• Transactions of the Korean Society of Mechanical Engineers
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• v.18 no.12
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• pp.3136-3148
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• 1994

This paper proposed the method of fracture integrity evaluation for semi-elliptical crack. Plane strain fracture toughnesses are used to compare with the thermal shock stress intensity factors for semi-elliptical crack obtained by Vainshtok weight function method. The method is applied to the finite Cr Mo V and 2.25Cr Mo steel plates with semi-elliptical crack under the thermal shock. For the purpose, tensile property and fracture toughness with respect to the temperature are measured. To verify the method, thermal shock experiments are carried. The theoretical predictions are in good agreement with the experiments.

• Journal of the Korean Society for Railway
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• v.6 no.2
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• pp.81-86
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• 2003

This study proposes the neural network simulator for the integrity evaluation of weld zone in railway rails. For these purposes, the ultrasonic signals for defects(crack) of weld zone in frames are acquired in the type of time series data and echo strength. The detection of the natural defects in railway truck is performed using the characteristics of echodynamic pattern in ultrasonic signal. And then their applications evaluated feature extraction based on the time-frequency-attractor domain(peak to peak, rise time, rise slope, fall time, fall slope, pulse duration, power spectrum, and bandwidth) and attractor characteristics (fractal dimension and attractor quadrant) etc. The constructed neural network simulator agrees fairly well with the measured results of test block(defect location, beam propagation distance, echo strength, etc). The Proposed neural network simulator in this study can be used for the integrity evaluation of weld zone in railway rails.

• Proceedings of the KSME Conference
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• 2004.11a
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• pp.160-165
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• 2004

An effective integrity evaluation system is essential to manage the fitness for service issues on infra-structure because the evaluation processes usually take long times and are detrimental for productivity point of view. In this paper, the key structures and procedures of four integrity evaluation programs which have been developed based on currently available codes and standards are described. The proposed programs are not only flexible to adopt advances in fitness for purpose type assessment methodologies but also convenient for field engineers. The developed programs which will be unified as an integrity evaluation system are expected to play a prominent role for integrity evaluation of major infra-structure.

• The KSFM Journal of Fluid Machinery
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• v.3 no.4 s.9
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• pp.7-14
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• 2000

This paper provides integrity evaluation and root cause analysis for defects observed at volute tongue, or cutwater, of the operating centrifugal pump in power plant. The cause of the cracks are analyzed and reviewed from the viewpoint of the operation and maintenance of the pumps, and the sample obtained from the cracked volute tongue of the pump are examined. At first, in-situ hardness test and microstructure examination were performed to understand the cause of cracking at volute tongue. The evaluation of structural integrity and the possibility of the crack propagation is also evaluated. Cracks were typical intergranular cracking and propagated along with prior austenite grain boundary. At easing volute tongue, the hardness was higher than ASTM requirement and a large amount of intergranular Cr carbide was precipitated. These were due to high C content in material. P content was also higher than ASTM requirement. Therefore, Cr carbide precipitation and P segregation at grain boundary, caused by higher C and P content in material, resulted in intergranular cracking of casing volute tongue. This procedure for integrity evaluation and root cause analysis is used to guide, and support the pump designer and manufacturer's material selection and process design to avoid a costly, unplanned outage of plant.

• Journal of the Korean Society for Precision Engineering
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• v.22 no.10 s.175
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• pp.137-143
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• 2005

The integrity of steam generator tubes in nuclear power plant should be maintained sufficiently during operation. For sake of this, complicated assessment procedures are required such as fracture mechanics analysis, etc. The integrity assessment of tubes has been performed by using conventional deterministic approaches while there are many uncertainties to carry out a rational evaluation. In this respect, probabilistic integrity assessment is considered as an alternative method for integrity assessment. The objectives of this study are to develop an integrity assessment system based on probabilistic fracture mechanics and to predict the failure probability of steam generator tubes containing an axial through-wall crack. The developed integrity assessment system consists of three evaluation modules, which apply first order reliability method, second order reliability method and Monte Carlo simulation method, respectively. The system has been applied to predict failure probability of steam generator tubes and the estimation results showed a promising applicability of the probabilistic integrity assessment system.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.14 no.7
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• pp.3121-3126
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• 2013

The integrity evaluation of pipes in nuclear power plant are essential for the safety of reactor vessel, and integrity must be assured when flaws are found. Accurate stress intensity analyses and crack growth rate data of surface-cracked components are needed for reliable prediction of their fatigue life and fracture strengths. Fatigue design and life assessment are the essential technologies to design the structures such as pipe, industrial plant equipment and so on. The effect of crack spacing on stress intensity factor K values was studied using three-dimensional finite element method (FEM). For the case of cylinder under internal pressure, a significant increase in K values observed at the deepest point of the surface crack. Also, this paper describes the fatigue analysis for cracked structures submitted to bending loads.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2092-2099
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• 2020

Many damages due to flow-accelerated corrosion and cracking have been observed during recent in-service inspections of nuclear power plants. To determine the operability or repair for damaged pipes, an integrity evaluation related to the damaged piping system should be performed by using already proven code and standards. One of them, the ASME Code Case is most popularly used to integrity assessment in nuclear power plants. However, the recent version of CC N-513 still recommends the simplified method which means a damaged elbow is assumed as an equivalent straight pipe. In addition, to enhance the accuracy integrity assessment in elbow, several previous studies recommend that the SIF and elastic COD values for an elbow with relatively large crack could be predicted by an interpolation technique. However, those estimates for elbow with relatively large crack might be derived to inaccurate results for crack growth analysis, such as for the allowable crack size and life estimation. Therefore, in this paper, the effect of crack length (0.3≤θ₁/π≤0.5) on SIF and elastic COD for elbow is systematically investigated. Then, for large crack in elbow, accurate estimates for SIF and elastic COD, which are widely used to assess the integrity of elbows, are proposed. Those proposed solutions are expected to be the technical basis for revisions of CC N-513-4 through the validation.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.25 no.2
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• pp.275-282
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• 2001

The reactor pressure vessel is usually cladded with stainless steel to prevent corrosion and radiation embrittlement. Number of subclad cracks may be found during an in-service-inspection due to the presence of cladding. It is specified, in ASME Sec. XI, that a subclad crack is characterized as a surface crack when the thickness of the clad is less than 40% of the crack depth. This condition is provided to keep the crack integrity evaluation conservative. In order to refine the fracture assessment procedures for such subclad cracks under a pressurized thermal shock condition, three dimensional finite element analyses are applied for various subclad cracks existing under cladding. A total of 36 crack geometries are analyzed, and the results are compared with those for surface cracks. The resulting stress intensity factors for subclad cracks are 6 to 44% less than those for surface cracks. It is proven that the flaw characterization condition as specified in ASME Sec. XI can be overly conservative for some subclad cracks.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.29 no.10 s.241
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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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.56-61
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• 2013

This paper presents the results of integrity assessment for the cracks happened in reactor vessel upper head penetration nozzles. The crack morphology for a boat sample from crack area was analyzed through microscope. The stress condition including weld residual stress around crack was analyzed using finite element analysis. From the results of crack morphology and stress condition, the crack was concluded as primary water stress corrosion cracking. The integrity of the cracked nozzle was assessed by the methodology provided in ASME Section XI. According to the assessment results, the remaining life of the cracked nozzle was 1.43 yrs. and the plant decided to repair it.