• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.1
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• pp.39-44
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• 2015

The fracture toughness of 2.25Cr-1Mo cast steel (SA217-WC9) samples which were taken from the check valves of feed water piping of Korean Standard Nuclear Power Plant(KSNPP) was measured by Master Curve method. The measured $T_0$ reference temperature of SA217-WC9 steel was $-30^{\circ}C$. The obtained $T_0$ was compared to the derived value from Charpy impact test data following to SINTEP procedure. The heat-to-heat variation in fracture toughness of SA217-WC9 steel was observed. It was found that the low toughness of a heat of SA217-WC9 steel was attributed to the coarse MnS inclusion originated by high sulfur content as the results of microanalyses.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.2794-2797
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• 2008

The pneumatic waste collection system, which is a complete solution for solving the waste collection problems, are constructed in many countries all over the world. However, research data for piping network design are insufficient. In this paper the pressure losses of the straight and curved pipes, pipe junctions are obtained using the numerical method in order to investigate the optimal pipe network design for the waste collection system.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.2
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• pp.61-67
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• 2015

3D FEM modeling based on 3D CAD data has been performed to evaluate the efficiency of CP system in a real operating nuclear power plant. The results of it successfully produced sophisticated profiles of electrolytic potential and current distributions in the soil of an interested area. This technology is expected to be a breakthrough for detection technology of damages on buried pipes when it comes into combining with a brand of area potential earth current (APEC) and ground penetrated radar (GPR) technologies. 2D current distribution and 2D current vectors on the earth surface from the APEC survey will be used as boundary conditions with exact 3D geometry data resulting in visualization of locations and extents of corrosion damages on the buried pipes in nuclear power plants.

• Nuclear Engineering and Technology
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• v.40 no.7
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• pp.607-614
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• 2008

Since pipes with wall-thinning defects can collapse at fluid pressure that are lower than expected, the collapse moment of wall-thinned pipes should be determined accurately for the safety of nuclear power plants. Wall-thinning defects, which are mostly found in pipe bends and elbows, are mainly caused by flow-accelerated corrosion. This lowers the failure pressure, load-carrying capacity, deformation ability, and fatigue resistance of pipe bends and elbows. This paper offers a support vector regression (SVR) model further enhanced with a fuzzy algorithm for calculation of the collapse moment and for evaluating the integrity of wall-thinned piping systems. The fuzzy support vector regression (FSVR) model is applied to numerical data obtained from finite element analyses of piping systems with wall-thinning defects. In this paper, three FSVR models are developed, respectively, for three data sets divided into extrados, intrados, and crown defects corresponding to three different defect locations. It is known that FSVR models are sufficiently accurate for an integrity evaluation of piping systems from laser or ultrasonic measurements of wall-thinning defects.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.1
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• pp.9-13
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• 2009

The containment local leakage rate testing in nuclear power plants is performed in accordance with ANSI/ANS-56.8(1994) in Korea. Two methods, the make-up flow rate and the pressure decay, are used for LLRT. Though ANSI/ANS-56.8 does not define clearly the minimum test duration for the make-up flow rate method, it requires obtaining the data after reaching the stable condition. Thus the prerequisite stable condition for data acquisition and the test period for type "C" LLRT is differently applied to each NPPs. Therefore, this study presents a unified test criteria for data stabilization and test duration through experiments to improve the test reliability for type "C".

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.42-47
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• 2012

The containment local leakage rate testing in nuclear power plants is performed in accordance with ANSI/ANS 56.8-1994 in Korea. Two methods, the make-up flow rate and the pressure decay, are used for local leakage rate testing. Though ANSI/ANS 56.8-1994 does not define clearly the minimum test duration for the make-up flow rate method, it requires obtaining the data after reaching the stable condition. Thus the prerequisite stable condition for data acquisition and the testing time is differently applied to each NPPs. Therefore, this study presents a standardized test procedure for data stabilization and testing time through experiments to improve the test reliability.

• Corrosion Science and Technology
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• v.15 no.4
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• pp.182-188
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• 2016

Significant piping wall thinning caused by Flow-Accelerated Corrosion (FAC) and Erosion-Corrosion (EC) continues to occur, even after the Mihama Power Station unit 3 secondary pipe rupture in 2004, in which workers were seriously injured or died. Nuclear power plants in many countries have experienced FAC and EC-related cases in steam cycle piping systems. Korea has also experienced piping wall thinning cases including thinning in the downstream straight pipe of a check valve in a feedwater pump line, the downstream elbow of a control valve in a feedwater flow control line, and failure of the straight pipe downstream of an orifice in an auxiliary steam return line. Cause analyses were performed by reviewing thickness data using Ultrasonic Techniques (UT) and, Scanning Electron Microscope (SEM) images for the failed pipe, and numerical simulation results for FAC and EC cases in Korea Nuclear Power Plants. It was concluded that the main cause of wall thinning for the downstream pipe of a check valve is FAC caused by water vortex flow due to the internal flow shape of a check valve, the main cause of wall thinning for the downstream elbow of a control valve is FAC caused by a thickness difference with the upstream pipe, and the main cause of wall thinning for the downstream pipe of an orifice is FAC and EC caused by liquid droplets and vortex flow. In order to investigate more cases, additional analyses were performed with the review of a lot of thickness data for inspected pipes. The results showed that pipe wall thinning was also affected by the operating condition of upstream equipment. Management of FAC and EC based on these cases will focus on the downstream piping of abnormal or unusual operated equipment.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.2
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• pp.146-154
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• 2023

In nuclear power plants, ultrasonic test for pipe wall thickness measurement is used during periodic inspections to prevent pipe rupture due to pipe wall thinning. However, when measuring pipe wall thickness using ultrasonic test, a significant amount of measurement error occurs due to the on-site conditions of the nuclear power plant. If the maximum pipe wall thinning rate is decided by the measured pipe wall thickness containing a significant error, the pipe wall thinning rate data have significant uncertainty and systematic overestimation. This study proposes preprocessing of pipe wall thinning measurement data using support vector machine regression algorithm. By using support vector machine, pipe wall thinning measurement data can be smoothened and accordingly uncertainty and systematic overestimation of the estimated pipe wall thinning rate data can be reduced.

• Water for future
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• v.22 no.2
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• pp.213-221
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• 1989

Gradual failure of an earth dam is caused by piping or overtopping. In this gradual failure, a breach will form and grow gradually under the erosive action of the waters. The process involved during an earth dam failure is very dynamic and complicated. The physical model of Fread and mathmatical model of Singh and Scarlatos are verified and compared in this study. Fread's model(BREACH) simulates dam failure well when sufficient data are given, and Singh and Scarlatos' model simulates it appoximately with a few simple data.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.29 no.5 s.236
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• pp.734-742
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• 2005

The present study proposes the integrity evaluation model for a straight pipe with local wall thinning defect, which reflects the characteristics of training shape and loading condition in the Piping of nuclear power plant. For this purpose, a series of finite element analyses are performed under various defect geometries and loading conditions, and real pipe experiment data performed previously is employed. The model includes the effect of thinning length as well as thinning depth and width, and also it considers the combined loading effect between internal pressure and bending moment. The proposed model has been validated using the results of finite element analysis and pipe experiment data. The results indicate that the proposed model provides more reliable predictions of pipe failure than the current existing model, in terms of accuracy, consistency, and conservativeness of results.