• 대한기계학회논문집A
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• 제38권6호
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• pp.655-662
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• 2014

본 논문에서는 상온과 원전 설계온도에서 증기발생기 전열관의 축방향과 원주방향 응력-변형률 거동과 인장물성치를 파악하기 위해서, 튜브 시편과 링 시편을 이용하여 상온과 $343^{\circ}C$에서 Alloy 690TT 전열관에 대한 인장시험을 수행하였다. 축방향 인장시험 결과 상온과 $343^{\circ}C$에서 모두 항복점 현상이 관찰되었으며, $343^{\circ}C$에서는 Serration이 관찰되었다. 축방향과 원주방향 모두 상온에 비해 $343^{\circ}C$에서 강도는 감소하였으나 연신율은 거의 변화가 없었다. $343^{\circ}C$에서 가공경화율은 상온에 비해 약간 감소하였으나, 가공경화 거동의 변화는 없었다. 시험 온도에 관계없이 축방향에 비해 원주방향의 항복강도와 인장강도가 약 5 10% 정도 낮았다. 시편 방향에 관계없이 상온 대비 $343^{\circ}C$에서 Alloy 690TT 전열관의 항복강도와 인장강도 감소는 ASME Sec.II의 온도 보정계수에 의해 예측된 것보다 큰 것으로 확인되었다.

• Tribology and Lubricants
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• 제21권1호
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• pp.33-38
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• 2005

Fretting is the oscillatory motion with very small amplitudes, which usually occurs between two solid surfaces in contact. Fretting wear is the removal of material from contacting surfaces through fretting action. Fretting wear of steam generator tubes in nuclear power plant becomes a serious problem in recent years. The materials for the tubes usually are Inconel 690 (I-690) and Inconel 600 (I-600). In this paper, fretting wear tests for I-690 and I-600 were performed under various applied loads in water at room temperature. Results showed that the fretting wear loss of I-690 and I-600 tubes was largely influenced by stick-slip. The fretting wear mechanisms were the abrasive wear in slip regime and the delamination wear in stick regime. Also, I-690 had somewhat better wear resistance than I-600.

• Tribology and Lubricants
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• 제24권6호
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• pp.362-367
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• 2008

Tubes in nuclear steam generators are held up by supports because the tubes are long and slender. Fluid flows of high-pressure and high-temperature in the tubes cause oscillating motions between tubes and supports. This is called as FIV (flow induced vibration), which causes fretting wear in contact parts of tube-support. The fretting wear of tube-support can threaten the safety of nuclear power plant. The tube and support materials were Inconel 690 and STS 409. The wear tests were conducted in various environments, which are in water without flow, in flowing water and in flowing water with air. The results showed that the flow of water influenced on the wear-life of tube. The wear-life of tube decreased in water flow as compared with wear-life in stationary water.

• Tribology and Lubricants
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• 제24권5호
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• pp.250-254
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• 2008

The steam generator of nuclear power plant is composed with the bundles of long tubes. It is exposed fluid flow and weak in the vibration. The tubes are supported by anti-vibration bars. Due to vibration the wear damage is called as the fretting wear. It should be minimized for the safety of the plants. The hard coatings are very effective to reduce the amount of wear. The coatings of TiN and CrN are introduced in this study to protect the fretting surfaces. The tube-on-flat type tester was used for fretting wear tests. The results show that the wear amounts of the coated tubes were decreased depending on coating thickness. CrN was very effective to reduce the wear. In case of TiN the wear amounts were dependent on the coating thickness. Thick coating of TiN was very effective for wear resistance.

• 대한기계학회논문집A
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• 제29권10호
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• pp.1361-1368
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• 2005

The structural and leakage integrity of steam generator tubes should be sustained all postulated loads with appropriate margin even if a crack is present. During the past three decades, for effective integrity evaluation, several limit load solutions have been used world-widely. However, to predict accurately load carrying capacities of specific components under different conditions, the solutions have to be modified by using lots of experimental data. The purpose of this paper is to propose a new burst pressure estimation scheme based on fracture mechanics analyses for steam generator tube with an axial or circumferential through-wall crack. A series of three dimensional elastic-plastic finite element analyses were carried out and, then, closed-form estimation equations with respect to both J-integral and crack opening displacement were derived through reference stress method. The developed engineering equations were utilized for structural integrity evaluation and the resulting data were compared to the corresponding ones fiom experiments as well as limit load solutions. Thereafter, since the effectiveness was proven by promising results, it is believed that the proposed estimation scheme can be used as an efficient tool for integrity evaluation of cracked steam generator tubes.

