• Journal of computational fluids engineering
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• v.21 no.4
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• pp.54-60
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• 2016

The valve used in nuclear power plant must be qualified but the limitation of the test facility leads to use the numerical analysis. The flow coefficient is calculated with the consideration of the pressure, velocity and geometry. And the flow coefficient is the important physical property which is prepared using experiment or analysis by valve manufacturer. In this study, the analysis model was made according to ISA 75.02.01 and the mass flow rate and pressure drop ratio was calculated. The model of the expansion factor was applied to the simulation result and the pressure drop ratio at the start of the choked flow in the valve was found. With the simulation result, the consideration was performed that the expansion factor is the important physical property to the system engineer in addition to the flow coefficient.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2005.11a
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• pp.394-399
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• 2005

A case history is presented pertaining to high frequency piping vibration and noise caused by globe valve in the spent fuel pool cooling system of nuclear power plant. Frequency analyses were performed on the system to diagnose the problem and develop a solution to reduce the piping vibration and noise. The source of the high frequency and noise energy was traced to the globe valve located immediately downstream of the centrifugal pump by performing valve throttling test. Measurements of vibration and noise are presented to show that the high frequency vibration and noise amplitude was dependent upon the valve disc position and flow rate. Strouhal vortex shedding frequencies were generated at the exit of the globe valve which exited structural resonance of valve disc and amplified the high frequency vibration and noise. The problem was identified as an interaction between the flow inside globe valve and the valve disc structure. Attempts to reduce the vibration and noise amplitudes of the piping system were successfully achieved by the modification of guide-disc diameter and disc-edge figure The valve disc was replaced by an alternative to eliminate the source of the harmful high frequency vibration and noise.

• 유체기계공업학회:학술대회논문집
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• 2002.12a
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• pp.493-498
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• 2002

The letdown system of pressurized water reactor (PWR) nuclear fewer plants had experienced instabilities in letdown system due to unacceptable flow characteristics of control valves. The Korean Standard Nuclear Power Plants (KSNPs) have three flow paths in parallel for letdown new control. Each flow path consists of two offices and one isolation valve. This study evaluates the effect of orifice arrangement and valve stroke time of letdown isolation valve on the system transients because sudden flow changes due to valve actuation can generate high pressure peaks in letdown line. A pressure transient analysis has been preformed to evaluate the impact of dynamic transients. This analysis uses MMS which is a simulation code developed by EPRI based on the method of characteristics. The result shows that the pressure peak is reduced in the continuous arrangement but negligible. Additionally, it shows that the stroke time of linear type flog valve greater than 15 seconds can give more stable performance.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.32 no.5
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• pp.400-405
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• 2008

Air-operated valve is one of principal valves that are used to control fluid flow in nuclear power plants. It is necessary to evaluate periodically the performance of the AOV(Air-Operated Valve) which is used for controlling flow and pressure in nuclear power plants. In this study, we developed the integrated program HAES including hardware system to evaluate the performance of AOV. HAES(HANVIT AOV Evaluation System) consists of many modules such as valve information, thrust and torque analysis, actuator analysis, weak analysis and margin analysis. We applied with the test valve and showed that results evaluated by HAES is similar to those offered in the plant.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2000.11a
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• pp.138-144
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• 2000

Degradation occurred at the 6 inch swing check valve in nuclear power plant. Valve replacement and maintenance were carried out during the plant O/H. This report examined the degradation mechanism of the 6 inch check valve by the experimental and theoretical study. Results shows that the degradation was caused by valve chattering which due to the structural and acoustic resonance.

• Proceedings of the KIEE Conference
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• 2002.11c
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• pp.537-541
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• 2002

This paper is about the development of the turbine valve actuator efficiency analysis system for the standard power plants. We developed hydraulic power unit and turbine valve actuator controller. We designed control algorithm for turbine valve actuator, implemented and verified it at the industrial plants.

• 유체기계공업학회:학술대회논문집
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• 2005.12a
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• pp.607-612
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• 2005

Check valves playa vital role in the operation and protection of nuclear power plants. Check valves failure in nuclear power plants often lead to a plant transient or trip. An overview of the failure history of check valves needs to identify key area where resources can be best applied to further improve their reliability, and provide cost effective means for failure reduction. The analysis of historical failure data gives information on the populations of various types of check valves, the systems they are installed in, failure modes, effects, methods of detection, and the mechanisms of the failures. The results presented are based on information derived from operating records, nuclear industry reports, manufacturer supplied information. A majority of check valve failures are caused by improper application. Failure modes are identified for swing and lift check valves. Failures involving improper seating and valve disc stuck comprised the largest percentage of failures.

• Nuclear Engineering and Technology
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• v.38 no.2
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• pp.155-162
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• 2006

It is necessary to monitor periodically the operability of safety-related motor-operated valves (MOVs) in nuclear power plants. However, acquiring diagnostic signals for MOVs is very difficult, and doing so requires an excessive amount of time, effort, and expenditure. This paper introduces an accurate and economical method to evaluate the performance of MOVs remotely. The technique to be utilized includes electrical measurements and signal processing to estimate the motor torque and the stem thrust, which have been cited as the two most effective parameters in diagnosing MOVs by the US Nuclear Regulatory Commission. The motor torque is calculated by using electrical signals, which can be measured in the motor control center (MCC). Some advantages of using the motor torque signature over other signatures are examined. The stem thrust is calculated considering the characteristics of the MOV and the estimated motor torque. The basic principle of estimating stem thrust is explained. The developed method is implemented in diagnostic equipment, namely, the Motor Operated Valve Intelligent Diagnostic System (MOVIDS), which is used to obtain the accuracy of and to validate the applicability of the developed method in nuclear power plants. Finally, the accuracy of the developed method is presented and some examples applied to field data are discussed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.1
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• pp.69-73
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• 2015

The study investigated effects of residual stresses and Charpy impact energy on brittle fractures of the butt weld between the valve and the piping, and of the valve body in nuclear power plants via a linear elastic fracture mechanics approach in the ASME B&PV Code, Sec.XI and finite element analysis. Weld residual stress in a butt weld between close check valve and piping, and residual stress in the valve due to casting process were assumed to be proportional to yield strength of base metal. Operating stresses in the butt weld and the valve body were calculated using approximate engineering formulae and finite element analysis, respectively. Applied stress intensity factors were calculated by assuming postulated cracks with specific sizes and then by substituting the residual stresses and the operating stresses into engineering formulae presented in the ASME B&PV Code, Sec.III. Plane strain fracture toughness was derived by using a correlation between Charpy V-notch impact energy and fracture toughness. Structural integrity of the weld and the body against brittle fracture was assessed by using the applied stress intensity factors, plane strain fracture toughness and the linear elastic fracture mechanics approach. As a result, it was identified that the structural integrity was maintained with decreasing the residual stress levels and increasing the Charpy V-notch impact energy.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.157-158
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• 2004

To investigate the effects of the upstream flow conditions on the opening performance of the swing check valve, the tests were performed at various conditions of the disturbance type and distance from the tested check valves. The results show that the upstream flow disturbances due to elbow and globe valve at $2{\sim}10$ diameters upstream of the check valve produced minor effects on the check valve performance compared to the uniform flow condition. However, more detail analysis and additional testing with the other disturbance source, such as orifice with a large number of holes, are needed to refine and confirm the present results.