• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2002.05a
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• pp.1049-1054
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• 2002

Fluid Pulsation in pipe usually cause several forces and these forces make mechanical vibration and noise. Protecting pipe from mechanical vibration is very important problem because vibration make pipe damage and break. To analyze pipe, we must formulate both the fluid pulsation force and vibration of pipe. In this paper fluid force from pulsation is modeled by Fluid Dynamics and solved by FEM(finite element method). The discharge pipe is also modeled by the FEM with use of 6 dof beam model. The acceleration of discharge pipe is estimated by the suggested method in this paper. The comparision of estimated results with experimental results show good agreement, which verified the validation of this method

• Proceedings of the Korea Water Resources Association Conference
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• 2003.05a
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• pp.493-496
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• 2003

No Abstract, See Full Text

• Journal of Korea Water Resources Association
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• v.42 no.6
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• pp.445-455
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• 2009

In this paper proportional hazards models for the first through seventh break of 150 mm cast iron pipes in a case study area are established. During the modeling process the assumption of the proportional hazards for covariates on the hazards is examined to include the time-dependent covariate terms in the models. As a result, the pipe material/joint type and the number of customers are modeled as time-dependent for the first failure, and for the second failure only the number of customers is modeled as time-dependent. From the analysis on the baseline hazard functions the failure hazards are found to be generally increasing for the first and second failure, while the hazards of the third break and beyond showed a form of a bath-tub. Furthermore, the changes in the baseline hazard rates according to the time and number of break reflect that the general condition of the pipes is deteriorating. The factors causing pipe break and their effects are analyzed based on the estimated regression coefficients and their hazard ratios, and the constructed models are verified using the deviance residuals of the models.

• Journal of computational fluids engineering
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• v.20 no.3
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• pp.87-93
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• 2015

In this study, a numerical analysis was performed to simulate the thermal-hydraulic response of the secondary side of a steam generator(SG) model equipped with an orifice-type SG outlet flow restrictor to a main steam line break(MSLB) at a pressurized water reactor(PWR) plant. The SG analysis model includes the SG upper steam space and the part of the main steam pipe between the SG outlet and the broken pipe end. By comparing the numerical calculation results for the present SG model to those obtained for a simple SG model having no flow restrictor, the effects of the flow restrictor on the thermal-hydraulic response of SG to the MSLB were investigated.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.6
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• pp.725-730
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• 2010

The leak-before-break behavior and the crack opening displacement were investigated of statically indeterminate piping system and statically determinate piping system after a crack penetration. The reduction in the ultimate strength caused by a crack was relatively small in the statically indeterminate piping system. The leak-before-break in the statically indeterminate piping system had a larger safety margin than that in the statically determinate piping system. The crack opening displacement after crack penetration in a pipe with a nonpenetrating crack was evaluated by using a plastic rotation angle.

• Nuclear Engineering and Technology
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• v.47 no.6
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• pp.712-718
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• 2015

Cast duplex stainless steels are susceptible to thermal aging during long-term service at temperatures ranging from $280^{\circ}C$ to $450^{\circ}C$. To analyze the effect of thermal aging on leak-before-break (LBB) behavior, three-dimensional finite element analysis models were built for circumferentially cracked pipes. Based on the elasticeplastic fracture mechanics theory, the detectable leakage crack length calculation and J-integral stability assessment diagram approach were carried out under different bending moments. The LBB curves and LBB assessment diagrams for unaged and thermally aged pipes were constructed. The results show that the detectable leakage crack length for thermally aged pipes increases with increasing bending moments, whereas the critical crack length decreases. The ligament instability line and critical crack length line for thermally aged pipes move downward and to the left, respectively, and unsafe LBB assessment results will be produced if thermal aging is not considered. If the applied bending moment is increased, the degree of safety decreases in the LBB assessment.

• Korean Journal of Hydrosciences
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• v.8
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• pp.31-40
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• 1997

Flow carrying capacity of water distribution systems is getting reduced by deterioration of pipes in the systems. The objective of this paper is to present a managerial decision-making model for the rehabilitation of water distribution systems with a mininum cost. The decisions made by the model also satisfy the requirements for discharge and pressure at demanding nodes in the systems. Replacement cost, pipe break repair cost, and pumping cost are considered in the economic evaluation of the decision along with the break rate and the interest rate to determine the optimal replacement time for each pipe. Then, the hydraulic integrity of the water distribution system is checked for the decision by a pipe network simulator, KYPIPE, if discharge and pressure requirements are satisfied. In case the system does not satisfy the hydraulic requirements, the decision made for the optimal replacement time is revised until the requirments are satisfied. The model is well applied to an existing water distribution system, the Seoul Metropolitan Water Supply System (1st Phase). The results show that the decisions for the replacement time determined by the economic analysis are accepted as optimal and hydraulic integrity of the system is in good condition.

• Journal of the Korean Society of Safety
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• v.25 no.2
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• pp.7-11
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• 2010

Difficulties associated with full-scale pipe tests are rather obvious. That is, it is not only difficult to perform them but also very expensive and it requires lots of experience. And the process of the fracture test for the pipe specimen is very difficult and complicated. Because the pipe specimen, the test jig and the test equipment are very large and heavy, it requires lots of costs and times. In this study, to easily perform the fracture toughness test for a pipe specimen, load line displacement data was obtained using the image processing method.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.25 no.7
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• pp.1031-1038
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• 2001

In order to verify the analytical methods predicting failure behavior of cracked piping, full-scale pipe tests are crucial in nuclear power plant piping. For this reason, series of international test programs have been conducted. However, full-scale pipe tests require expensive testing equipment and long period of testing time. The objective of this paper is to develop a test system which can economically simulate the full-scale pipe test regarding the integrity evaluation. This system provides the failure behavior of cracked pipe by testing a wide-plate specimen. The system provides the failure behavior of cracked pipe by testing a wide-plate specimen. The system was developed for the integrity evaluation of nuclear piping based on the methodology of hardware-in-the-loop (HiL) simulation. Using this simulator, the piping integrity can be evaluated based on the elastic-plastic behavior of full-scale pipe, and the high cost full-scale pipe test may be replaced with this economical system.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.1
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• pp.19-25
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• 2022

Nuclear Power Plants (NPP) should be designed to have sufficient safety margins and to ensure seismic safety against earthquake that may occur during the plant life time. After the 9.12 Gyeongju earthquake accident, the structural integrity of nuclear power plants due to the beyond design basis earthquake is one of key safety issues. Accordingly, it is necessary to conduct structural integrity evaluations for domestic NPPs under beyond design basis earthquake. In this study, the Level 3 LBB (Leak Before Break) evaluation was performed by considering the beyond design basis earthquake for the surge line of a OPR1000 plant of which design basis earthquake was set to be 0.2g. The beyond design basis earthquake corresponding to peak ground acceleration 0.4g at the maximum stress point of the surge line was considered. It was confirmed that the moment behaviors of the hot leg and pressurized surge nozzle were lower than the maximum allowable loading in moment-rotation curve. It was also confirmed that the LBB margin could be secured by comparing the LBB margin through the Level 2 method. It was judged that the margin was secured by reducing the load generated through the compliance of the pipe.