• Korean Journal of Materials Research
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• v.8 no.8
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• pp.726-732
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• 1998

가압 경수로형 원전에 사용되는 Alloy 600 증기발생기 전열관재료의 입계응력부식균열 거동에 미치는 냉간변형의 영향을 1차 냉각수 모사조건에서 정속인장시험방법으로 조사하였다. 인장 냉간변형은 응력부식균열을 크게 가속화 시키지는 않았으며 변형량이 25%이상인 경우에는 응력부식균열이 발생하지 않았다. 이 현상은 냉간 변형량 및 형태에 따른 미소변형 및 응력의 불균질성에 영향을 받는 것으로 사려되며 응력의 크기는 직접적인 영향을 주지 않는 것으로 보인다. 국부적인 큰 응력구배가 존재하는 경우 균열의생성 및 성장이 현저히 가속화되었는데 이는 원전 1차측 응력부식균열 기구가 응력구배에 의존하는 과정과 연관되어 있다는 증거이다. Hump 시편을 이용한 정속인장시험방법은 짧은 실험기간내에 원전 1차측 응력부식균열 특성을 평가할 수 있는 방법이었다.

• Journal of the Korean Society for Nondestructive Testing
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• v.29 no.2
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• pp.137-144
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• 2009

In this paper, the signal analysis of eddy current array probe was performed to analyze the electromagnetic characteristics with the variation of FBH(flat bottomed hole) defects size on steam generator tube in NPP(nuclear power plants) using the electromagnetic finite element method. To obtain the electromagnetic characteristic of probes, the governing equation was derived from Maxwell's equations, and the individual problem was analyzed by using the 3-dimensional finite element method. For the simulation FBH defects were used. The depth of FBH defects were 20%, 40%, 60%, 80% and 100% of steam generator(SG) tube thickness, and it was assumed that the defects were located on the tube outside. And the operation frequencies of 100 kHz, 300 kHz and 400 kHz were used. Material of specimen was Inconel 600 which is usually used for SG tubes in NPP. The signal difference could be observed according to the size variation of depth of FBH defects and operation frequencies. The results in this paper can be helpful when the ECT(eddy current testing) signals from EC array probe are evaluated and analyzed.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.20 no.9
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• pp.2894-2900
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• 1996

The stream generator tubes represent an integral part of a major barrier against the fission product release to the environment. So, the rupture of these tubes could permit flow of reactor coolant into the secondary system and injure the safety of reactor coolant system. Therefore, if the crack was detected during In-Service Inspection of tubes the cracked tube should be evaluated by the pulgging criteria and plugged or not. In this study, the fracture mechanics evaluation is carried out on the thru-wall axial crack due to Primary Water Stress Corrosion Cracking in the roll transition aone of steam generator tube to help the assurence the integrity of tubes and estabilish the plugging criteria. Due to the Inconel which is used as tube material is more ductile than others, the plastic instability repture theory was used to calculate the critical and allowable crack length. Based on Leak Before Break concept the leak rate for the critical crack length and the allowable leak rate are compared and the safety of tubes was given.

• Journal of Ocean Engineering and Technology
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• v.21 no.2 s.75
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• pp.50-59
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• 2007

Steam generator tubes are degraded from wear, stress corrosion cracking, rupture and fatigue and so on. Therefore, the failure assessment of steam generator tube is very important for the integrity of energy plants. In the steam generator tubes, sometimes, the local wall thinning may result from severe degradations such as erosion-corrosion damage and wear due to vibration. In this paper, the elasto-plastic analysis was performed by FE code ANSYS on steam generator tubes with wall thinning. Also, the four-point bending tests were performed on the wall thinned specimens, and then it was compared with the analysis results. We evaluated the failure mode, fracture strength and fracture behavior from the experiment and FE analysis. Also, it was possible to predict the crack initiation point by estimating true fracture ductility under multi-axial stress conditions at the center of the thinned area from FE analysis.

• Journal of the Korean Society for Nondestructive Testing
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• v.28 no.4
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• pp.339-345
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• 2008

The heat exchanger tube in nuclear power plants is mainly fabricated from nonferromagnetic material such as a copper, titanium, and inconel alloy, but the moisture separator & reheater tube in the turbine system is fabricated from ferromagnetic material such as a carbon steel or ferrite stainless steel which has a good mechanical properties in harsh environments of high pressure and temperature. Especially, the moisture separator & reheater tubes, which use steam as a heat transfer media, typically employ a tubing with integral fins to furnish higher heat transfer rates. The ferromagnetic tube typically shows superior properties in high pressure and temperature environments than a nonferromagnetic material, but can make a trouble during the normal operation of power plants because the ferrous tube has service-induced damage forms including a steam cutting, erosion, mechanical wear, stress corrosion cracking, etc. Therefore, nondestructive examination is periodically performed to evaluate the tube integrity. Now, the remote field testing(RFT) technique is one of the solution for examination of ferromagnetic tube because the conventional eddy current technique typically can not be applied to ferromagnetic tube such as a ferrite stainless steel due to the high electrical permeability of ferrous tube. In this study, we have designed RFT probes, calibration standards, artificial flaw specimen, and probe pusher-puller necessary for field application, and have successfully carry out RFT examination of the moisture separator & reheater tube of nuclear power plants.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.82-87
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• 1996

