• Journal of the Korean Society for Nondestructive Testing
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• v.21 no.5
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• pp.501-509
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• 2001

For the enhancement of ECT reliability on the primary water stress corrosion cracks of nuclear steam generator tubes, of which the occurrence is on the increase, it is important to comprehend the signal characteristics on crack morphology and to select an appropriate probe type. In this paper, the sizing accuracy and the detectability for the inner wall axial cracks of tubes were quantitatively evaluated using the following specimens: the electric discharge machined notches and the corrosion cracks which were developed on the operating steam generator tubes. The difference of eddy current signal characteristics between pancake and axial coil were also Investigated. The results obtained from this study provide a useful information for more precise evaluation on the inner wall axial tracks oi stram generator tubes.

• Journal of the Korean Society for Nondestructive Testing
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• v.20 no.5
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• pp.451-456
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• 2000

The development of an integrated technology including fabrication of corrosion defect specimens and their practical use is needed to enhance the reliability of eddy current test for nuclear steam generator tubing. In this paper, the necessity and importance are presented from the viewpoint of the structural integrity, simulation specimens for real defects, and experiences from the destructive examination of pulled tubes. The models for several corrosion defects we also briefly introduced, with the scheme for their practical use.

• Korean Chemical Engineering Research
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• v.57 no.1
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• pp.28-41
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• 2019

We carried out performance analysis of Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor. We analyzed transient-dynamic behavior of fluids inside the steam generator to vent into a sodium dump tank or a water dump tank when tubes in the steam generator were broken to cause a large-water-leak accident. Accordingly, we preliminarily evaluated design requirements of our system. Our results showed that sodium in the shell side of the steam generator and in Intermediate Heat Transport System was completely vented within 50 s and feed water in the tube side of the steam generator was completely vented within 2.5 s. It was analyzed that pressure of the tube side of the steam generator was higher than pressure of the shell side of the steam generator, which showed that sodium in the shell side did not flow into the tube side. Our results are expected to be used as basis information to performance analysis of Sodium-Water Reaction Pressure Relief System of Prototype Generation-IV Sodium-Cooled Fast Reactor.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05b
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• pp.110-117
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• 1997

Alloy 690 제1열 시제 전열관을 U 굽힘 가공할 시 전열관에 도입된 표면 잔류응력 및 굽힘 단면에서 치수변화 (벽두께, 진원도)를 위치별로 측정하여 평가하였다. 외측호(extrados)의 표면 잔류응력은 $\psi$=0$^{\circ}$에서 축 방향 응력이 -319 MPa (압축)로 가장 높았으며, 내측호(intrados)는 $\psi$=0$^{\circ}$, 160$^{\circ}$ 위치인 천이영역 부관에서 응력 변화가 크게 되는 경향을 보였다 측면(flank)은 인장 잔류응력으로 $\psi$=90$^{\circ}$(apex)에서 최대 190 MPa 로 축방향 응력으로 나타났다. 잔류응력치는 벽두께 보다는 진원도 변화와 일치되어 나타났으며, 시제 전열관의 벽두께 및 진원도는 ASTM의 치수 허용치 내에 포함되는 것으로 평가되었다. 잔류응력 측정은 스트레인 게이지를 이용한 구멍뚫기 방법 (Hole-Drilling Method)을 사용하였다.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.85-91
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• 1996

원전 증기발생기 시제 전열관으로 제조된 Alloy 600 및 690 에 대하여 ASTM 규정 (B163-86a)에 따라 확관실험을 실시하여 평가하였으며, 관 요소에 작용하는 응력을 해석하였다. 실험 결과 시제 전열관은 ASTM에서 요구하는 확관율 30% 및 그 이상의 35% 까지 확관할 경우에도 양호한 확관상태를 보였다. 확관에 따른 유동곡선의 축력은 Alloy 690 이 Alloy 600 에 비해 높았으며, 확관율의 증가에 따라 차이가 점진적으로 크지는 경향을 보였다. 얇은 벽 튜브의 확관에 대한 응력 해석식은 Modified Tresca's Yield Criterion를 도입하여 얻었으며, 소성변형식을 이용하여 확관율에 따른 응력을 예측하였다. 유동곡선의 이론 계산치와 실험치를 비교한 결과 Alloy 600의 경우 이론치는 실험치보다 약간 낮은 값으로 잘 일치되었으나, Alloy 690 경우는 Alloy 600에 비하여 확관율의 증가에 따라 차이가 커지는 경향을 보였다.

