• Transactions of the Korean Society of Mechanical Engineers B
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• v.34 no.4
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• pp.429-435
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• 2010

In approximately fifty utilities, including KHNP (Korea Hydro & Nuclear Power), CHECWORKS is used as a tool for predicting and managing the wall thinning of carbon steel piping; this wall thinning is caused by flow-accelerated corrosion (FAC). It is known that CHECWORKS is only applicable to predict the wall thinning of piping with large bores. When dealing with small-bore steel piping, FAC engineers measure the thickness of the susceptible area that is selected on the basis of the experience and judgment of the engineer. This paper proposes the application of CHECWORKS for the management of wall thinning of small-bore piping. Four small-bore pipelines of a domestic nuclear power plant were analyzed from the viewpoints of FAC and fluid dynamics by using CHECWORKS and FLUENT code. Depending on the engineer's skill, CHECWORKS can also be used for the management of wall thinning of small-bore piping.

• Nuclear Engineering and Technology
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• v.36 no.1
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• pp.112-120
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• 2004

The purpose of this paper is to analyze the piping failure frequency for the main feedwater system in domestic nuclear power plants(NPPs) for the application to an in-service inspection(ISI), leak before break(LBB) concept, aging management program(AMP), and probabilistic safety analysis(PSA). First, a database was developed for piping failure events in domestic NPPs, and 23 domestic piping failure events were collected. Among the 23 events, 12 locations of wall thinning due to flow accelerated corrosion(FAC) were identified in the main feedwater system in 4 domestic WH 3-loop NPPs. Two types of the piping failure frequency such as the damage frequency and rupture frequency were considered in this study. The damage frequency was calculated from both the plant population data and damage(s) including crack, wall thinning, leak, and/or rupture, while the rupture frequency was estimated by using both the well-known Jeffreys method and a new method considering the degradation due to FAC. The results showed that the damage frequencies based on the number of the base metal piping susceptible to FAC ranged from $1.26{\times}10^{-3}/cr.yr\;to\;3.91{\times}10^{-3}/cr.yr$ for the main feedwater system of domestic WH 3-loop NPPs. The rupture frequencies obtained from the Jeffreys method for the main feedwater system were $1.01{\times}10^{-2}/cr.yr\;and\;4.54{\times}10^{-3}/cr.yr$ for the domestic WH 3-loop NPPs and all the other domestic PWR NPPs respectively, while those from the new method considering the degradation were higher than those from the Jeffreys method by about an order of one.

• Nuclear Engineering and Technology
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• v.42 no.5
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• pp.568-575
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• 2010

Flow Accelerated Corrosion is an active degradation mechanism of CANDU feeder. The tight bend downstream to Gray loc weld connection, close to reactor face, suffers significant wall thinning by FAC. Extensive in-service inspection of feeder wall thinning is very difficult because of the intense radiation field, complex geometry, and space restrictions. Development of a knowledge-based inspection program is important in order to guarantee that adequate wall thickness is maintained throughout the whole life of feeder. Research results and plant experiences are reviewed, and the plant inspection databases from Wolsong Units One to Four are analyzed in order to support developing such a knowledge-based inspection program. The initial thickness before wall thinning is highly non-uniform because of bending during manufacturing stage, and the thinning rate is non-uniform because of the mass transfer coefficient distributed non-uniformly depending on local hydraulics. It is obvious that the knowledge-based feeder inspection program should focus on both fastest thinning locations and thinnest locations. The feeder wall thinning rate is found to be correlated proportionately with QV of each channel. A statistical model is proposed to assess the remaining life of each feeder using the QV correlation and the measured thicknesses. W-1 feeder suffered significant thinning so that the shortest remaining life barely exceeded one year at the end of operation before replacement. W-2 feeder showed far slower thinning than W-1 feeder despite the faster coolant flow. It is believed that slower thinning in W-2 is because of higher chromium content in the carbon steel feeder material. The average Cr content of W-2 feeder is 0.051%, while that value is 0.02% for W-1 feeder. It is to be noted that FAC is reduced substantially even though the Cr content of W-2 feeder is still very low.

• Proceedings of the KSME Conference
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• 2005.11a
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• pp.114.1-114.1
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• 2005

• Proceedings of the Korean Institute of Industrial Safety Conference
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• 2003.05a
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• pp.345-353
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• 2003

탄소강은 가공성과 용접성이 우수하기 때문에 각종 산업설비의 배관재로 많이 사용되고 있으며, 특히 가압중수로형 원전의 1차측 배관과 가압경수로형 원전의 2차측 배관에 주로 사용되고 있다. 그러나 탄소강 배관은 부식에 취약하므로 유동가속부식(FAC, Flow Accelerated Corrosion) 현상에 의한 배관의 두께가 감소하는 감육 손상이 중요하게 대두되고 있는 실정이다. 이러한 감육현상은 다른 어떤 설비보다 안전성의 확보가 강조되고 있는 원전 배관의 경우에 있어서는 특히 중요한 건전성 저해요인으로 인식되고 있다.(중략)

