• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.12
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• pp.1897-1905
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• 2004

In many part of machines or structures that made of bimaterial bonded with two dissimilar materials, most failures occur at their interface. Therefore, the accurate analysis of fracture characteristics and the evaluation of mechanical strength for interfacial crack are essential when we design those structures. In this research, stress and displacement components in the vicinity of stationary interfacial crack tip in the two dissimilar isotropic bimaterials are established. Hereafter, the stress components established in this research can be applied to the photoelastic hybrid method which can be used to analyze the fracture behavior of the two dissimilar isotropic bimaterials.

• Transactions of the Korean Society of Mechanical Engineers
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• v.16 no.11
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• pp.2072-2081
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• 1992

Theoretical fringe patterns were calculated and regenerated by using power series type Williams equations and coefficients estimated from the photoelastic data. Results of calculated values were evaluated by comparing experimental data points with the regenerated theoretical fringe loops. Statistical accuracy evaluation between regenerated fringe values and experimental ones showed that standard deviation was minimum and correlation coefficient was maximum when the first four terms of Wiliams equations were used.

• Proceedings of the KSME Conference
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• 2000.11a
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• pp.286-291
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• 2000

The reactor pressure vessel is usually cladded with stainless steel to prevent corrosion and radiation embrittlement, and number of subclad cracks have been found during an in-service-inspection. Therefore assessment for subclad cracks should be made for normal operating conditions and faulted conditions such as PTS. Thus, in order to find the optimum fracture assessment procedures for subclad cracks under a pressurized thermal shock condition, in this paper, three different analyses were performed, ASME Sec. XI code analysis, an LEFM(Liner elastic fracture mechanics) analysis and an EPFM(Elastic plastic fracture mechanics) analysis. The stress intensity factor and the Maximum $RT_{NDT}$ were used for characterizing. Analysis based on ASME Sec. XI code does not completely consider the actual stress distribution of the crack surface, so the resulting Maximum allowable $RT_{NDTS}$ can be non-conservative, especially for deep cracks. LEFM analysis, which does not consider elastic-plastic behavior of the clad material, is much more non-conservative than EPFM analysis. Therefore, It is necessary to perform EPFM analysis for the assessment of subclad cracks under PTS.

• Steel and Composite Structures
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• v.47 no.2
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• pp.239-251
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• 2023

The stiffness evaluation of cracked base metal is of great guidance to fatigue crack reinforcement. By carrying out fatigue tests and numerical simulation of typical cracking details in steel box girder, the strain-degradation law of cracked base metal was analyzed and the relationship between base metal stress and its displacement (stiffness) was explored. The feasibility of evaluating the stress of cracked base metal based on the stress field at the crack tip was verified. The results demonstrate that the stiffness of cracked base metal shows the fast-to-slow degradation trend with fatigue cracking and the base metal at 50mm or more behind the crack tip basically lose its bearing capacity. Drilling will further accelerate stiffness degradation with the increase of hole diameters. The base metal stress has a negative linear relation with its displacement (stiffness), The stress of cracked base metal is also related to stress intensity factor and its relative position (distance, included angle) to the crack tip, through which the local stiffness can be effectively evaluated. Since the stiffness is not uniformly distributed along the cracked base metal, the reinforcement patch is suggested to be designed according to the stiffness to avoid excessive reinforcement for the areas incompletely unloaded.

• Journal of the Korean Society of Industry Convergence
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• v.26 no.4_2
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• pp.571-576
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• 2023

In this study, the harmless crack size(a_hml) according to the residual stress depth was evaluated using the fatigue limit of SCM440 steel by quenching-tempering(QT) and induction hardening(IH), and threshold stress intensity factor of QT steel. Because the residual stress increased rapidly as the crack depth increased, a_hml was determined at the depth of all the crack aspect ratio(As) regardless of Type I-III, and a_hml also increased according to the residual stress depth. a_hml was minimal at As=1.0 and maximal at As=0.1, but was almost similar on each Type. a_hml was small the dependence on As.

• Journal of the Korean Society for Precision Engineering
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• v.19 no.11
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• pp.174-182
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• 2002

The integrity evaluation of pressure tube is essential for the safety of CANDU reactor, and integrity must be assured when flaws or contacts between pressure tube and surrounding calandria tube are found. In order to complete the integrity evaluation, not only complicated and iterative calculation procedures but also a lot of data and knowledge are required. For this reason, an integrity evaluation system, which provides an efficient way of the evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec.？ and FFSG issued by the AECL, and applicable for the evaluation of blister, sharp flaw and blunt notch. Delayed hydride cracking and blister evaluation modules are included in the general flaw and notch evaluation module. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• Proceedings of the KSME Conference
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• 2000.11a
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• pp.843-848
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• 2000

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle and it's containment vessel. If a flaw is found during the periodic inspection from the pressure tubes, the integrity evaluation must be carried out, and the safety requirements must be satisfied for continued service. In order to complete the integrity evaluation, complicated and iterative calculation procedures are required. Besides, a lot of data and knowledge for the evaluation are required for the entire integrity evaluation process. For this reason, an integrity evaluation system, which provides efficient way of evaluation with the help of attached databases, was developed. The developed system was built on the basis of ASME Sec. XI and FFSG(Fitness For Service Guidelines for zirconium alloy pressure tubes in operating CANDU reactors) issued by the AECL, and covers the delayed hydride cracking(DHC). Various analysis methods are provided for the integrity evaluation of pressure tube. In order to verify the developed system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.534-540
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• 2001

One approach to testing the suitability of an adhesive joint for a particular application is to build and test to destruct ion a representative sample of the joint. The nondestructive test will not measure strength directly but will measure a parameter which can be correlated to strength. It is therefore, essential that a suitable nondestructive test is chosen and that its results are correctly interpreted. In this paper, typical defects found in adhesive joints are described together with their significance. The limits and likely success of current physical nondestructive tests are described, and future trends outlined.

• Journal of Mechanical Science and Technology
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• v.17 no.7
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• pp.947-957
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• 2003

The pressure tube is a major component of the CANDU reactor, which supports nuclear fuel bundle. In order to complete the integrity evaluation of pressure tube, expert knowledge, iterative calculation procedures and a lot of input data are required. More over, results of integrity assessment may be different according to the evaluation method. For this reason, an integrity evaluation system, which provides efficient way of evaluation with the help of attached database, was developed. The present system was built on the basis of 3D FEM results, ASME Sec. XI, and Fitness For Service Guidelines for CANDU pressure tubes issued by the AECL (Atomic Energy Canada Limited). The present system also covers the delayed hydride cracking and the blister evaluation, which are the characteristics of pressure tube integrity evaluation. In order to verify the present system, several case studies have been performed and the results were compared with those from AECL. A good agreement was observed between those two results.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.8 s.179
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• pp.2083-2090
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• 2000

In order to manage nuclear power plants safely and cost effectively, it is necessary to develop integrity evaluation methodologies for the main components. Recently, the integrity evaluation techniques were broadly studied regarding the license renewal of nuclear power plants which were approaching their design lives. Since the integrity evaluation process requires special knowledges and complicated calculation procedures, it has been allowed only to experts in the specified area. In this paper, an integrity evaluation system for reactor pressure vessel was developed. RVIES(Reactor Vessel Integrity Evaluation System) provides four specific integrity evaluation procedures covering PTS(Pressurized Thermal Shock) analysis, P-T(Pressure-Temperature) limit curve generation, USE(Upper Shelf Energy) analysis and Fatigue analysis. Each module was verified by comparing with published results.