• Journal of Advanced Marine Engineering and Technology
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• v.19 no.1
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• pp.28-35
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• 1995

The fluid flow for a energy transfer is essential for the design and operation of power plants, petrochemical plants and ships including a process. When the operating conditions of a plant are changed or any transitional event occured, the flow controls of a fluid must be performed to follow the new operating state or mitigate the results of a event. Generally these flow controls to accommodate the new operating state of a plant are made by the use of various valves. The refore the design of valves and the related techniques are very important to the system and component designs. However the system and component design are not familiar with the practical theory of the valve since the derivative procedures of the flow equations in a valve are difficult and it is not easy to found the theoretical foundamentals and informations about the design of a valve from the present references. In this study the flow equations applicable to a valve for liquid are theoretically derived in detail. And the definition of valve reynolds number and its boundary values between the tubulent and laminar flow is described compared with the values of a circular pipe flow.

• Proceedings of the Korean Society of Machine Tool Engineers Conference
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• 2001.10a
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• pp.304-310
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• 2001

Flow pulsation often causes vibration and noise in piping systems and therefore has been a troublesome concern for fluid system engineers. According to frequency increase in this paper under the influence wave form of velocity in springly flow and viscosity are drop coefficient of viscosity become increase so that impedance and resistance. The transient variations of flow rate are measured by a modified impedance tube method which is realized by virtue of the present analytical technique. At pipe line in order to eliminate vibration, confirm happened intermittently impedance characteristics. We make a test and frequency analysis and have to minimize obstructive component at hydraulic circuit.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 1997.10a
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• pp.232-238
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• 1997

In this paper, a theoretical method is presented to design a hydraulic control valve system that consist of an important component in the hydraulic snubber. The hydraulic snubber is used essentially to support the piping systems at power plants. To calculate the force due to pressure drop and flow rate in the valve orifice and by-pass hole, Bernoulli equation is used. The Reynolds equation are numerically analyzed in the clearance gap between the valve cone and valve seat to estimate the friction force and leakage flow rate. Based on the detailed theoretical data, we developed successfully the hydraulic snubber for power plants.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.1
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• pp.37-42
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• 2010

The stay cylinder weld at the steam generator of Korean Standard Nuclear Power Plants is safety class I component and is subjected to be inspected by the volumetric examination such as ultrasonic method. As accessibility of this area is limited due to the narrow space and high radiation, the existing manual inspection method involves various difficulties. Moreover operators may be exposed to internal contamination by contaminated dust during the surface buffing process to improve the inspection reliability of this area. Recently the new automatic inspection system for stay cylinder welds has been developed. The inspection system basically consists of a driving assembly, data acquisition device and signal processing units. The driving assembly is classified by 1) the scanner for inspecting and buffing the weld, 2) pillars for guiding the scanner and 3) the base frame for loading and supporting pillars. The scanner has 4 sensor modules to inspect in 4 refracted angles and 4 incident directions. These components can be inserted into the skirt of the stay cylinder through the manway hole and assembled easily by one-touch in the skirt. Data acquisition device and signal processing units developed in previous works are also newly upgraded for better processing of data analysis and evaluation. The system has been successfully demonstrated not only in the mock-up but also in the field. In this paper, newly developed inspection system for the stay cylinder weld of the steam generator is introduced and their field applications are discussed.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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• pp.40-45
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• 2008

There exist many types of pipe and component fatigue through vibrations, thermal fatigues or shifting. In some cases of thermal stratification/thermal fatigue, pipes & components are receiving thermal stress by means of material expansion and shrinkage by continuous thermal repetitive variation. Small cracks initially occur on the inside surface by thermal stress. These cracks grow in depth the pipe wall and finally come to a rupture. Pipe parts of susceptibility to thermal stratification and thermal fatigue are now being examined by conventional UT(ultrasonic test) as volumetric examination. It is difficult to fully satisfy the code & standards requirements because 3" weldolet weldments of RCS 16" pipe to 3" branch connection lines have complex structural shape. To solve the problems of conventional UT examination, we made a realistic mock-up and UT calibration block. We performed a simulation of phased array UT utilizing CIVA as NDE(Non-Destructive Examination) simulation software. Also we designed phased array UT transducer and wedge, optimal frequency by using simulation data. We performed phased array UT experiment through mock-up including artificial flaws(notch). The phased array UT technique is finally developed to improve the reliability of ultrasonic test at RCS 16" pipe to 3" branch connection weld.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.13 no.2
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• pp.1-18
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• 2017