• Corrosion Science and Technology
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• v.11 no.3
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• pp.89-95
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• 2012

U.S. Electric Power Research Institute (EPRI) has developed CHECWORKS program and applied it to power plant piping lines since some lines were ruptured by flow-accelerated corrosion (FAC) in 1978. Nowadays the CHECWORKS program has been used to manage pipe wall thinning phenomena caused by FAC. However, various erosion mechanisms can occur in carbon-steel piping. Most common forms of erosion are cavitation, flashing, liquid droplet impingement erosion (LDIE), and Solid Particle Erosion (SPE). Those erosion mechanisms cause pipe wall thinning, leaking, rupturing, and even result in unplanned shutdowns of utilities. Especially, in two phase condition, LDIE damages a wide scope of plant pipelines. Furthermore, LDIE is the major culprit to cause such as power runback by pipe leaking. This paper describes the methodologies that manage wall thinning and also predict LDIE wall thinning area. For this study, current properties of two-phase condition are investigated and LDIE areas are selected. The areas are checked by B-Scan method to detect the effect of wall thinning phenomena.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.30 no.7 s.250
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• pp.707-714
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• 2006

A numerical analysis study has performed in terms of fluid dynamics to identify the wall thinning generated in the main feedwater isolation valve body of a nuclear power plant. To review the relations between flow characteristics and the wall thinning induced by flow accelerated corrosion (FAC), numerical analysis using FLUENT code and ultrasonic tests (UT) were performed. The local velocities according to the analysis results were compared with the distribution of the measured wall thickness by ultrasonic tests. The comparison results show that the local velocity in the x-direction had no correlation with the wall thinning but the local velocity in the y-direction and turbulence intensity had a great influence on that. These results provide a good match to those of the previous studies - locations colliding vertically against components undergo severe wall thinning. These results may be utilized to the design modification and the wall thinning management for main feedwater isolation valves for preventing the wall thinning degradation.

• Proceedings of the KSME Conference
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• 2007.05a
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• pp.1702-1707
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• 2007

The reduction of pipe-thickness induced by flow accelerated corrosion (FAC) is one of the most serious problems on the maintenance of piping system in nuclear power plants (NNP). If the thickness of a pipe component is reduced below the critical level, it cannot sustain pressure and consequently results in leakage or rupture. For this reason, wall thinning by FAC has been inspected in secondary side piping systems in NPPs. In this research Round Robin Test (RRT) was conducted to verify confidence of wall thinning measurement system in NPP. 12 inspectors from 3 companies participated and 23 specimens were used according to standard practice in RRT. The gage R&R analysis was introduced in regard to repeatability and reproducibility that are affected to measurement system errors. Confidence intervals of thickness measurement system were obtained.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.11 no.1
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• pp.61-68
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• 2015

The function of Feed Water Ring is to provide the flow path from Feedwater Nozzle to inside of SG(steam generator). Significant amounts of general FAC on the outside of the Feed Water Ring are not likely due to the low flow velocities in this area. However, on the interior of the Feed Water Ring, there may be areas of local higher flow velocity which could lead to higher FAC rates. These may include the inlet tee from the Feedwater Nozzle into the Feed Water Ring, the areas where the Feed Water Ring changes diameter, and especially the entrance area to the J-Nozzles. In this paper, the results of root cause analysis of through-wall hole observed at domestic WH 51F SG Feed Water Ring and its effect on the integrity and performance of SG are described. And, the maintenance strategy for WH 51F SG Feed Water Ring and the monitoring strategy for Downcomer Feed Water Ring of CE System 80 SG are presented.

• Bulletin of the Korean Chemical Society
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• v.32 no.12
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• pp.4270-4274
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• 2011

A method is proposed for determining chromium content in the carbon steel pipes of a nuclear power plant (NPP) to evaluate wall thinning caused by flow-accelerated corrosion (FAC). A flat file was used to obtain filings samples. To assess sampling quality, a disk form of SRM 1227 was ground with the flat file, and the amount of Cr in the filings was determined by ICP-AES. The content of chromium in the filings of SRM 1227 was estimated as six times higher than the certified value due to the contamination of chromium in the file. To eliminate chromium contamination from the file, it was coated with WC-12Co using high-velocity oxygen-fuel (HVOF) spraying systems. After obtaining filings samples using the coated file, Cr content in four types of disk-form SRMs was determined by ICP-AES. The recoveries of Cr in the disk-form SRMs were in the range of 95.4-102.6%, with relative standard deviations from 0.43 to 3.0%. The Cr contents in the filings collected from the used outlet headers of the nuclear power plants using the flat file coated were in the range of 0.11-0.19%.