• 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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.30 no.12 s.255
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• pp.1618-1625
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• 2006

During the last two decades, several guidelines have been developed and used for assessing the integrity of a defective steam generator (SG) tube that is generally caused by stress corrosion cracking or wall-thinning phenomenon. However, as some of SG tubes are also failed due to fretting and so on, alternative failure estimation schemes are required for relevant defects. In this paper, parametric three-dimensional finite element (FE) analyses are carried out under internal pressure condition to simulate the failure behavior of SG tubes with different defect configurations; elliptical wear, tapered and flat wear type defects. Maximum pressures based on material strengths are obtained from more than a hundred FE results to predict the failure of SG tube. After investigating the effect of key parameters such as defect depth, defect length and wrap angle, simplified failure estimation equations are proposed in relation to the equivalent stress at the deepest point in wear region. Comparison of failure pressures predicted by the proposed estimation scheme with corresponding burst test data showed a good agreement.

• 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.

• 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.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.12 no.4
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• pp.261-271
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• 2002

The fluid-elastic instability analysis of the U-tube bundle inside the steam generator is very important not only for detailed design stage of the SG but also for the change of operating condition of the nuclear powerplant. However the calculation procedure for the fluid-elastic instability was so complicated that the consolidated computer program has not been developed until now. In this study, the numerical calculation procedure and the computer program to obtain the stability ratio were developed. The thermal-hydraulic data in the region of secondary side of steam generator was obtained from executing the ATHOS3 code. The distribution of the fluid density can be calculated by using the void fraction, enthalpy, and operating pressure. The effective mass distribution along the U-tube was required to calculate natural frequency and dynamic mode shape using the ANSYS ver. 5.6 code. Finally, stability ratios for selected tubes of the CE type steam generator were computed. We considered the YGN 3.4 nuclear powerplant as the model plant, and stability ratios were investigated at the flow exit region of the U-tube. From our results, stability ratios at the central and the outside region of the tube bundle are much higher than those of other region.

• Nuclear Engineering and Technology
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• v.35 no.6
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• pp.566-580
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• 2003

To resolve the concern of the SWR possibility in LMR and improve the economic feature of LMR, relative performance of various SG designs using a double tube bundle configuration is evaluated and a new SG design concept is proposed. The new steam generator design houses two tube bundles that are functionally different and its tube bundle region is radially divided into two. It prevents the occurrence of sodium water reaction while sodium is still used as the coolant for the primary heat transport system. The feasibility of the SG with a double tube bundle for actual use in an LMR plant is evaluated by setting up the skeleton of the NSSS for various possible configurations of the SG tube bundles. The evaluation revealed the relative advantages and disadvantages of the configurations and the new SG design concept performs good and can be actually used in an LMR plant.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2002.10b
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• pp.203-204
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• 2002

The wear behaviour of steam generator (SG) tubes (Inconel 600 and 690) against support materials (405 and 409 ferritic stainless steels) has been experimentally studied in room temperature water using reciprocating wear apparatus with tube-an-plate configuration. The results showed that the wear rate of Inconel 690 was lower than that of lnconel 600 with increasing normal loads and sliding amplitudes. Also, plastic deformation layers appear below the surface of both SG tubes, which have a specific thickness and are small compared with their grain size. This means that wear rate of SG tubes in water condition is closely related to the formation and fracture of plastic deformation layers.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.26 no.2
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• pp.203-209
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• 2016

According to the wear detection history for the steam generator tubes in the nuclear power plant, the outer tubes inside the steam generator have more problems on the flow-induced vibration than inner tubes. Many researchers and engineers have used a specified added mass coefficient for a given tube array during the design stage of the steam generator even though the coefficient is not constant for entire tube in cylindrical shell. The aim of this study is to find out the distribution of added mass coefficients for each tube along the radial location. When numbers of tubes inside a cylindrical shell are increased, values of added mass coefficients are also increased. Added mass coefficients at outer tubes are less than those of inner tubes and they are decreased with increasing the gap between the outermost tube and the cylindrical shell. It also turns out when the gap between the outermost tube and the cylindrical shell approaches infinite value, the added mass coefficient converges to an asymptotic value of given tube array in a free fluid field.

• Journal of the Korean Society for Nondestructive Testing
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• v.30 no.4
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• pp.302-310
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• 2010

In this paper, we studied Bagging neural network for predicting defect size of steam generator(SG) tube in nuclear power plant. Bagging is a method for creating an ensemble of estimator based on bootstrap sampling. For predicting defect size of SG tube, we first generated eddy current testing signals for 4 defect patterns of SG tube with various widths and depths. Then, we constructed single neural network(SNN) and Bagging neural network(BNN) to estimate width and depth of each defect. The estimation performance of SNN and BNN were measured by means of peak error. According to our experiment result, average peak error of SNN and BNN for estimating defect depth were 0.117 and 0.089mm, respectively. Also, in the case of estimating defect width, average peak error of SNN and BNN were 0.494 and 0.306mm, respectively. This shows that the estimation performance of BNN is superior to that of SNN.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.7-14
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• 2014

There has been extensive experience associated with the operation of SGs wherein it was believed, based on NDE, that throughwall tube indications were present within the tubesheet. The installation of the SG tubes usually involves the development of a short interference fit, referred to as the tack expansion, at the bottom of the tubesheet. The tack expansion was usually effected by a hard rolling process and thereafter, in most instance, by the expansion of a urethane plug inserted into the tube end and compressed in the axial direction. The rolling process by its very nature is considered to be intensive with regard to metalworking at the inside surface of the tube and would be expected to lead to higher residual surface stresses. Alternate repair criteria(ARC) in the tack expansion area have been developed and applied to nuclear power plants in USA, however domestic nuclear power plants have not applied ARC for tubes in tubeheet area yet. In consideration of the degradation characteristics of tubes in the Steam Generator tubesheet, this paper suggests ARC application for tubes in the steam generator tubesheet of the domestic nuclear power plants in order to assure life time of the steam generator as well as nuclear power plants.