• Journal of the Korea Society for Simulation
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• v.14 no.2
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• pp.35-43
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• 2005

Feedwater heaters of many nuclear power plants have recently experienced wall thinning damage, which will increase as operating time progresses. As it is judged that the wall thinning damages have generated due to local fluid behavior around the impingement baffle installed in downstream of the high pressure turbine extraction steam line to avoid colliding directly with the tubes, numerical analyses using PHOENICS code were performed for two models with original clogged impingement baffle and modified multi-hole impingement baffle. To identify the relation between wall thinning and fluid behavior, the local velocity components in x-, y-, and z-directions based on the numerical analysis for the model with the clogged impingement baffle were compared with the wall thickness data by ultrasonic test. From the comparison of the numerical analysis results and the wall thickness data, the local velocity component only in the y-direction, and not in the x- and z-direction, was analogous to the wall thinning configuration. From the result of the numerical analysis for the modified impingement baffle to mitigate the shell wall thinning, it was identified that the shell wall thinning may be controlled by the reduction of the local velocity in the y-direction.

• Journal of the Korean Society of Safety
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• v.17 no.4
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• pp.52-60
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• 2002

Failure criterion is a parameter to represent the resistance to failure of locally wall thinned pipe, and it depends on material characteristics, defect geometry, applied loading type, and failure mode. Therefore, accurate prediction of integrity of wall thinned pipe requires a failure criterion adequately reflected the characteristics of defect shape and loading in the piping system. In the present study, the finite element analysis was performed and the results were compared with those of pipe experiment to develop a sound criterion for failure of internally wall thinned pipe subjected to combined pressure and bending loads. By comparing the predictions of failure to actual failure load and displacement, an appropriate criterion was investigated. From this investigation, it is concluded that true ultimate stress criterion is the most accurate to predict failure of wall thinned pipe under combined loads, but it is not conservative under some conditions. Engineering ultimate stress estimates the failure load and displacement reasonably for al conditions, although the predictions are less accurate compared with the results predicted by true ultimate stress criterion.

• Corrosion Science and Technology
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• v.17 no.6
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• pp.317-323
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• 2018

Assumptions have always been that wall thinning on the secondary side piping in nuclear power plants is mostly caused by Flow-Accelerated Corrosion (FAC). Recent studies have showed that wall thinning on the secondary side piping is caused by Liquid Droplet Impingement Erosion (LDIE), Solid Particle Erosion (SPE), cavitation, and flashing. To manage those aging mechanisms, several software such as CHECWORKS, COMSY, and BRT-CICERO have been used in nuclear power plants. Korean nuclear power plants have been using the CHECWORKS program since 1996 to date. However, many site engineers have experienced a lot of inconveniences and problems in using the CHECWORKS program. In order to work through the inconveniences and to remedy problems, KEPCO-E&C has developed a "3D-based pipe wall thinning management program (ToSPACE)" based on the experience of over 30 years in relation to the pipe wall thinning management. This study compares the results of FAC and LDIE analysis using both the CHECWORKS and ToSPACE programs with respect to validation of the wall thinning analysis results.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.25 no.7
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• pp.377-385
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• 2013

In this study, the performance of an impeller according to blade length and pitch angle was studied experimentally by building a variable pitch impeller while changing blade length to review the effect of blade length and pitch angle on a fan's performance. The pitch angle was changed in six steps from $20^{\circ}{\sim}45^{\circ}$ at intervals of $5^{\circ}$ while the blade lengths were changed to 90 mm, 100 mm, 110 mm and 120 mm with an identical airfoil shape while carrying out the experiment. The results are summarized as follows : The air flow per static pressure of axial fans increased linearly with increase of pitch angle, but the high static pressure showed a decrease at a pitch angle of $35^{\circ}$. The shaft power increased proportionally to the pitch angle at all blade lengths; the larger the pitch angle, the larger the measured increase of shaft power. This is because the drag at the fan's front increases with the pitch angle. In the axial fans considered in this research, the flow and incre.

• Journal of the Korean Society of Manufacturing Technology Engineers
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• v.22 no.3_1spc
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• pp.518-524
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• 2013

Recently, probabilistic assessments of nuclear power plant components have generated interest in the nuclear industries, either for the efficient inspection and maintenance of older nuclear plants or for improving the safety and cost-effective design of newly constructed nuclear plants. In the present paper, the partial safety factor (PSF) of wall-thinned nuclear piping is evaluated based on a reliability index method, from which the effect of each statistical variable (assessment parameter) on a certain target probability is evaluated. In order to calculate the PSF of a wall-thinned pipe, a limit state function based on the load and resistance factor design (LRFD) concept is first constructed. As for the reliability assessment method, both the advanced first-order second moment (AFOSM) method and second-order reliability method (SORM) are employed to determine the PSF of each probabilistic variable. The present results can be used for developing maintenance strategies considering the priorities of input variables for structural integrity assessments of wall-thinned piping, and this PSF concept can also be applied to the optimal design of the components of newly constructed plants considering the target reliability levels.

