• Journal of Industrial Convergence
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• v.18 no.4
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• pp.15-21
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• 2020

Because there are no isolation valves in condensate system of nuclear power plants, circulating water pump was shutdown for the condenser repair. When circulating water pump was shutdown, power plant output decreased about 45%. These output decreasing can minimize by establishing isolation valves. In this paper, evaluated effect to flow conditions change of condensate system, structural integrity of system, condenser pressure of in case of establish isolation valves to condensate system. Results of the evaluation, the flow rate due to the installation of the isolation valve decreased 0.3% when the valve was fully opened and 4.5% when fully closed. In addition, it was found that the vacuum degree of the condenser decreased with decreasing flow rate, but the integrity of the system was maintained.

• Journal of the Korea Society of Computer and Information
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• v.19 no.7
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• pp.95-102
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• 2014

There are a variety of structures below the surface are buried. In particular, if It is experiencing problems in the city gas pipe or electrical wire, our safety would be greatly jeopardized. Therefore, the underground structures which encounter a variety of pollutants are highly sensitive to corrosion. So if you are not identify the degree of corrosion, it can lead to large accidents such as gas leakage. Until now, person visit directly every underground structure to measure and record manually, but This approach requires a lot of human and material resources and the continuity of management. Therefore, the research to find out the risk factors quickly via the continuous management is needed, and in this paper the structures management systems in the vehicle being moved by combining ICT underground structures for state information wirelessly collects and analyzes system is proposed.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.7
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• pp.729-736
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• 2015

In this study, the effects of specimen size and side-groove on the results of the J-R test for leak-before-break (LBB) evaluation were investigated. A series of J-R tests were conducted at both RT and $316^{\circ}C$, using three different sizes of compact tension (CT) specimens machined from SA508 Gr.1a piping material: 12.7 mm-thick 1T-CT, 25.4 mm-thick 1T-CT, and 25.4 mm-thick 2T-CT with and without side-groove. The results showed that side-grooving reduced the J-R curve for all specimens and the effect of side-grooving was more significant at $316^{\circ}C$ than at RT. As the thickness of the specimens decreased and the width of the specimens increased, the J-R curve slightly decreased at RT but it increased at $316^{\circ}C$. However, the variation in the J-R curve of SA508 Gr.1a with the thickness and width of CT specimen was insignificant.

• Corrosion Science and Technology
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• v.12 no.5
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• pp.227-232
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• 2013

A number of components installed in the secondary system of nuclear power plants are exposed to aging mechanisms such as FAC (Flow-Accelerated Corrosion), Cavitation, Flashing, and LDIE (Liquid Droplet Impingement Erosion). Those aging mechanisms can lead to thinning of the components. In April 2013, one (1) inch small bore piping branched from the main steam line experienced leakage resulting from wall thinning in a 1,000 MWe Korean PWR nuclear power plant. During the normal operation, extracted steam from the main steam line goes to condenser through the small bore piping. The leak occurred in the downstream of an orifice. A control valve with vertical flow path was placed on in front of the orifice. This paper deals with UT (Ultrasonic Test) thickness data, SEM images, and numerical simulation results in order to analyze the extent of damage and the cause of leakage in the small bore piping. As a result, it is concluded that the main cause of the small bore pipe wall thinning is liquid droplet impingement erosion. Moreover, it is observed that the leak occurred at the reattachment point of the vortex flow in the downstream side of the orifice.

• Transactions of the Korean Society of Mechanical Engineers
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• v.19 no.4
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• pp.1051-1061
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• 1995

In order perform leak-before-break design of nuclear piping systems and integrity evaluation of reactor vessels, full stress-strain (.sigma. - .epsilon.) curves and fracture resistance (J-R) curves are required. However it is time-consuming and expensive to obtain J-R curves experimentally. The objective of this paper is to develop two methods for J-R curve prediction. In the first method, elastic-plastic finite element analyses for a series of crack length / specimen width ratio were performed. Accordingly the load versus load line displacement (P .delta.) curve corresponding to the fracture strain is obtained and the J-R curve based on the generalized locus method is obtained. In the second method, the correlation between .sigma.-.epsilon. curves and J-R curves was statistically analyzed and an empirical equation to predict the J-R curve from the .sigma.-.epsilon. test result is proposed. A good correlation between the predicted results based on the proposed methods and the experimental ones is obtained.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.6
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• pp.1099-1106
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• 2002

This paper provides engineering J-integral and crack opening displacement (COD) estimation equations for circumferential through-wall cracked pipes under internal pressure and under combined internal pressure and bending. Based on selected 3-D finite element calculations for the circumferential through-wall cracked pipes under internal pressure using the idealized power law materials, the elastic and plastic influence functions for fully plastic J-integral and COD solutions are found as a function of the normalized crack length and the mean radius-to-thickness ratio. These developed GE/EPRI-type solutions are then re-formulated based on the enhanced reference stress method. Such re-formulation not only provides simpler equations for J-integral and COD estimations, but also can be easily extended to combined internal pressure and bending. The proposed equations are compared with elastic-plastic finite element results using actual stress-strain data, which shows overall excellent agreement.

• 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.

• Journal of the Korean Society of Safety
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• v.32 no.6
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• pp.1-8
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• 2017

P-PIE is a program developed to estimate failure probability of pipes and pressure vessels considering fatigue and stress corrosion crack growth. Using the program, crack growth simulation was performed with an initially existing crack in order to examine the effects of initial crack depth distribution on the leak probability of pipes. In the simulation stress corrosion crack growth was considered and several crack depth distribution models were used. From the results it was found that the initial crack depth distribution gives great effect on the leak probability of pipes. The log-normal distribution proposed by Khaleel and Simonen gives lower leak probability compared other exponential distribution models. The effects of the number and the quality of pre-service and in-service inspections on the leak probability were also examined and it was recognized that the number and the quality of pre-service and in-service inspections are also give great effect on the leak probability. In order to reduce the leak probability of pipes in plants it is very important to improve the quality of inspections. When in-service inspection is performed every 10 years and the quality of inspection is above the very good level, the leak probability shows nearly constant value after the first inspection for an initially existing crack.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.21 no.11
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• pp.1786-1795
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• 1997

In order to perform leak-before-break design of nuclear piping systems and integrity evaluation of reactor vessels, full stress-strain curves and fracture resistance (J-R) curves are required. However it is time-consuming and expensive to obtain J-R curves experimentally. The objective of this paper is to modify two J-R curve prediction methods previously proposed by the authors and to propose an additional J-R curve prediction method for nuclear piping materials. In the first method which is based on the elastic-plastic finite element analysis, a blunting region handling procedure is added to the existing method. In the second method which is based on the empirical equation, a revised general equation is proposed to apply to both carbon steel and stainless steel. Finally, in the third method, both full stress-strain curve and finite element analysis results are used for J-R curve prediction. A good agreement between the predicted results based on the proposed methods and the experimental ones is obtained.

• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 2009.11a
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• pp.78-81
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• 2009

A series of hydraulic tests of a fuel pump are performed using water at a room temperature. The pump is under development for 75-ton class liquid rocket engines of the open-loop gas generator type. According to the test results, the fuel pump satisfies its design requirement and its head and efficiency at the design flowrate are higher than the expected value by the computational analysis. Also, it is found that the pressure at the rear bearing outlet is higher than expected because the inlet of bypass pipe line is narrow. Furthermore, the flowrate of the secondary flow is estimated using the pressure difference of the elbow of the bypass pipe line.