• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2005년도 춘계학술대회 논문집
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• pp.1803-1807
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• 2005

The pipe which it uses from the nuclear power plant or factory by a long period use and a corrosive action the inside defect occurs on the inside. abstract here. The ESPI method is in order to investigate the laser light in the measurement object it will be able to measure the wide territory whole in once, does not receive an effect in direction of defect not to be, has the strong point it will be able to measure a change of place arrowhead real-time defect. It measured a inside defect of pressure vessel by using Out of plane ESPI and In plane of ESPI. It compared a each method result.

• 한국정밀공학회지
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• 제24권12호
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• pp.104-110
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• 2007

Internal defects are mainly caused by a corrosive action and degradation in the pipe used in a nuclear power plant or factory. The ESPI method have the many advantages when compared with conventional method. The advantage are the area measurement ability at one time and non-contact measurement ability in the real-time. In this paper, we studied on the measurement of a internal defect by using out of plane ESPI technique. Here, we compared the experimental results using out of plane ESPI with the FEM results.

• 대한기계학회논문집A
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• 제25권9호
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• pp.1350-1358
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• 2001

During the last decade, a number of experiments and numerical analyses had been performed in conjunction with the development of simplified analytical methods to estimate the fracture behavior of cracked piping in nuclear power plant. However, the necessity of further investigation for the analytical methods was issued because of the discrepancies with the experimental data. The objective of this paper is to find out the optimum methods to evaluate the load-carrying capacities for cracked pipes. To do this, numerous analytical and finite element analyses were carried out for various pipe and crack geometries and materials. These results were synthesized for crack shapes and can be used as basic data for leak before analyses and risk informed inspections.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2006년 추계학술발표대회 개요집
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• pp.15-17
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• 2006

Weld residual stresses can be a major concern in structural integrity assessments such as a nuclear power plant. In this paper, detailed weld residual stress analyses were presented for a typical multi-pass weld of pipe-butt weld and plate T-butt weld. The calculated residual stress distributions were compared with those of the measured data and recommended profiles in R6 and BS7910. Defect assessment which is based on the stress intensity factor(SIF) calculations was carried out for a plate T-butt weld with cracks considering the weld residual stress distributions.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2000년도 추계학술대회논문집A
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• pp.730-735
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• 2000

The finite element analysis by the computer program SYSWELD in consideration of phase transformation was carried out to simulate the multi-pass welding process of SA106 Gr. C which is used for the main steam pipe in nuclear power plant. All the numerical results such as temperatures, the size of heat affected zone and the residual stresses were compared to the experimental results.

• Journal of Welding and Joining
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• 제19권3호
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• pp.334-341
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• 2001

The chemical and geometric effects of weld on flow-accelerated corrosion (FAC) of SA106 Gr.C low alloy steel pipe in 3.5wt% NaCl and simulated feedwater of nuclear power plant have been investigated by using rotating cylinder electrode. Polarization test and weight loss test were conducted and compared at rotating speed of 2000rpm (3.14m/s) with the variables of chemical and geometric parameters. The results showed that the chemical effects were relatively larger than the geometric effects, and the welded parts were the local anode and preferentially corroded, which could be explained by the differences between microstructural and compositional parameters. On the other hand, under active corrosion conditions, the heat affected zone were severely corroded and microstructural effects became the important role in the whole process.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2005년도 추계학술대회 논문집
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• pp.910-913
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• 2005

The pipe which it uses from the nuclear power plant or factory by a long period use and a corrosive action the inside defect occurs on the inside. abstract here. The ESPI method is in order to investigate the laser light in the measurement object it will be able to measure the wide territory whole in once, does not receive an effect in direction of defect not to be. has the strong point it will be able to measure a change of place arrowhead real-time defect. It measured a inside defect of pressure vessel by using ESPI and FEM. It compared a each method result.

• 한국압력기기공학회 논문집
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• 제17권1호
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• pp.56-63
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• 2021

The feedwater ring is an assembly in steam generator internal piping, which distributes feedwater into the secondary side of the steam generator. It consists of an assembly of carbon steel piping, pipe fittings and J-nozzles which are inserted into the top of the feedwater ring and welded to the diameter of the ring. The feedwater ring at the attachment region of the J-nozzle may be susceptible to flow accelerated corrosion (FAC) due to flow turbulence which increases local fluid velocities. If a J-nozzle becomes a loose part, it can cause damage to tubing near the tube sheet. In this paper, the structural stress analysis for a wall thinned feedwater ring and integrity evaluations under assumed loading conditions are carried out in compliance with ASME B&PV SecIII, NB-3200.

• 한국지진공학회논문집
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• 제17권1호
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• pp.21-31
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• 2013

In this study, dynamic characteristics and seismic capacity of the nuclear power plant piping system are evaluated by model test results using multi-platform shake table. The model is 21.2 m long and consists of straight pipes, elbows, and reducers. The stainless steel pipe diameters are 60.3 mm (2 in.) and 88.9 mm (3 in.) and the system was assembled in accordance with ASME code criteria. The dynamic characteristics such as natural frequency, damping and acceleration responses of the piping system were estimated using the measured acceleration, displacement and strain data. The natural frequencies of the specimen were not changed significantly before and after the testing and the failure and leakage of the piping system was not observed until the final excitation. The damping ratio was estimated in the range of 3.13 ~ 4.98 % and it is found that the allowable stress(345 MPa) according to ASME criteria is 2.5 times larger than the measured maximum stress (138 MPa) of the piping system even under the maximum excitation level of this test.

• 한국재난정보학회 논문집
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• 제12권1호
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• pp.1-9
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• 2016

국부적인 두께감소 문제에 대하여 직관부위를 대상으로 한 건전성 평가는 잘 알려져 있으나 엘보우를 대상으로 한 건전성 평가는 최근에 원자력 분야에서는 많은 연구가 이루어지고 있으나 석유화학 플랜트에서 이용되는 건전성 평가 지침서 중 하나인 API579 코드에는 아직 없는 실정이다. 이에 본 연구에서는 엘보우를 대상으로 엘보우의 외부(extrados)와 내부(intrados)에 두께 감소가 있다고 가정한 후 유한요소해석법을 이용하여 두께감소된 엘부우의 건전성평가 해석을 수행하였다. 본 해석 결과는 석유화학 플랜트에서 이용되는 엘보우의 건전성 평가에 이용될 수 있음을 보여주었다.