To find the effects of pressure on critical heat flux (CHF) for the conditions of low pressures (especially up to 10 bar) and low mass flux (~300 kg/$m^2$s), a series of experiments have been accomplished by using uniformly heated Inconel-625 tube. The experimental ranges are as follows: pressure (from 1.2 to 8 bar). mass velocities (from 100 to 250 kg/$m^2$s) and the inlet subcooling ($\Delta$h$_{i}$ ＝ 350 kJ/kg). According to the experimental data, it is found that the CHF is nearly independent of the pressure and increases with mass flux. From the results of the CHF correlation assessment for this experimental data, we could find somewhat different tendency of CHF behavior from every other CHF prediction correlation and table.ation and table.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.05a
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• pp.961-966
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• 1995

국내 상업적으로 운용증인 경수로 원자력 발전소중, 중기발생기의 건전성 유지를 위하여 보수 유지에 많은 비용을 소비하고 있다 특히 중기발생기 전열관으로 사용되는 inconel 600 재질에 많은 문제점이 발생되었다. 전열관 파손에 대한 보수 및 방지기술은 plugging, sleeving, shot penning, Ni-plating 등이 있다. 특히 최근에 개발된 고출력 Nd:YAG 레이저를 이용한 sleeving 보수 기술이 개발되었다. Nd:YAG 레이저를 이용한 보수 방식은 미국의 WH 및 일본의 MHI 등이 선정하여 실용화 단계에 있으며, 이는 광섬유로 전송이 가능한 Nd:YAG 레이저를 이용하여 원격으로 가공할 수 있는 기술이다. 현재 한국 원자력 연구소에서는 전열관 레이저 보수 용접에 대한 개념을 확립하고 장비 및 기구를 개발하였으며, 고리 1 호기 전열관규격에(7/8") 3/4" sleeve tube를 삽입하여 약 50 m 떨어진 곳으로부터 원격 레이저 용접을 실험실적 규모로 실증 하였다.모로 실증 하였다.

• Proceedings of the KIEE Conference
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• 1997.07a
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• pp.83-85
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• 1997

In this paper a numerical analysis method for eddy current testing (ECT) with axisymmetric crack is studied. The finite element method(FEM) is used for electromagnetic solution. In this paper the outer diameter crack of INCONEL 600 tube is modelled and the impedance signal is obtained using the differential probe. The characteristics of the crack depth variation in the signal are analyzed.

• Proceedings of the KWS Conference
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• 2010.05a
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• pp.43-46
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• 2010

As the nuclear power plant has been constructed continuously for several decades in Korea, the welding technology for components manufacturing and installation has been improved largely. Standardization for weld test and qualification was also established systematically according to the concerned code. The welding for the main components requires the high reliability to keep the constant quality level, which means the repeatability of weld quality. Therefore the weld process qualified by thorough test and evaluation is able to be applied for manufacturing. Narrow gap SAW and GTAW process are usually applied for girth seam welding of pressure vessel like Reactor vessel, steam generator, and etc. For the surface cladding with stainless steel and Inconel material, strip welding process is mainly used. Inside cladding of nozzles is additionally applied with Hot wire GTAW and semi-auto welding process. Especially the weld joint having elliptical weld line on curved surface needs a specialized weld system which is automatically rotating with adjusting position of the head torch. The small sized pipe, tube, and internal parts of reactor vessel requests precise weld processes like an automatic GTAW and electron beam welding. Welding of dissimilar materials including Inconel690 material has high possibility of weld defects like a lack of fusion, various types of crack. To avoid these kinds of problem, optimum weld parameters and sequence should be set up through the many tests. As the life extension of nuclear power plant is general trend, weld technologies having higher reliability is required gradually. More development of specialized welding systems, weld part analysis and evaluation, and life prediction for main components should be taken into a consideration extensively.

• Journal of the Korean Society for Precision Engineering
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• v.26 no.2
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• pp.101-108
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• 2009

A fretting wear test rig for dry ambient condition, which employs a piezoelectric actuator, has been developed. It is driven and loaded in a very simple manner with acceptable experimental accuracy. By using the rig, Inconel 690 tube has been tested under the normal load of 10 and 15N with sliding amplitude of less than $100{\mu}m$ during $10^6$cycles. The wear resistance of the material has been characterized in terms of the wear coefficient based on the work rate model. SEM micrographs show the complex structures of the scars, which consist of risen peaks, plate-type thin layers and locally exposed bare surfaces. The cracks spread over the layers give clue to the fretting wear mechanism of the material.