• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.5
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• pp.665-671
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• 2013

In a Korean Generation IV prototype sodium-cooled fast reactor (SFR), various types of high-temperature heat exchangers such as IHX (intermediate heat exchanger), DHX (decay heat exchanger), AHX (air heat exchanger), FHX (finned-tube sodium-to-air heat exchanger), and SG (steam generator) are to be designed and installed. In this study, the high-temperature design and integrity evaluation of the sodium-to-air heat exchanger AHX in the STELLA-1 (sodium integral effect test loop for safety simulation and assessment) test loop already installed at KAERI (Korea Atomic Energy Research Institute) and FHX in the SEFLA (sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger) test loop to be installed at KAERI have been performed. Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two heat exchangers according to the high-temperature design codes, and the integrity of the high-temperature design of the two heat exchangers was confirmed.

• Journal of the Korean Society for Nondestructive Testing
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• v.21 no.5
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• pp.518-525
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• 2001

As steam generator (SG) tubes have aged, new and subtle flaws have appeared. Most of them start growing from outside the tubes. Since signals from outer diameter (OD) defects are very weak compared to those from inner diameter (ID) defects in the conventional eddy current testing due to skin effect, this paper studies the feasibility of using remote field eddy current (RFEC) technique, which has shown equal sensitivity to ID and OD defects in the ferromagnetic pipe inspection. Finite element modeling studies show that the operating frequency needs to be increased up to a few hundred kHz in order for RFEC effects to occur in the nonmagnetic SG tube. The proper distance between exciter and sensor coils is also found to be about 1.5 OD, which is half the distance used in the ferromagnetic pipe inspection. Defect signals obtained by the designed RFEC probe show equal sensitivity to ID and OD defects and the existence of linear relationship between defect depth and phase signal strength. These results tell us that RFEC inspection is feasible even in nonmagnetic steam generator tubes.

• Nuclear Engineering and Technology
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• v.46 no.6
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• pp.869-874
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• 2014

It was discovered in a Korean PWR that an extensive number of very short and shallow cracks in the SG tubes were undetectable by eddy current in-service-inspection because of the masking effect of sludge deposits. Axial stress corrosion cracks at the outside diameter of the steam generator tubes near the line contacts with the tube support plates are the major concern among the six identical Korean nuclear power plants having CE-type steam generators with Alloy 600 high temperature mill annealed tubes, HU3&4 and HB3~6. The tubes in HB3&4 have a less susceptible microstructure so that the onset of ODSCC was substantially delayed compared to HU3&4 whose tubes are most susceptible to ODSCC among the six units. The numbers of cracks detected by the eddy current inspection jumped drastically after the steam generators of HB4 were chemically cleaned. The purpose of the chemical cleaning was to mitigate stress corrosion cracking by removing the heavy sludge deposit, since a corrosive environment is formed in the occluded region under the sludge deposit. SGCC also enhances the detection capability of the eddy current inspection at the same time. Measurement of the size of each crack using the motorized rotating pancake coil probe indicated that the cracks in HB4 were shorter and substantially shallower than the cracks in HU3&4. It is believed that the cracks were shorter and shallower because the microstructure of the tubes in HB4 is less susceptible to ODSCC. It was readily understood from the size distribution of the cracks and the quantitative information available on the probability of detection that most cracks in HB4 had been undetected until the steam generators were chemically cleaned.

• 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.

• Journal of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.24 no.1
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• pp.63-70
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• 2010

In this paper, we study the structure of neural network for predicting defect size of steam generator tube. After extracting the features from the eddy current testing (ECT) signals, multi-layer neural networks are used to predict the defect size. In order to maximize the prediction performance for the defect size, we should carefully choose the structure of neural networks, especially the number of neurons in the hidden layer. In this paper, it is shown that, for the prediction of defect size, the number of neurons in the hidden layer can be efficiently determined by using cross-validation.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1008-1016
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• 2019

In the SG (steam generator) of PWR (pressurized water reactor) for a nuclear plant, hundreds of U-shaped tubes are used for the heat exchanger system. They interact with primary pressurized cooling water flow, generating flow-induced vibration in the secondary flow region. A simplified U-tube model is proposed in this study to apply for experiment and its counterpart computation. Using the commercial code, ANSYS-CFX, we first verified the Moody chart, comparing the straight pipe theory with the results derived from CFD (computational fluid dynamics) analysis. Considering the virtual mass of fluid, we computed the major modes with the low natural frequencies through the comparison with impact hammer test, and then investigated the effect of pump flow in the frequency domain using FFT (fast Fourier transform) analysis of the experimental data. Using two-way fluid-structure interaction module in the CFD code, we studied the influence on mean flow rate to generate the displacement data. A feasible CFD method has been setup in this research that could be applied potentially in the field of nuclear thermal-hydraulics.

