• 한국윤활학회:학술대회논문집
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• 한국윤활학회 2002년도 제35회 춘계학술대회
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• pp.304-309
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• 2002

Fretting wear test in room temperature water was performed to evaluate the wear coefficient of Inconel 600,690 (Pressurized Water Reactor, PWR) and Alloy 800 (CANadian DeuteriumUranium, CANDU) steam generator (SG) tubes against ferritic and martensitic stainless steels. The main focus is to compare the wear behaviors between Alloy 800 and Inconel alloys. Test conditions are $10{\sim}30N$ of normal load, $200{\sim}450{\mu}m$ of sliding amplitude and 30Hz of frequency. The result indicated that the wear rate of Alloy 800 was higher than those of Inconel 690 at various test condition such as normal loads, sliding amplitudes etc. From the results of SEM observation, there was little evidence of plastic deformation layer that were dominantly formed on the worn surfaces of Inconel 690. Also, wear particles in Alloy 800 were released from contacting asperities deformed by severe plastic flow during fretting wear. Main cause of wear rate between Alloy 800 and Inconel 690 may be due to the difference of hardness between martensitic and ferritic stainless steel. The wear rate and wear mechanism of two tubes in room temperature water are discussed.

• Corrosion Science and Technology
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• 제15권2호
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• pp.84-91
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• 2016

The transport and deposition of corrosion products in pressurized water nuclear reactor (PWR) steam generators have led to corrosion (SCC, denting etc.) problems. Lancing, mechanical cleaning and chemical cleaning have been used to reduce these problems. The methods of lancing and mechanical cleaning have limitations in removing corrosion products due to the structure of steam generator tubes. But high temperature chemical cleaning (HTCC) with EDTA is the most effective method to remove corrosion products regardless of the structure. However, EDTA in chemical cleaning aqueous solution and chemical cleaning environments affects the integrity of materials used in steam generators. The nuclear power plants have to perform the pre-test (also called as qualification test (QT)) that confirms the effect on the integrity of materials after HTCC. This is one of the series studies that assess the effect, and this study determines the effects of 20 % EDTA aqueous solution on defective tubes in high temperature chemical cleaning environments. The depth and magnitude of defects in steam generator (SG) tubes were measured by eddy current test (ECT) signals. Surface analysis and magnitude of defects were performed by using SEM/EDS. Corrosion rate was assessed by weight loss of specimens. The ECT signals (potential and depth %) of defective tubes increased marginally. But the lengths of defects, oxides on the surface and weights of specimens did not change. The average corrosion rate of standard corrosion specimens was negligible. But the surfaces on specimens showed traces of etching. The depth of etching showed a range on the nanometer. After comprehensive evaluation of all the results, it is concluded that 20 % EDTA aqueous solution in high temperature chemical cleaning environments does not have a negative effect on defective tubes.

• Nuclear Engineering and Technology
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• 제54권4호
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• pp.1374-1381
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• 2022

Under the influence of nuclear radiation, the reliability of steam generators (SGs) is an important factor in the efficiency and safety of nuclear power plant (NPP) reactors. Motion planning that remotely manipulates an SG mobile tube-inspection robot to inspect SG heat transfer tubes is the mainstream trend of NPP robot development. To achieve motion planning, conditional traversal is usually used for base position optimization, and then the A^* algorithm is used for path planning. However, the proposed approach requires considerable processing time and has a single expansion during path planning and plan paths with many turns, which decreases the working speed of the robot. Therefore, to reduce the calculation time and improve the efficiency of motion planning, modifications such as the matrix method, improved parent node, turning cost, and improved expanded node were proposed in this study. We also present a comprehensive evaluation index to evaluate the performance of the improved algorithm. We validated the efficiency of the proposed method by planning on a tube sheet with square-type tube arrays and experimenting with Model SG.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3225-3234
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• 2022

Alloy 690 tubing has been shown to be highly resistant to primary water stress corrosion cracking (PWSCC). Nevertheless, predicting the failure by PWSCC in Alloy 690 SG tubes is indispensable. In this work, a Bayesian-based statistical approach is proposed to predict the occurrence of failure by PWSCC in Alloy 690 SG tubing. The prior distributions of the model parameters are developed based on the prior knowledge or information regarding the parameters. Since Alloy 690 is a replacement for Alloy 600, the parameter distributions of Alloy 600 tubing are used to gain prior information about the parameters of Alloy 690 tubing. In addition to estimating the model parameters, analysis of tubing reliability is also performed. Since no PWSCC has been observed in Alloy 690 tubing, only right-censored free-failure life of the tubing are available. Apparently the inference is sensitive to the choice of prior distribution when only right-censored data exist. Thus, one must be careful in choosing the prior distributions for the model parameters. It is found that the use of non-informative prior distribution yields unsatisfactory results, and strongly informative prior distribution will greatly influence the inference, especially when it is considerably optimistic relative to the observed data.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2009년도 제40회 하계학술대회
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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.

• 전기학회논문지
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• 제58권12호
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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.

