• Smart Structures and Systems
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• v.4 no.5
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• pp.517-530
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• 2008

This paper deals with civil infrastructures in general, sensor and smart structure technology, and smart steel structures in particular. Smart structures technology, an integrated engineering field comprising sensor technology, structural control, smart materials and structural health monitoring, could dramatically transform and revolutionize the design, construction and maintenance of civil engineering structures. The central core of this technology is sensor and sensor networks that provide the essential data input in real time for condition assessment and decision making. Sensors and robust monitoring algorithms that can reliably detect the occurrence, location, and severity of damages such as crack and corrosion in steel structures will lead to increased levels of safety for civil infrastructure, and may significantly cut maintenance or repair cost through early detection. The emphasis of this paper is on sensor technology with a potential use in steel structures.

• Structural Engineering and Mechanics
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• v.24 no.1
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• pp.91-105
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• 2006

In this paper, post-earthquake capacity evaluation method of reinforced concrete buildings was studied. Substructure pseudo-dynamic test and static loading test of first story column in a four-story R/C building was carried out in order to investigate the validity of the evaluation method proposed in the Damage Assessment Guideline (JBDPA 2001). In pseudo-dynamic test, different levels of damage were induced in the specimens by pre-loading, and input levels of seismic motion, at which the specimens reached to the ultimate stage, were examined. From the experimental result, no significant difference in damage levels such as residual crack width between the specimens under static and pseudo-dynamic loading was found. It is shown that the seismic capacity reduction factors ${\eta}$ can provide a reasonable estimation of post-earthquake seismic capacity of R/C buildings suffered earthquakes.

• Proceedings of the Korea Concrete Institute Conference
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• 2010.05a
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• pp.165-166
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• 2010

Collision scenario was 12 cases considering gate location, water level and lateral location of collision etc. And then, analysis result of trunnion by collision loads (reservoir side gate). Compressive fracture may not occur because the maximum compressive stress of concrete is below the allowable compressive strength. but, it is possible to appear some local crack because the maximum tensile stress exceed the tensile strength.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.7
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• pp.1006-1013
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• 2004

Operating experience of steam generators shows that the tubes are degraded by stress corrosion cracking, fretting wear and so on. These defected tubes could stay in service if it is proved that the tubes have sufficient structural margin to preclude the risk of tube bursting. This paper provides detailed plastic limit pressure solutions for through-wall cracks in the steam generator tubes. These are developed based on three dimensional(3D) finite element analyses assuming elastic-perfectly plastic material behavior. Both axial and circumferential through-wall cracks in free span and in u-bend regions are considered. The resulting limit pressure solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes.

• Nuclear Engineering and Technology
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• v.33 no.5
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• pp.498-513
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• 2001

Performed here is a comparative assessment study for the generation of the pressure- temperature (P/T) limit curve of the reactor vessel. Using the cooling or heating rate and vessel material properties, the stress distribution is obtained to calculate stress intensity factors, which are compared with the material fracture toughness to determine the relations between operating pressure and temperature during cool-down and heat-up. P/T limit curves are generated with respect to crack direction, clad thickness, toughness curve, cooling or heating rate and neutron fluence, and their results are compared.

• Journal of the Korean Society for Nondestructive Testing
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• v.9 no.2
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• pp.25-36
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• 1989

This essay is a general review of the application of ultrasonic NDE techniques to the performance assessment and characterization of composite materials. A brief review of ultrasonic input-output characterization of a composite plate by shear waves is presented. A theoretical development of ultrasonic wave propagation in isotropic and anisotropic media excited, respectively, by a circular transducer and an oscillatory point source is summarized. Some experimental results are described in which ultrasonic velocity and attenuation measurements give insight into material degradation of fatigued composite laminates. Ultrasonic determination of the elastic constants of a composite plate and an experimental attempt at ultrasonic testing of an isotropic plate containing a crack are also included. A recent effort for the characterization of viscoelastic materials using the ultrasonic NDE technique is outlined. Finally, the reliability of ultrasonic NDE is briefly touched upon.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.32 no.12
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• pp.1123-1131
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• 2008

This paper provides V-factor estimation under combined mechanical and thermal load for circumferential cracks. Results are based on finite element analyses and effect of types and magnitudes of the thermal stress, crack geometry, the loading mode and plastic strain hardening on variations of the V-factor are investigated. The results of finite element analyses are compared with R6 values. As a result, it is shown that R6 gives generally conservative results. The conservatism is especially increased for the combination of large mechanical and thermal load. As a result, new estimation method which uses failure assessment line in R6 is proposed for V-factor and gives less conservative results.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.459-464
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• 2001

For major nuclear power plant components periodic inspections and integrity assessments are needed for the safety. But many flaws are undetectable due to sampling inspection. Probabilistic integrity assessment is applied to take into consideration of uncertainty and variance of input parameters arise due to material properties, applied load and undetectable flaws. This paper describes a Probabilistic Fracture Mechanics(PFM) analysis based on Monte Carlo(MC) algorithms. Taking important parameters as probabilistic variables such as fracture toughness, crack growth rate and flaw shape, failure probability of major nuclear power plant components is archived as a results of MC simulation. For the verification of these analysis, a comparison study of the PFM analysis using other commercial code, mathematical method is carried out and a good agreement was observed between those results.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.322-327
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• 2001

The internal leaks of RCS pressure boundary valves may cause thermal fatigue crack because of the TASCS in RCS branch line. After experienced unisolable piping failures in several PWR plants, many studies have peformed to understand these phenomena and various methods were applied to ensure the structural integrity of piping. In this paper, the cause of unisolable piping failures and the alternatives to prevent recurrence of failure were reviewed. Also, the severity of piping failure including susceptibility of valve leaks was evaluated for the Westinghouse 2-loop plant. The length of turbulent penetration on RHR inlet piping was measured and, thermal fluid analysis and fatigue analysis was performed for this piping. As a means of ensuring the structural integrity, temperature monitoring and specialized UT and other alternatives were compared for the further application.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.1
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• pp.39-47
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• 2002

Performed here is an assessment study for deterministic fracture mechanics analysis of a pressurized thermal shock(PTS). The PTS event means an event or transient in pressurized water reactors(PWRs) causing severe overcooling(thermal shock) concurrent with or followed by significant pressure in the reactor vessel. The problems consisting of two transients and 10 cracks are solved and maximum stress intensity factors and maximum allowable nil-ductility reference temperatures are calculated. Their results are compared each other to address the general characteristics between transients, crack types and analysis methods. The effects of elastic-plastic material behavior and clad coating on the inner surface are explored.