• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.186-192
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• 2022

Flow accelerated corrosion (FAC) is a type of pipe corrosion in which the pipe thickness decreases depending on the fluid flow conditions. In nuclear power plants, FAC mainly occurs in the carbon steel pipes of a secondary system. However, because the temperature of a secondary system pipe is over 150 ℃, in situ monitoring using a conventional ultrasonic non-destructive testing method is difficult. In our previous study, we developed a waveguide ultrasonic thickness measurement system. In this study, we applied a waveguide ultrasonic thickness measurement system to monitor the thinning of the pipe according to the change in pH. The Korea Atomic Energy Research Institute installed FAC-proof facilities, enabling the monitoring of internal fluid flow conditions, which were fixed for ~1000 h to analyze the effect of the pH. The measurement system operated without failure for ~3000 h and the pipe thickness was found to be reduced by ~10% at pH 9 compared to that at pH 7. The thickness of the pipe was measured using a microscope after the experiment, and the reliability of the system was confirmed with less than 1% error. This technology is expected to also be applicable to the thickness-reduction monitoring of other high-temperature materials.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.255.2-255
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• 2001

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2977-2982
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• 2007

Feedwater heaters of many nuclear power plants have recently experienced severe wall thinning damage, which will increase as operating time progresses. Several nuclear power plants in Korea have experienced wall thinning damage in the area around the impingement baffle - installed downstream of the high pressure turbine extraction steam line - inside number 5A and 5B feedwater heaters. At that point, the extracted steam from the high pressure turbine is two phase fluid at high temperature, high pressure, and high speed. Since it flows in reverse direction after impinging the impingement baffle, the shell wall of the number 5 high pressure feedwater heater may be affected by flow-accelerated corrosion. This paper describes the comparisons between the numerical analysis results using the FLUENT code and the down scale experimental data which effect on disclosing of the shell wall thinning of the high pressure feedwater heaters by porous plate.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.1
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• pp.56-63
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• 2021

The feedwater ring is an assembly in steam generator internal piping, which distributes feedwater into the secondary side of the steam generator. It consists of an assembly of carbon steel piping, pipe fittings and J-nozzles which are inserted into the top of the feedwater ring and welded to the diameter of the ring. The feedwater ring at the attachment region of the J-nozzle may be susceptible to flow accelerated corrosion (FAC) due to flow turbulence which increases local fluid velocities. If a J-nozzle becomes a loose part, it can cause damage to tubing near the tube sheet. In this paper, the structural stress analysis for a wall thinned feedwater ring and integrity evaluations under assumed loading conditions are carried out in compliance with ASME B&PV SecIII, NB-3200.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.2
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• pp.40-49
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• 2016

Flow accelerated corrosion (FAC) of the carbon steel piping has been a significant problem in nuclear power plants. FAC occurs under certain hydrodynamic, environmental, and material conditions, and extensive research into the factors of FAC has been conducted. The basic process of FAC is now relatively well understood; however, a full mechanistic model has not yet been established. Recently, the Korea Atomic Energy Research Institute (KAERI) has built a large experiment loop system for FAC. To produce significant experimental results using this system, the factors affecting on FAC should be analyzed quantitatively, and a model needs to be developed. In this work, a statistical modeling methodology to develop an empirical model is described in detail, and a preliminary model is suggested. Firstly, FAC data were collected from the research literature in Japan and the results of domestic experiments. The flow rate, water temperature, pH at room temperature, and the Cr content are selected as major factors, and nonlinear regression is used to find the best fit of the available data. An iterative procedure between suggesting and evaluating a model is used until an optimum model is obtained. The developed model gives the FAC rate comparable to the measured FAC rate. The developed model is going to be refined using additional laboratory data in the future.

• Corrosion Science and Technology
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• v.12 no.6
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• pp.274-279
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• 2013

Flow accelerated corrosion (FAC) of the carbon steel piping in nuclear power plants (NPPs) has been major issue in nuclear industry. During the FAC, a protective oxide layer on carbon steel dissolves into flowing water leading to a thinning of the oxide layer and accelerating corrosion of base material. As a result, severe failures may occur in the piping and equipment of NPPs. Effect of alloying elements on FAC of pipe materials was studied with rotating cylinder FAC test facility at $150^{\circ}C$ and at flow velocity of 4m/s. The facility is equipped with on line monitoring of pH, conductivity, dissolved oxygen(DO) and temperature. Test solution was the demineralized water, and DO concentration was less than 1 ppb. Surface appearance of A 106 Gr. B which is used widely in secondary pipe in NPPs showed orange peel appearance, typical appearance of FAC. The materials with Cr content higher than 0.17wt.% showed pit. The pit is thought to early degradation mode of FAC. The corrosion product within the pit was enriched with Cr, Mo, Cu, Ni and S. But S was not detected in SA336 F22V with 2.25wt.% Cr. The enrichment of Cr and Mo seemed to be related with low, solubility of Cr and Mo compared to Fe. Measured FAC rate was compared with Ducreaux's relationship and showed slightly lower FAC rate than Ducreaux's relationship.

• Nuclear Engineering and Technology
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• v.30 no.2
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• pp.148-163
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• 1998

Flow Induced Material Degradation (FIMD) is reviewed focusing on Flow Accelerated Corrosion (FAC) models. Several examples of FAC related incidents are described, which include nuclear and fossile power plants. Lastly, mitigation techniques such as inspection, material selection, water chemistry, temperature, and hydrodynamic factor are discussed.

• 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.

• Journal of Energy Engineering
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• v.11 no.1
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• pp.18-25
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• 2002

One of the most serious concern in nuclear power plant piping maintenance is thickness reduction due to flow-accelerated corrosion (FAC). Since the FAC occurs under specific conditions of pH, dissolved oxygen, temperature, flow velocity, steam quality of the fluid and materials and geometry of the piping, a systematic approach is required for managing the FAC problem. In this study, construction of a secondary piping database, analyzing the FAC rate using the database and predicting the residual life was performed for a domestic CANDU nuclear power plant. Also FAC mechanism and factors affecting FAC were reviewed. By showing a case study on analysis for a pipe line between a separator and a flash tank, a procedure for managing FAC problem is suggested. The procedure proposed in this paper can be widely applied to the secondary piping of other domestic nuclear polder plants.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.2
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• pp.51-60
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• 2015

Local wall thinning and integrity degradation caused by several mechanisms, such as flow accelerated corrosion, cavitation, flashing and/or liquid droplet impingement, is a main concern in secondary steam cycle piping system of nuclear power plants in terms of safety and operability. Thinned pipe management program (TPMP) has being developed and optimized to reduce the possibility of unplanned shutdown and/or power reduction due to pipe failure caused by wall thinning. In this paper, newest technologies, standards and regulations related to the integrity assessment, repair and replacement of thinned pipe component are reviewed. And technical improvement items in TPMP to secure the reliability and effectiveness are also presented.