• Journal of the Korean Society for Nondestructive Testing
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• v.33 no.6
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• pp.519-523
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• 2013

A steam generator (SG) tube is an important component of a nuclear power plant (NPP). It works as a pressure boundary between the primary and secondary systems. The integrity of a SG tube can be assessed by an eddy current test every outage. The eddy current technique(adopting a bobbin probe) is currently the main technique used to assess the integrity of the tubing of a steam generator. An eddy current signal analyst for steam generator tubes continuously analyzes data over a given period of time. However, there are possibilities that the analyst conducting the test may get tired and cause mistakes, such as: missing indications or not being able to separate a true defect signal from one that is more complicated. This error could lead to confusion and an improper interpretation of the signal analysis. In order to avoid these possibilities, many countries of opted for automated analyses. Axial ODSCC (outside diameter stress corrosion cracking) defects on the tubes of OPR1000 steam generators have been found on the tube that are in contract with tube support plates. In this study, automated analysis software called CDS (computer data screening) made by Zetec was used. This paper will discuss the results of introducing an automated analysis system for an axial ODSCC on the tubes of an OPR1000 steam generator.

• Proceedings of the Korean Powder Metallurgy Institute Conference
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• 2006.09a
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• pp.585-586
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• 2006

Several chemicals were studied to suppress the damage due to stress corrosion cracking (SCC) of steam generator (SG) tubes in nuclear power plants. The effects on the SCC of the compounds, $TiO_2$, TyzorLA and $CeB_6$, were tested for several types of SG tubing materials. The test with the addition of $TiO_2$ and $CeB_6$ showed an effect in decreasing the SCC for the SG tubing material. However, $CeB_6$ caused some more SCC for Alloy 800. The penetration property into a crevice of the inhibitors was investigated by using Alloy 600 specimens with different gap.

• Corrosion Science and Technology
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• v.3 no.1
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• pp.30-33
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• 2004

It has been reported that some events of a rupture of seam generator tube have occurred in nuclear power plants around the world. Main causes of the leakage are from various types of corrosion in the steam generator(SG) tubings. Primary water stress corrosion cracking(PWSCC) of steam generator tubings have occurred in many tubes in Korean plant, and they were repaired using sleeves or plugs, In order to develop proper repair criteria, it is necessary to ascertain the leak behavior of the tubings. A high pressure leak and burst testing system was manufactured. Various types of Electro Discharged Machined (EDM) notches were developed on the SG tubes. Leak rate and burst pressure were measured on the tubes at room temperature. Burst pressure of the part through wall defected tubes depends on the defect depth, Water flow rates after the burst were independent of the t1aw types; tubes having 20 to 60 mm long EDM notches showed similar flow rates regardless of the defect depth. A fast pressurization rate gave the tube a lower burst pressure than the case of a slow pressurization.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.889-896
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• 2020

The vertically inverted U-tube steam generator (UTSG) is widely used in the pressurized water reactor (PWR). The reverse flow behavior generally exists in some U-tubes of a steam generator (SG) under both single- and two-phase natural circulations (NCs). The behavior increases the flow resistance in the primary loop and reduces the heat transfer in the SG. As a consequence, the NC ability as well as the inherent safety of nuclear reactors is faced with severe challenges. The theoretical models for calculating single- and two-phase flow pressure drops in U-tubes are developed and validated in this paper. The two-phase reverse flow characteristics in two types of SGs are investigated base on the theoretical models, and the effects of the U-tube height, bending radius, inlet steam quality and primary side pressure on the behavior are analyzed. The conclusions may provide some promising references for SG optimization to reduce the disadvantageous behavior. It is also of significance to improve the NC ability and ensure the PWR safety during some accidents.