• Journal of the Korean Society for Nondestructive Testing
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• v.36 no.6
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• pp.496-503
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• 2016

Leakages at plant structures of power and petrochemistry plants have led to casualties and economic losses. These leakages are caused by fatigue failure of pipelines and their wall thickness. Vibration measurement methods for plant pipelines mainly use acceleration and laser sensors. These sensors are difficult to install and operate and thus lead to an increase in operational cost especially for wide area surveillance. Recently, measurements of leak and vibration displacements using cameras have attracted the interest of many researchers. This method has advantages such as simple installation, long distance monitoring, and wide area surveillance. Therefore, in this paper, we have developed a system that can measure the leakage and vibrational displacement by using a camera. Furthermore, the developed system was verified with experimental data.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.4
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• pp.295-301
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• 2010

Flow accelerated corrosion(FAC) of carbon steel pipes in nuclear power plants has been known as one of the major degradation mechanisms. It could have bad influence on the plant reliability and safety. Also detection of FAC is a significant cost to the nuclear power plant because of the need to remove and replace insulation. Recently, the interest of the guided wave testing(GWT) has grown because it allows long range inspection without removing insulation of the pipe except at the probe position. If GWT can be applied to detection of FAC damages, it will can significantly reduce the cost for the inspection of the pipes. The objective of this study was to determine the capability of GWT to identify location of FAC damages. In this paper, three kinds of techniques were used to measure the amplitude ratio between the first and the second welds at the elbow area of mock-ups that contain real FAC damages. As a result, optimal inspection technique and minimum detectability to detect FAC damages drew a conclusion.

• Corrosion Science and Technology
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• v.12 no.3
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• pp.125-131
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• 2013

The most common pipe wall thinning degradation mechanisms that can occur in the steam and feedwater systems are FAC (Flow Acceleration Corrosion), cavitation, flashing, and LDIE (Liquid Droplet Impingement Erosion). Among those degradation mechanisms, FAC has been investigated by many laboratories and industries. Cavitation and flashing are also protected on the piping design phase. LDIE has mainly investigated in aviation industry and turbine blade manufactures. On the other hand, LDIE has been little studied in NPP (Nuclear Power Plant) industry. This paper presents the development of prediction system for pipe wall thinning caused by LDIE in terms of erosion rate based on air-water ratio and material. Experiment is conducted in 3 cases of air-water ratio 0.79, 1.00, and 1.72 using the three types of the materials of A106B, SS400, and A6061. The main control parameter is the air-water ratio which is defined as the volumetric ratio of water to air (0.79, 1.00, 1.72). The experiments were performed for 15 days, and the surface morphology and hardness of the materials were examined for every 5 days. Since the spraying velocity (v) of liquid droplets and their contact area ($A_c$) on specimens are changed according to the air-water ratio, we analyzed the behavior of LDIE for the materials. Finally, the prediction equations(i.e. erosion rate) for LDIE of the materials were determined in the range of the air-water ratio from 0 to 2%.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.35 no.10
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• pp.1097-1103
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• 2011

Liquid droplet impingement erosion (LDIE) is caused by the impact of high-velocity droplets entrained in steam or air on metal. The degradation caused by the LDIE has been experienced in steam turbine internals and high-velocity airplane components (particularly canopies). Recently, LDIE has also been observed in the pipelines of nuclear plants. LDIE among the pipelines occurs when two-phase steam experiences a high pressure drop (e.g., across an orifice in a line to the condenser). In 2011, a nuclear power plant in Korea experienced a steam leak caused by LDIE in a pipe through which a two-phase fluid was flowing. This paper describes a study on the design change of a pipe affected by LDIE in order to mitigate the damage. The design change has been reviewed in terms of fluid dynamics by using the FLUENT code.

• Plant Journal
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• v.15 no.1
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• pp.25-30
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• 2019

In this study, by varying flow patterns, which is one of the hydraulic factors of FAC, a strategy to reduce pipe wall thinning by mass transfer has been investigated. A similarity between heat transfer and mass transfer was verified via theoretical analysis, and local convective heat transfer coefficients were analyzed using a commercial numerical analysis program. When ribs were installed inside and outside of the internal surface in the straight section of the pipe, the maximum local heat transfer coefficient was shown to decrease substantially by up to 24.9% compared to the basic flow depending on the position and shape of ribs. If a guide vein was inserted in the pipe elbow, the maximum local heat transfer coefficient decreased by up to 26.7% compared to the basic flow depending on the internal surface area of the pipe by the guide vein.