• Journal of the Korean Magnetics Society
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• v.15 no.5
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• pp.292-297
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• 2005

Steam generator of a nuclear power plant has important rolls for the heat transfer and the isolation of radioactive materials. So bursting of the steam generator tube is directly related to the accident of nuclear power plants. Incone1600 has been used for the steam generator tube material. The material shows non-magnetic and metallic properties, eddy current NDT method has been employed for defects detection. In this work, a differential type of eddy current probe was developed to improve resolution of defect detection. To verify properties of the developed differential type eddy current probe, we have made reference material with SUS304 which has similar magnetic and electrical properties of Inconel600. Using the developed differential type eddy current probe, we can detect defect size of 0.25 mm in diameter and 0.2 mm in depth (volume of $1{\times}10^{-3}\;mm^3$) with the reference material.

• Journal of Energy Engineering
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• v.12 no.2
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• pp.84-92
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• 2003

A one-dimensional thermal-hydraulic analysis computer program is developed for thermal sizing of a copper bonded steam generator. It is assumed that the conduction heat transfer of copper region between the hot side and the cold side tube is one-dimensional and its thermal resistance is derived as a function of a tube pitch. The flow regions of the water/steam side are divided into four regions: subcooled, saturated, film boiling, and super-heated. The number of tube selected ranges from 250 to 3500 and the pitch to tube diameter (P/D) ratios are 1.4, 1.6 and 1.8 for the parametric study calculation. The calculation results showed that when the number of tube was 2500, the length of the heating tube was about 12 m and the outside diameter of the steam generator was about 3 m. If the P/D ratio increases, the thermal resistance of copper component also increases, however the length of the heating tube is not so much increased.

• Journal of the Korean Society for Nondestructive Testing
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• v.28 no.2
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• pp.184-190
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• 2008

Various ECT techniques have been applied basically to assess the integrity of steam generator tithing in nuclear power plant. Among these techniques, the bobbin probe technique is applied generally to examine the volumetric flaws such as a crack-like defect and wear which is generally occurred on steam generator tubing, and additionally MRPC probe is used to examine closely tile top of tubesheet and bending regions due to the high possibility of cracking. Dent and bulge also may be formed on tube during installation process and operation of steam generator, but the dent and bulge indications greater than specific size criteria are recorded on examination report because these indications are not considered as flaw. These indications can be easily detected with bobbin probe and approximately sized with profile bobbin probe, but the size and shape can not be accurately verified. Accordingly, in this study, the $8{\times}1$ multi-array EC probe was designed to increase the measurement accuracy of the sectional profiling EC testing of tube. As a result, we would like to propose the application of $8{\times}1$ multi-array EC probe for the measurement of size and shape of profile change on steam generator tube in OPR-1000 nuclear power plant.

• Korean Chemical Engineering Research
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• v.43 no.4
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• pp.525-530
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• 2005

The erosion rate of water wall tube has been measured and discussed in a commercial circulating fluidized bed combustor (200 ton steam/hr, $4.97{\times}9.90{\times}28.98m\;height$). Tube thickness was measured with ultrasonic method. Severe tube erosion rate was observed in the splash region on all waterwalls including wingwalls. The tube erosion rate increased after an initial decrease as height from the distributor increased. The difference of erosion rate among wing walls was found due to unbalanced distribution of gas and solid flow rates. The erosion rate of the wing wall increased as location of the wing wall became closer to the center of combustor crosssection.

• Proceedings of the Korea Society for Energy Engineering kosee Conference
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• 1995.05a
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• pp.207-212
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• 1995

본 연구에서는 발전소 응축기를 시뮬레이션 할 수 있는 프로그램을 개발하였다. 관 내외 측 열전달계수의 계산에는 기존 상관식들과 응축 모델을 사용하였고 $\varepsilon$-NTU 방법을 사용하여 응축기를 해석하였다. 실제 응축기를 모사하기 위하여 관다발 보정계수 및 화울링 계수도 도입하였다. 이 프로그램을 사용하여 기존 평관을 대체할 전열촉진관의 형상을 도출하였다. 시뮬레이션 결과 전열촉진관을 사용하면 증기 응축 온도를 6 - 8 $^{\circ}C$ 정도 낮출 수 있음을 알 수 있었다.