• Corrosion Science and Technology
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• v.8 no.5
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• pp.188-192
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• 2009

A forced outage due to a steam generator tube leak in a Korean nuclear power plant was reported.1) Primary water stress corrosion cracking has occurred in many tubes in the plant, and they were repaired using sleeves or plugs. In order to develop proper repair criteria, it is necessary to understand the leak behavior of the tubes containing stress corrosion cracks. Stress corrosion cracks were developed in 0.1 M sodium tetrathionate solution at room temperature. Steam generator(SG) tubes with short cracks were successfully fabricated with a restricted solution contact method. The leak rates of the degraded tubes were measured at room temperature. Some tubes with 100 % through wall cracks showed an increase of leak rate with time at a constant pressure.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.58 no.12
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• pp.2532-2537
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• 2009

For defect characterization in steam generator tubes in nuclear power plant, artificial neural network has been extensively used to classify defect types. In this paper, we study the effectiveness of Bagging for improving the performance of neural network for the classification of tube defects. Bagging is a method that combines outputs of many neural networks that were trained separately with different training data set. By varying the number of neurons in the hidden layer, we carry out computer simulations in order to compare the classification performance of bagging neural network and single neural network. From the experiments, we found that the performance of bagging neural network is superior to the average performance of single neural network in most cases.

• Nuclear Engineering and Technology
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• v.41 no.10
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• pp.1315-1322
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• 2009

The present article is concerned with the application of an electrochemical noise (EN) monitoring technique to analyze the initiation and propagation of Pb-assisted stress corrosion cracking (SCC) of an Alloy 600 material in a simulated environment of a steam generator (SG) sludge pile at high temperatures. A typical increase of electrochemical current noise (ECN) and electrochemical potential noise (EPN) was frequently recorded from the EN measurement in a caustic solution with such impurities as PbO and CuO, indicating that there are localized corrosion events occurring. With the aid of microscopic and spectral analyses, the EN data involving information on such stochastic processes as uniform corrosion and the initiation and propagation of SCC, were analyzed based on a stochastic theory.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2001.06a
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• pp.115-121
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• 2001

The wear test has been performed to evaluate the wear mechanism of steam generator (SG) tube materials against ferritic stainless steel in water environment. The wear rates of SG tube materials depend on the change of mechanical properties between contact surfaces during wear test. From the subsurface hardness test, Inconel 690 is more work-hardened than Inconel 600 even though these materials have similar hardness values before the wear test. Main cause is due to the difference of stacking fault energy with the chromium content. In water environment, wear mechanism is closely related with the continuous formation and fracture of deformation layers at the contact surfaces.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.15 no.2
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• pp.64-70
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• 2019

In recent years, attention has been paid to the integrity of steam generator (SG) tubes due to severe accident and beyond design basis accident conditions. In these transient conditions, steam generator tubes may be damaged by high temperature and pressure, which might result in a risk of fission products being released to the environment due to the failure. Alloy 690 which has increased the Cr content has been replaced for the SG tube due to its high corrosion resistance against stress corrosion cracking (SCC). However, there is lack of research on the high temperature creep rupture and life prediction model of Alloy 690. In this study, creep test was performed to estimate the high temperature creep rupture life of Alloy 690 using tube specimens. Based on manufacturer's creep data and creep test results performed in this study, creep life prediction was carried out using the Larson-Miller (LM) Parameter, Orr-Sherby-Dorn (OSD) parameter, Manson-Haford (MH) parameter, and Wilshire's approach. And a hyperbolic sine (sinh) function to determine master curves in LM, OSD and MH parameter methods was used for improving the creep life estimation of Alloy 690 material.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.1955-1962
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• 2020

In a sudden shutdown of primary pump or coolant loss accident in a marine nuclear power plant, the primary flow decreases rapidly in a transition process from forced circulation (FC) to natural circulation (NC), and the lower flow enters the steam generator (SG) causing reverse flow in the U-tube. This can significantly compromise the safety of nuclear power plants. Based on the marine natural circulation steam generator (NCSG), an experimental loop is constructed to study the characteristics of reverse flow under middle-temperature and middle-pressure conditions. The transition from FC to NC is simulated experimentally, and the characteristics of SG reverse flow are studied. On this basis, the experimental loop is numerically modeled using RELAP5/MOD3.3 code for system analysis, and the accuracy of the model is verified according to the experimental data. The influence of the flow variation rate on the reverse flow phenomenon and flow distribution is investigated. The experimental and numerical results show that in comparison with the case of adjusting the mass flow discontinuously, the number of reverse flow tubes increases significantly during the transition from FC to NC, and the reverse flow has a more severe impact on the operating characteristics of the SG. With the increase of flow variation rate, the reverse flow is less likely to occur. The mass flow in the reverse flow U-tubes increases at first and then decreases. When the system is approximately stable, the reverse flow is slightly lower than obverse flow in the same U-tube, while the flow in the obverse flow U-tube increases.