• Proceedings of the KIEE Conference
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• 2009.07a
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• pp.790_791
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• 2009

본 논문에서는 전자기 유한요소 해석을 통하여 원전 증기 발생기(SG, Steam Generator) 세관의 결함 변화에 따른 배열와전류프로브의 와전류탐상 특성을 해석하였다. 프로브의 전자기적 특성을 위해 맥스웰 방정식을 이용하여 지배방정식을 유도하였고, 이를 3차원 전자기 유한요소법을 이용하여 문제를 해석하였다. 해석을 위한 선정한 결함은 프로브의 특성파악을 위한 표준시험편과 원전 SG세관에 발생 가능한 결함인 Pitting, SCC, Wear, Multi SCC 결함을 선정하였다. 해석 대상으로는 원자력발전소 증기발생기 세관으로 사용되고 있는 Inconel 600 도체관을 사용하였다. 본 논문으로 통하여 결함의 형상, 크기, 시험주파수의 변화에 따른 탐상신호의 변화를 확인할 수 있었다. 본 논문의 결과는 배열와전류프로브의 와전류탐상 신호 평가시 도움이 될 것이다.

• 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.

• 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.

• Nuclear Engineering and Technology
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• v.36 no.2
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• pp.127-141
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• 2004

Various types of integrated steam generators, which integrate IHTS and a steam generator into a single unit of equipment for an LMR, were analyzed using an analytic solution with some simplification. The analysis showed that the undesirable reversed heat transfer, of which occurrence was previously observed only in an integrated single-region bundle type, can also occur in an integrated double-region bundle type. The mechanism of the reversed heat transfer occurrence in the double-region type is explained and it is shown the mechanism in the double-region type is completely different from that in the single-region type. Based on this finding, a method for preventing the aforementioned heat transfer is suggested. The performance of the four types of the integrated steam generators is assessed. For this assessment, a SG is actually designed for each type and the optimization in the geometric parameters and flow rate are optimized.

• 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.

• Journal of the Korean Society for Nondestructive Testing
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• v.27 no.5
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• pp.383-392
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• 2007

In this paper, we study the prediction of depth and width of a defect in steam generator tube in nuclear power plant using neural network. To this end, we first generate eddy current testing (ECT) signals for 4 defect patterns of SG tube: I-In type, I-Out type, V-In type, and V-Out type. In particular, we generate 400 ECT signals for various widths and depths for each defect type by the numerical analysis program based on finite element modeling. From those generated ECT signals, we extract new feature vectors for the prediction of defect size, which include the angle between the two points where the maximum impedance and half the maximum impedance are achieved. Using the extracted feature vector, multi-layer perceptron with one hidden layer is used to predict the size of defects. Through the computer simulation study, it is shown that the proposed method achieves decent prediction performance in terms of maximum error and mean absolute percentage error (MAPE).

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.1
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• pp.27-35
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• 2009

In this paper, we study the classification of steam generator tube defects using an improved feature extraction. We consider 4 axisymmetric defect patterns of tube: I-In type, I-Out type, V-In type, and V-Out type. Through numerical analysis program based on finite element modeling, 400 ECT signals are generated by varying width and depth of each defect type. From those generated ECT signals, we propose new feature vectors that include an angle between the two points where the Maximum impedance and half the Maximum impedance, and angles between Maximum impedance point and 10%, 20%, 30%, 40% of Maximum impedance points. Also, multi-layer perceptron with one hidden layer is used to classify the defect patterns. Through the computer simulation study, it is shown that the proposed method achieves an improved defect classification performance in terms of Maximum Error and mean square Error.

• 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.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.