• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.109-118
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• 2023

In this paper, the evaluation method of fluidelastic instability (FEI) of newly designed steam generator tubing in pressurized water reactor (PWR) nuclear power plants is discussed. To obtain the parameters for prediction of the critical velocity of FEI for steam generator tubes, experimental research is carried out, and the design parameters are determined. Using CFD numerical simulation, the tube array scale of the model experiment is determined, and the experimental device is designed. In this paper, 7 groups of experiments with void fractions of 0% (water), 10%, 20%, 50%, 75%, 85% and 95% were carried out. The critical damping ration, fundamental frequency and critical velocity of FEI of tubes in flowing water were measured. Through calculation, the total mass and instability constant of the immersed tube are obtained. The critical damping ration measured in the experiment mainly included two-phase damping and viscous damping, which changed with the change in void fraction from 1.56% to 4.34%. This value can be used in the steam generator design described in this paper and is conservative. By introducing the multiplier of frequency and square root of total mass per unit length, it is found that the difference between the experimental results and the calculated results is less than 1%, which proves the rationality and feasibility of the calculation method of frequency and total mass per unit length in engineering design. Through calculation, the instability constant is greater than 4 when the void fraction is less than 75%, less than 4 when the void fraction exceeds 75% and only 3.04 when the void fraction is 95%.

• 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.

• Journal of the Korean Society of Safety
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• v.30 no.4
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• pp.168-173
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• 2015

A steam generator tubing is one of the main pressure boundary of the reactor coolant system in the nuclear power plants. Structural integrity refers to maintaining adequate margins against failure of the tubing. Burst pressure of a tube at tube support plate can be higher than that for a free-span tube because failure behaviors could be interfered from the tube support plate. Alternative repair criteria for out-diameter stress corrosion cracking indications in tubes to the drilled type tube support plate were developed, however, there are very limited information to the eggcrate type tube support plate. This paper discussed reinforcement effect of steam generator tube burst pressure with axial out-diameter stress corrosion cracking within an eggcrate type tube support plate. A series of tube burst tests were performed under the room temperature and it was found out that there is no significant but marginal effects.

• 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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.4
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• pp.19-25
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• 2011

Steam Generator(SG) tube is an important component of Nuclear Power Plant(NPP), which is the pressure boundary between the primary and secondary systems. The integrity of SG tube has been confirmed by the eddy current test every outage. The eddy current technique adopting bobbin probe is currently the primary technique for the steam generator tubing integrity assesment. The bobbin probe is one of the essential components which consist of the whole ECT examination system and provides us a decisive data for the evaluation of tube integrity. Until now, all of the ECT bobbin probes in Korea which is necessary to carry out inspection are imported from overseas. However, KHNP has recently developed the bobbin probe design technology and transferred it to domestic manufacturers to fabricate the probes. This study has been conducted to establish technical requirements applicable to the steam generator tube inspection using the bobbin probes fabricated by the domestic manufactures. The results have been compared with the results obtained by using foreign probe to identify the availability to the steam generator tube inspection. As a result, it is confirmed that the domestic bobbin probe is generally applicable to SG tube inspection in the NPPs.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.344.1-344
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.288-288
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• 1999

• Corrosion Science and Technology
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• v.13 no.3
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• pp.112-119
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• 2014

In this work, it is aimed to develop the fabrication method of axial stress corrosion cracking (SCC) defects having various sizes, on the outer diameter surface of the steam generator (SG) tubings. To control the length of the artificial SCC defect, the specific area of the SG tubing samples was exposed to an acidic solution after a sensitization heat treatment. During the exposure to an acidic solution, a direct current potential drop (DCPD) method was adopted to monitor the crack depth. The size of the SCC defect was first evaluated by an eddy current test (ECT), and then confirmed by a destructive examination. From the comparison, it was found that the actual crack length was well controlled to be similar to the length of the surface exposed to an acidic solution (5, 10, 20 or 30 mm in this work) with small standard deviation. From in-situ monitoring of the crack depth using the DCPD method, it was possible to distinguish a non-through wall crack from a through wall crack, even though the depth of the non-through wall crack was not able to be precisely controlled. The fabrication method established in this work was useful to simulate the SCC defect having similar size and ECT signals as compared to the field cracks in the SG tubings of the operating Korean PWRs.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3225-3234
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• 2022

Alloy 690 tubing has been shown to be highly resistant to primary water stress corrosion cracking (PWSCC). Nevertheless, predicting the failure by PWSCC in Alloy 690 SG tubes is indispensable. In this work, a Bayesian-based statistical approach is proposed to predict the occurrence of failure by PWSCC in Alloy 690 SG tubing. The prior distributions of the model parameters are developed based on the prior knowledge or information regarding the parameters. Since Alloy 690 is a replacement for Alloy 600, the parameter distributions of Alloy 600 tubing are used to gain prior information about the parameters of Alloy 690 tubing. In addition to estimating the model parameters, analysis of tubing reliability is also performed. Since no PWSCC has been observed in Alloy 690 tubing, only right-censored free-failure life of the tubing are available. Apparently the inference is sensitive to the choice of prior distribution when only right-censored data exist. Thus, one must be careful in choosing the prior distributions for the model parameters. It is found that the use of non-informative prior distribution yields unsatisfactory results, and strongly informative prior distribution will greatly influence the inference, especially when it is considerably optimistic relative to the observed data.

• Journal of the Korean Society for Nondestructive Testing
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• v.32 no.5
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• pp.545-550
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• 2012

Steam generator is one of the major components of nuclear power plant and steam generator tubes are the pressure boundary between primary and secondary side, which makes them critical for nuclear safety. As the operating time of nuclear power plant increases, not only damage mechanisms but also scaled deposits on steam generator tubes are known to be problematic causing tube support flow hole blockage and heat fouling. The ability to assess the extent and location of scaled deposits on tubes became essential for management and maintenance of steam generator and eddy current bobbin data can be utilized to measure thickness of scale on tubes. In this paper, tube reference standards with various thickness of scaled deposit has been set up to provide information about the overall deposit condition of steam generator tubes, providing essential tool for steam generator management and maintenance to predict and prevent future damages. Also, methodology to automatically measure scale thickness on tubes has been developed and applied to field data to estimate overall scale amount.