• Nuclear Engineering and Technology
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• v.33 no.2
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• pp.145-155
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• 2001

This simulation first allows us to define a transition zone from a bubble to jet mode of the argon out-flow and hereinafter to define a similar area for water-steam leak in the KALIMER SG (Korea Advanced Liquid Metal Reactor Steam Generator) using a water mock-up system, taking into account the KALIMER leak classification and tube bundle design, as a simulation of a real water-steam into sodium leak. in accordance with leak conditions in the KALIMER SG, the transition from bubbling to jetting is studied by means of turbulence regime simulation for argon out-flow through a very small orifice, which has the equivalent diameter of about 0.253 mm. finally the noise generation mechanism is explained from the existing experimental data. We also confirmed the possibility of micro-leak detection from the information of the bubbling mode through simulations and the experiment in this study.

• Corrosion Science and Technology
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• v.10 no.4
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• pp.118-124
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• 2011

Alloy 600 (Ni 75 wt%, Cr 15 wt%, Fe 10 wt%) as a heat exchanger tube of the steam generator (SG) in nuclear power plants (NPP) has been degraded by various corrosion mechanism during the long-term operation. Especially lead (Pb) is known to be one of the most deleterious species in the secondary system causing outer diameter stress corrosion cracking (ODSCC). Oxide formation and breakdown is requisite for SCC initiation and propagation. Therefore it is expected that a property change of the oxide formed on SG tubing materials by lead addition into a solution is closely related to PbSCC. In the present work, the SCC susceptibility was assessed by using a slow strain rate test (SSRT) in caustic solutions with and without lead for Alloy 600 and Alloy 690 (Ni 60 wt%, Cr 30 wt%, Fe 10 wt%) used as an alternative of Alloy 600 because of outstanding superiority to SCC. The results were discussed in view of the oxide property formed on Alloy 600 and Alloy 690. The oxides formed on Alloy 600 and Alloy 690 in aqueous solutions with and without lead were examined by using a transmission electron microscopy (TEM), equipped with an energy dispersive x-ray spectroscopy (EDXS).

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.18 no.2
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• pp.37-42
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• 2022

An influence analysis on multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout is performed to compare the plant responses according to the number of ruptured u-tubes under the assumption of a total of 10 ruptured u-tubes. In all calculation cases, the transient behaviour of major thermal-hydraulic parameters, such as the discharge flow rate through the ruptured u-tubes, reactor header pressure, and void fraction in the fuel channels is found to be overall similar to that of the base case having a single SG with 10 u-tubes ruptured. Additionally, as the conditions of low-flow coolant with high void fraction in the broken loop continued, causing the degradation of decay heat removal, the peak cladding temperature (PCT) would be expected to exceed the limit criteria for ensuring nuclear fuel integrity. However, despite the same total number of ruptured u-tubes, because of the different connection configuration between the SG and pressurizer, a difference is foud in time between the pressurizer low-level signal and reactor header low-pressure signal, affecting the time to trip the reactor and to reach the PCT limit. The present study is expected to provide the technical basis for the accident management strategy for mSGTR transient conditions of CANDU-6 plants.

• Journal of the Korean Society for Nondestructive Testing
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• v.35 no.1
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• pp.61-67
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• 2015

Recently, the initiation of outside diameter stress corrosion cracking (ODSCC) at the tube support plate region of domestic steam generators (SG) with Alloy600 HTMA tubes has been increasing. As a result, SGs with Alloy600 HTMA tubes must be replaced early or are scheduled to be replaced prior to their designed lifetime. ODSCC is one of the biggest threats to the integrity of SG tubes. Therefore, the accurate evaluation of tube integrity to determine ODSCC is needed. Eddy current testing (ECT) is conducted periodically, and its results could be input as parameters for evaluating the integrity of SG tubes. The reliability of an ECT inspection system depends on the performance of the inspection technique and abilty of the analyst. The detection probability and ECT sizing error of degradation are considered to be the performance indices of a nondestructive evaluation (NDE) system. This paper introduces an optimized evaluation method for ECT, as well as the sizing error, including the analyst performance. This study was based on the results of a round robin program in which 10 inspection analysts from 5 different companies participated. The analysis of ECT sizing results was performed using a linear regression model relating the true defect size data to the measured ECT size data.