• 전자공학회논문지S
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• 제35S권2호
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• pp.137-145
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• 1998

In this paper, we propose a tube-hole center detection vision algorithm verifying the position of a tele-controlled robot and providing visual information for increasing reliability and efficiency in the diagnosis of steam generator (SG) tubes in nuclear power plant. A tele-controlled robot plays a role in carrying the probe used in inspecting the integrity of SG tubes. Thus accurately locating a tele-controlled robot on the desired tube-hole center is important issue for reliability of inspection. To do this work, we have to find the tube-hole center locations from the input image. At first, we apply the three-class segmentation method modified for this application. WE extract minimum bounding rectangles (MBRs) in the theresholded binary image. Second, for discriminating between MBR by tube and MBR by noise, we introduce the MBR rejection rules as knowledge-based rule set. MBRs are divided into the very dark region MBRs and the very bright region MBRs. In order to describe the region of complete tube-hole, the MBRs need a process of pairing each other. We then can find the tube-hole center from the paired MBR. For more accurately finding the tube-hole center in several sequential images, the centers of some frames need to be averaged. We tested the performance of our method using hundreds of real images.

• 한국정보통신설비학회:학술대회논문집
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• 한국정보통신설비학회 2009년도 정보통신설비 학술대회
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• pp.54-58
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• 2009

In this paper, the ECT(eddy current testing) signal analysis of eddy current array probe for inspection of SG(steam generator) tube in NPP(nuclear power plant) using electromagnetic FEM(finite element method) was performed. To obtain the electromagnetic characteristics of probes, the governing equation was derived from Maxwell's equation, and the problem was solved by using the 3-dimensional FEM. The types of defects were FBH(flat bottomed hole) and OD groove, Spiral groove, natural defects(pitting, SCC, multiple SCC, wear). The depth of FBH defects were 20%, 40%, 60%, 80%, 100 of SG tube thickness, and it was assumed that the defects were located on the tube outside. And the operation frequency of 100kHz, 300kHz and 400kHz 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 variation of size and depth on FBH defects and operation frequencies. The results in this paper can be helpful when the ECT signals from EC array probe are evaluated and analyzed.

• 한국정보통신설비학회:학술대회논문집
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• 한국정보통신설비학회 2006년도 하계학술대회
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• pp.241-245
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• 2006

In the examination of Steam Generator (SG) tube in Nuclear Power Plant (NPP) Eddy Current Testing (ECT) probes play an Important role in detecting the defects. Bobbin probe and Rotating Pancake Coil (RPC) probe is usually used for the inspection of SG tube. Bobbin probe is good at high speed inspection, but ability of detection of circumferential defect is very weak. On the contrary RPC probe, which moves for inspection in the direction of axial and circumferential simultaneously, has very slow inspection speed, but it was excellent detection capability fur small cracks, which is hardly detected by bobbin probe. Many examinations of SG tube examination of NPP are achieved during short period. Therefore, solution about this must develop probe of new form for examination performance and examination time shortening of other probe. In this paper, analyzed technological present condition of Bob-bin probe and RPC probe been using in Nondestructive Testing (NDT) for SG tube defect detection and Appeared about background theory and view of developed probe newly.

• 화약ㆍ발파
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• 제17권4호
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• pp.18-31
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• 1999

The explosive forming has been used for many year to expand tubes into tubesheets. this process has demonstrated ability to direct carefully the energy of an explosive to expand tubes into tubesheet holes without damaging the tubesheet and without causing the excessive cold work at the tube I.D. that is normally associated with mechanical expansion. The success of explosive tube expansion provided the background for the development of the explosive tube plug. The main results are as follows : (1) The optimum explosives and explosive qualities are PETN, RDX, HMS and about 18~31gr/ft of explosive plugging in nuclear steam generator. (2) Explosive plugging's thickness is 0.9~1.8mm. If groove of 0.4 mm formed in plug outside, For the hydraulic leakage is go up, explosive plugging of formed groove are applicate tube and tubrplate. (3) Sheath is designed on the polyethylene of low density, In thermal impact test of the $430^\circ{C}$, hydraulic leakage is $300kg/cm^2$. (4) About 10~60mm oxide inclusions are existed on the space of explosive plug and tube protect to the leakage.

• 대한기계학회논문집A
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• 제30권10호
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• pp.1255-1260
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• 2006

To maintain the structural integrity of steam generator tubes, 40% of wall thickness plugging criterion has been developed. The approach is for the steam generator tube with single crack, so that the interaction effect of multiple cracks can not be considered. Although, recently, several approaches have been proposed to assess the integrity of steam generator tube with two identical cracks whilst actual multiple cracks reveal more complex shape. In this paper, the coalescence pressure of steam generator tube containing multiple cracks of different length is evaluated based on the detailed 3-dimensional (3-D) elastic-plastic finite element (FE) analyses. In terms of the crack shape, two collinear axial through-wall cracks with different length were considered. Furthermore, the resulting FE coalescence pressures are compared with FE coalescence pressures and experimental results for two identical collinear axial through-wall cracks to quantify the effect of crack length ratio on failure behavior of steam generator tube with multiple cracks. Finally, based on 3-D FE results, the coalescence evaluation diagrams were proposed.