Metal fatigue is an important aging mechanism that material characteristics can be deteriorated when even a small load is applied repeatedly. An accurate fatigue evaluation is very important for component structural integrity and reliability. In the design stage of a nuclear power plant, the fatigue evaluations of the Class 1 components have to be performed. However, operating experience shows that the design evaluation can be very conservative due to conservatism in the transient severity and number of occurrence. Therefore, the fatigue monitoring system has been considered as a practical mean to ensure safe operation of the nuclear power plants. The fatigue monitoring system can quantify accumulated fatigue damage up to date for various plant conditions. The purpose of this paper is to describe the fatigue monitoring procedure and to introduce the fatigue monitoring program developed by the authors. The feasibility of the fatigue monitoring program is demonstrated by comparing with the actual operating data and finite element analysis results.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.20-27
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• 2010

In order to prevent the corrosion of component contacted primary water designed alloy 600 material in the nuclear power plant. But the primary water stress corrosion cracking(PWSCC) of alloy 600 and weld area occurs continuously due to the residual stress. The leakage accident resulted from PWSCC in the drain nozzle of the steam generator of domestic power plants. Heater sleeves of the pressurizer are welded with alloy 600 weld material and therefore exposed to the primary water environment. PWSCC occurred in heater sleeve material and weld area of many foreign power plants. The current issue of domestic nuclear power plants are consequently concentrated to PWSCC of similar material. In order to improve the detection and the sizing of the PWSCC in the welding sleeve of the pressurizer, the automatic UT system and multi-directions probe sets have been developed. The experimental studies have been performed using the mock-up block containing artificial reflectors(ID connected EDM notch) and semi-artificial cracks made from thermal fatigue. The automatic UT System is applied in the detection and the length sizing of the ID/OD on the tube and the J-groove weld area of the artificial reflectors and results of the detection and the sizing are compared respectively. Also, the developed automatic UT system is successfully accomplished to inspect the heater sleeve and the J-groove weld area on the pressurizer for the detection of PWSCC.

• Journal of the Korean GEO-environmental Society
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• v.11 no.9
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• pp.15-25
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• 2010

The main objective of this study was to offer informations about long-term seepage behavioral characteristics and to find a leakage safety management method for Juam Dam and Imha Dam, a central cored rockfill dams in Korea by the evaluating the automatically monitored leakage data. In the water leakage monitoring of fill dam, the generation of abnormal water leakage is difficult to directly detect due to the effect of outside factors such as the component of rainfall inherent in the observation value. Therefore, conventionally estimation methods of water leakage quantity were applied by multiple regression analysis considering reservoir water level, rainfall, etc.. However, the estimated error of rainfall component is relatively big in these method. This paper identifies the seepage characteristic of each dams which is not directly affected by rainfall through the hydrograph separation analysis and 3 dimensional analytical method, and thinks a leakage management method. It was noticed that two dams had site specific seepage behaviour features and were in stable state with the decreasing leakage quantity. It was also found that hydrograph separation method might be applicable to leakage safety management method.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2119-2129
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• 2020

Flow-accelerated corrosion (FAC), liquid droplet impingement erosion (LDIE), cavitation and flashing can cause continuous wall-thinning in nuclear secondary pipes. In order to prevent pipe rupture events resulting from the wall-thinning, most NPPs (nuclear power plants) implement their management programs, which include periodic thickness inspection using UT (ultrasonic test). Meanwhile, it is well known in field experiences that the thickness measurement errors (or deviations) are often comparable with the amount of thickness reduction. Because of these errors, it is difficult to estimate wall-thinning exactly whether the significant thinning has occurred in the inspected components or not. In the previous study, the authors presented an approximate estimation procedure as the first step for thickness measurement deviations at each inspected component and the statistical & quantitative characteristics of the measurement deviations using plant experience data. In this study, statistical significance was quantified for the current methods used for wall-thinning determination. Also, the authors proposed new estimation procedures for determining local wall-thinning to overcome the weakness of the current methods, in which the proposed procedure is based on analysis of variance (ANOVA) method using subgrouping of measured thinning values at all measurement grids. The new procedures were also quantified for their statistical significance. As the results, it is confirmed that the new methods have better estimation confidence than the methods having used until now.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.20 no.6
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• pp.1843-1853
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• 1996

The objective of this paper is to evaluate the crack stability of the nuclear main stresm pipes, considering the load reduction effect due to the presence of circumferential throuth-wall crack. Also, the optimization techniques are adoped tosimulate the crack effect on the elbow component of the piuping system. By using a general beam elemetn which contains a discontinuous cross-section, the piping analysis is accomplished to acquire the reduced load. Considering this reduced load, it is feasible for the LBB application in nuclear main stresm pipe. Also, by combining an optimization program and a genaral finite element analysis program, the appropriate dimensions of the simplified beam elemtn which represents the effect of crack in elbow could be successfully determined.