• Journal of Energy Engineering
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• v.10 no.4
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• pp.335-341
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• 2001

Pressure pulsations generated by the blade-tongue interaction induce vibration of the piping systems and the structure connected to pumps, resulting in the severe noise and fatigue fracture. Experiments were made on the natural frequencies of liquid columns in piping systems with a single suction, single stage, centrifugal volute pump. Experimental results show that the natural frequencies of the liquid columns in the pump piping systems depend on the dimensions of the pipes and the impeller shapes, and are not affected substantially by the rate of discharge and the rotating speed of the pump.

• Smart Structures and Systems
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• v.19 no.5
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• pp.473-485
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• 2017

Ratcheting behavior of $90^{\circ}$ elbow piping subject to internal pressure 20 MPa and reversed bending 20 kN was investigated using experimental method. The maximum ratcheting strain was found in the circumferential direction of intrados. Ratcheting strain at flanks was also very large. Moreover, the effect of temperature on ratcheting strain of $90^{\circ}$ elbow piping was studied through finite element analysis, and the results were compared with room condition ($25^{\circ}$). The results revealed that ratcheting strain of $90^{\circ}$ elbow piping increased with increasing temperature. Ratcheting boundary of $90^{\circ}$ elbow piping was determined by Chaboche model combined with C-TDF method. The results revealed that there was no relationship between the dimensionless form of ratcheting boundary and temperature.

• Proceedings of the KSME Conference
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• 2000.11a
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• pp.432-437
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• 2000

The purpose of this paper is to determine a two dimensional axisymmetric model through a comparative study between a three dimensional and an axisymmetric finite element analysis of the reactor coolant piping nozzle subject to internal pressure. The finite element analysis results show that the stress adopting the axisymmetric model with the radius of equivalent spherical vessel are well agree with that adopting the three dimensional model. The the radii of equivalent spherical vessel are 3.5 times and 7.3 times of the radius of the reactor coolant piping for the safety injection nozzle and for the residual heat removal nozzle, respectively.

• Journal of the Korean Society of Fisheries and Ocean Technology
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• v.33 no.3
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• pp.194-201
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• 1997

Recently with the rapid development in the industries such as an iron mill and chemical plants, these are enlarged by the use of the piping. This piping was encountered the stress corrosion cracking(SCC) because of stress by water pressure and residual stress of welding etc. under industrial water. In this paper, the behaviour of stress corrosion cracking on the weld zone of steel pipe piping water(SPPW) were investigated according to pre-heat before welding in natural seawater(specific resistance : 25$\Omega$-cm). The main results obtained are as follows :1) The stress corrosion cracking for SPPW and SB 41 is most ready to propagate in heat affected zone of weldment. 2) The SCC sensitivity of SPPW on weldment is more susceptible than that of SB 41. 3) The stress corrosion cracking growth of heat affected zone is delayed by the preheat and dry of base metal and electrode before welding.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.15-22
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• 2011

The chemistry factor and RTPTS margin for domestic reactor pressure vessel materials were analyzed by using the surveillance data which have been obtained from 8 nuclear power plants in Korea. The surveillance data have been used to assess the integrity of the pressure vessel under the pressurized thermal shock (PTS) event. The chemistry factor, which is determined by the Cu and Ni contents of vessel materials, is considered a proper tool to assess the $RT_{PTS}$. The chemistry factors, which were obtained from the surveillance data of domestic reactor pressure vessels, were investigated and compared with those of Regulatory Guide 1.99 in this study. Regressions for ${\Delta}RT_{NDT}$ were performed to expect the chemistry factor as a function of Cu and Ni, and to estimate $RT_{PTS}$ margin. The margin analysis was performed by comparing the regression graphs and standard deviations with those of Regulatory Guide 1.99. The standard deviations calculated by using the domestic surveillance data for base metal and welds are almost same as the standard deviations which are suggested on Regulatory Guide 1.99, Rev.2.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.1
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• pp.18-27
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• 2019

The Advanced Power Reactor 1400 (APR1400) Steam Generator (SG) uses alloy 690 as a tube material and SA-508 Grade 3 Class 1 as a tubesheet material to form tube-to-tubesheet joint through hydraulic expansion process. In this paper, the residual stresses in the SG tube-to-tubesheet contact area was investigated by applying Model-Based System Engineering (MBSE) methodology and the V-model. The use of MBSE transform system description into diagrams which clearly describe the logical interaction between functions hence minimizes the risk of ambiguity. A theoretical and Finite Element Methodology (FEM) was used to assess and compare the residual stresses in the tube-to-tubesheet contact area. Additionally, the axial strength of the tube to tubesheet joint based on the pull-out force against the contact joint force was evaluated and recommended optimum autofrettage pressure to minimize residual stresses in the transition zone given. A single U-tube hole and tubesheet with ligament thickness was taken as a single cylinder and plane strain condition was assumed. An iterative method was used in FEM simulation to find the limit autofrettage pressure at which pull-out force and contact force are of the same magnitude. The joint contact force was estimated to be 20 times more than the pull-out force and the limit autofrettage pressure was estimated to be 141.85MPa.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.1
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• pp.25-31
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• 2009

Steam Generator(SG) tube is an important component of Nuclear Power Plant(NPP), which comprises of the pressure boundary of primary system. The integrity of SG tube has been confirmed by the eddy current test every outage. In Korea, Bobbin probe and MRPC probe have been generally used for the eddy current test. Meanwhile the usage of Array probe has gradually increased in U.S., Japan and other countries. In this study, we investigated the defect detection capability of the Array probe through its preliminary application to SG tube inspection. The Array probe has the equivalent capability in the defect detection and sizing as the conventional methods. Thus it is desirable that the Array probe is generally applied to SG tube inspection in the domestic NPPs.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.1
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• pp.20-24
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• 2009

There is extensive tubing pipe in the nuclear power plant under high temperature and pressure. Erosion and corrosion defects are expected on this tubing pipe due to environmental and mechanical factors. In this study, Guided Ultrasonic Wave technique was applied to detect defects. The technique explores the advantages of the Guided Ultrasonic Wave method that inspects along the wall of the pipe and can travel long distances, providing rapid collection of data. This paper presents a case study of the Guided Ultrasonic Wave testing of 3/8" tubing pipe. This study offers to understand detected signals through correlation between amplitude and depth of 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.11 no.1
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• pp.79-86
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• 2015

The purpose of this study is to evaluate structural integrity of a weldment which is partially screwed and then welded. Two finite element models are constructed and solved: operating temperature and internal pressure are considered in the first simple model, and welding process and normal operating condition including heat-up process are simulated in the second model. Structural integrity assessment criteria are satisfied with both finite element models, therefore the repair weldment finely sustains structural integrity of this assembly and prevents leakage. Stresses are dramatically increased when weld residual stress is considered, but it should be considered as a secondary stress.