results, the ice plugging process in short pipes can be safely carried out without any plastic deformation and fracture.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.3
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• pp.231-236
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• 2005

HIC transport packaging to transport a high integrity container(HIC) containing dry spent resin generated from nuclear power plants is to comply with the regulatory requirements of Korea and IAEA for Type B packaging due to the high radioactivity of the content, and to maintain the structural integrity under normal and accident conditions. It must withstand 9 m free drop impact onto an unyielding surface and 1 m drop impact onto a mild steel bar in a position causing maximum damage. For the conceptual design of a cylindrical HIC transport package, three dimensional dynamic structural analysis to ensure that the integrity of the package is maintained under all credible loads for 9 m free drop and 1 m puncture conditions were carried out using ABAQUS code.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.20 no.1
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• pp.39-48
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• 2024

Research reactors are operated under ambient temperature and atmospheric pressure, which is much less severe conditions compared to those in typical nuclear power plants. Due to the high temperature, heat resistant materials such as austenite stainless steel should be used for the reactors in typical nuclear power plants. Whereas, as the effect of temperature is low for research reactors, materials with high resistance to neutron irradiation, such as zircaloy and beryllium, are used. Therefore, these conditions should be considered when performing integrity assessment for research reactors. In this study, a computational technique through finite element (FE) analysis was developed considering the operating conditions and materials of research reactor when conducting integrity assessment. Neutron irradiation analysis techniques using thermal expansion analysis were proposed to consider neutron irradiation growth and swelling in zirconium alloys and beryllium. A user subroutine program that can calculate the strain rate induced by neutron irradiation creep was developed for use in the commercial analysis program Abaqus. To validate the proposed technique and the user subroutine, FE analysis results were compared with hand-calculation results, and showed good agreement. Consequently, developed technique and user subroutine are suitable for evaluating structural integrity of research reactors.