• Nuclear Engineering and Technology
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• 제32권2호
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• pp.108-127
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• 2000

Hot leg break LBLOCA(Large Break LOCA) had a potential to be a containment maximum pressure accident in YGN3&4, which was induced from excessive conservatism in the CE analysis methodology of mass and energy release. This study conducted mass and energy release experiment for the hot leg break LBLOCA during post blowdown with an integral test facility, SNUF(Seoul National University Facility). This facility simulated YGN 3&4 with volume ratio of 1/1140 based on Ishii's three level scaling. Experiment showed that SI(Safety Injection) water refilled cold leg first and core later. SI water was vaporized in the core, which resulted in the repressurization of reactor. This increase of pressure drove the water in cold leg to flow up half height of U tubes. However, since the water was drained back soon, the release through the SG side broken section by evaporation was negligibly small. This study also provided experimental assessment of RELAP5 results by KAERI for the release through the SG side broken section.

• Nuclear Engineering and Technology
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• 제52권6호
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• pp.1304-1317
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• 2020

The tubes in a heat exchanger, such as a steam generator (SG), are subjected to crossflow, and interaction between tubes and supports can happen, which can cause fretting wear of tubes. Although many experiments and models have been established, some detailed mechanisms are still not sufficiently clear. In this work, more attention is paid to obtain the regulation of impact and sliding in the complex process and many factors, such as excitation forces and clearances. The responses and contact forces were analyzed to obtain clear understanding of the influences of these factors. Room temperature tests in the air were established. The results show that the effect of clearance on the normal work rate is not monotonous and instead has two peaks. The force ratio can influence the normal work rate by changing the distribution of contact angles, which can result in higher sliding in the contact process. Fretting wear tests are conducted, and the wear surfaces are analyzed by a scanning electron microscope (SEM) and energy dispersive X-ray spectrometer (EDX). The results of this work can serve as a reference for impactsliding contact analysis between AVBs and tubes in steam generators.

• Nuclear Engineering and Technology
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• 제45권4호
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• pp.513-522
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• 2013

The ODSCC detected in the TSP position of Ulchin 3&4 SGs are typical ODSCC of Alloy 600MA tubes. The causative chemical environment is formed by concentration of impurities inside the occluded region formed by the tube surface, egg crate strips, and sludge deposit there. Most cracks are detected at or near the line contacts between the tube surface and the egg crate strips. The region of dense crack population, as defined as between $4^{th}$ and $9^{th}$ TSPs, and near the center of hot leg hemisphere plane, coincided well with the region of preferential sludge deposition as defined by thermal hydraulics calculation using SGAP computer code. The cracks developed homogeneously in a wide range of SGs, so that the number of cracks detected each outage increased very rapidly since the first detection in the $8^{th}$ refueling outage. The root cause assessment focused on investigation of the difference in microstructure and manufacturing residual stress in order to reveal the cause of different susceptibilities to ODSCC among identical six units. The manufacturing residual stress as measured by XRD on OD surface and by split tube method indicated that the high residual stress of Alloy 600MA tube played a critical role in developing ODSCC. The level of residual stress showed substantial variations among the six units depending on details of straightening and OD grinding processes. Youngwang 3&4 tubes are less susceptible to ODSCC than U3 and U4 tubes because semi-continuous coarse chromium carbides are formed along the grain boundary of Y3&4 tubes, while there are finer less continuous chromium carbides in U3 and U4. The different carbide morphology is caused by the difference in cooling rate after mill anneal. There is a possibility that high chromium content in the Y3&4 tubes, still within the allowable range of Alloy 600, has made some contribution to the improved resistance to ODSCC. It is anticipated that ODSCC in Y5&6 SGs will be retarded more considerably than U3 SGs since the manufacturing residual stress in Y5&6 tubes is substantially lower than in U3 tubes, while the microstructure is similar with each other.

• 비파괴검사학회지
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• 제29권2호
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• pp.137-144
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• 2009

본 논문에서는 전자기 유한요소 해석을 통하여 원전 증기발생기(SG, steam generator) 세관의 결함 크기 변화에 따른 배열 와전류 프로브의 와전류탐상 특성을 해석하였다. 프로브의 전자기적 특성을 해석하기 위하여 맥스웰 방정식을 이용하여 지배방정식을 유도하였고, 이를 3차원 전자기 유한요소법을 이용하여 문제를 해석하였다. 해석을 위해 선정한 결함은 평저공(FBH, flat bottomed hole) 결함을 선정하였다. FBH결함에 대해 결함의 위치를 관의 외부표면에 존재하게 하고 결함의 깊이는 세관 두께의 20%, 40%, 60%, 80%, 100%로 하였다. 또한 결함의 크기변화 및 시험주파수를 100 kHz, 300 kHz, 400 kHz로 변화시켜 해석하였다. 해석 대상으로는 원자력발전소 증기발생기 세관으로 사용되고 있는 Inconel 600 도체관을 사용하였다. 본 논문을 통하여 결함형상, 깊이 및 크기, 시험주파수의 변화에 따른 탐상신호의 변화를 확인할 수 있었다. 본 논문의 결과는 배열 와전류 프로브의 와전류탐상 신호 평가시 도움이 될 것이다.