• Transactions of the Korean Society of Automotive Engineers
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• v.15 no.5
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• pp.139-145
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• 2007

Fretting is a kind of surface damage mechanism observed in mechanically jointed components and structures. The initial crack under fretting damage occurs at lower stress amplitude and lower cycles of cyclic loading than that under plain fatigue condition. This can be observed in automobile and railway vehicle, fossil and nuclear power plant, aircraft etc. In the present study, railway axle material RSA1 used for evaluation of fretting fatigue life. Plain and fretting fatigue tests were carried out using rotary bending fatigue tester with proving ring and bridge type contact pad. Through these experiments, it is found that the fretting fatigue limit decreased about 37% compared to the plain fatigue limit. In fretting fatigue, the wear debris is observed on the contact surface, and oblique cracks at an earlier stage are initiated in contact area. These results can be used as useful data in a structural integrity evaluation of railway axle.

• Transactions of the Korean Society of Machine Tool Engineers
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• v.10 no.1
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• pp.133-140
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• 2001

Recently, to assure the integrity of a structural components such as piping pressure vessels and thinning structure, Lamb wave inspection technique has been used in material evaluation. It is very important to select the optimal Lamb wave mode and to analyze the signal accurately because of its unique dispersion properties grnerating several modes within the speci-men. It this study, the feasibility of material evaluation applications using wavelet analysis of Lamb wave has been veir-fied experimentally. These results show as follows; 1)dispersion characteristic of each mode in dispersion curve is demon-strated that A0 mode propagating material surface is useful mode having the lest energy loss and not sensitive to surface condition. 2) it can be detected even the micro defect ($1\times2mm$) fabricated in ultrasonic probe flaw distance (290mm) to axis direction. 3) the wavelet transform which is called "time-frequency analysis" shows the Lamb wave propagation due to the change of materials characterization can be evaluated at each frequency and experimental group velocity of Lamb wave agrees quite well with that of simulated dispersion curve.ion curve.

• Journal of the Korean Society of Safety
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• v.32 no.2
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• pp.1-6
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• 2017

Modern society has been experiencing by population growth and urbanization that bring, a change of eating habits which has occurred a various types of waste in a large amount. Even though these wastes are required an immediate treatment with difficulties unsanitary handling and existing waste treatment method are by incineration, fermentation, drying and etc. however a bad smell occurs after the treatment that need's a lot of energy in processing organic wastes with high moisture contents and wasteful and inefficient problem. The strength assessment of the organic waste agitating vessel is required in terms of safety due to the differences of loading on the shaft that was treated by agitating the mixture of food waste. The damage of agitating axis is depended on steam pressure, temperature condition and the force moment that exerted by the food waste. Thus the strength assessment and stability evaluation are very important, especially to handle a hard waste. In this study the rotation capacity of agitation is about 5 tons considering general structural rolled steel pressure vessel strength and steam pressure. The purpose is to estimate the safety and strength evaluation for a agitator axis and impellers according to the rotating angle of the axis under the condition of the 3.2 ton capacity reactor.

• Transactions of the Korean Society of Automotive Engineers
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• v.18 no.2
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• pp.74-82
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• 2010

Fretting damage can be observed in automobile and railway vehicle, fossil and nuclear power plant, aircraft etc. In the present study, railway axle material RSA1 used for evaluation of fretting fatigue life. Plain and fretting fatigue tests were carried out using rotary bending fatigue tester with proving ring and bridge type contact pad. Through these test, the following results are obtained: 1) it is found that the fretting fatigue limit of standard specimen decreased about 37% compared to the plain fatigue limit. 2) The early crack of Shinkansen type specimens initiated in contact area and final fractured below samp=214 MPa. 3) The early crack of all TGV type specimens initiated in rounded area and fractured. 4) Tire tracks and rubbed scars were observed in the oblique crack region and fatigue crack growth region of fracture surface. 5) The wear debris is observed on the contact surface, and oblique cracks at an earlier stage are initiated in contact area. These results can be used as useful data in a structural integrity evaluation of railway axle.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.895-906
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• 2015

Background: Mitigation of primary water stress corrosion cracking (PWSCC) is a significant issue in the nuclear industry. Advanced nickel-based alloys with lower susceptibility have been adopted, although they do not seem to be entirely immune from PWSCC during normal operation. With regard to structural integrity assessments of the relevant components, an accurate evaluation of crack growth rate (CGR) is important. Methods: For the present study, the extended finite element method was adopted from among diverse meshless methods because of its advantages in arbitrary crack analysis. A user-subroutine based on the strain rate damage model was developed and incorporated into the crack growth evaluation. Results: The proposed method was verified by using the well-known Alloy 600 material with a reference CGR curve. The analyzed CGR curve of the alternative Alloy 690 material was then newly estimated by applying the proven method over a practical range of stress intensity factors. Conclusion: Reliable CGR curves were obtained without complex environmental facilities or a high degree of experimental effort. The proposed method may be used to assess the PWSCC resistance of nuclear components subjected to high residual stresses such as those resulting from dissimilar metal welding parts.

• Journal of Industrial Convergence
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• v.18 no.4
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• pp.15-21
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• 2020

Because there are no isolation valves in condensate system of nuclear power plants, circulating water pump was shutdown for the condenser repair. When circulating water pump was shutdown, power plant output decreased about 45%. These output decreasing can minimize by establishing isolation valves. In this paper, evaluated effect to flow conditions change of condensate system, structural integrity of system, condenser pressure of in case of establish isolation valves to condensate system. Results of the evaluation, the flow rate due to the installation of the isolation valve decreased 0.3% when the valve was fully opened and 4.5% when fully closed. In addition, it was found that the vacuum degree of the condenser decreased with decreasing flow rate, but the integrity of the system was maintained.

• Journal of the Korean Society for Nondestructive Testing
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• v.26 no.1
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• pp.12-17
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• 2006

A steam generator of nuclear power plant has thousands of thin tubes. These tubes play an important role in maintaining the pressure boundary between the primary and secondary side of nuclear power plant. The steam generator tube is easy to be damaged because of the severe operating conditions such as the high temperature and pressure. Therefore, tremendous efforts are made to assess the structural integrity of the steam generator tubes. The eddy current test is the most popular non-destructive test to assess the integrity of the tubes. However, the eddy current test has the limitation to size the flaw accurately because the eddy current signal behavior depends on the total volume of flaw. This paper shows the possibility that the ultrasonic test method can be applied to detect the flaws in the steam generator tubes and to measure them quantitatively. From the test results, it is expected that if the ultrasonic test is put to practical use in the steam generator tube inspection, the inspection results will be improved.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.41 no.2
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• pp.157-164
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• 2017

A parallel slide gate valve (PSGV) is located between the heat recovery steam generator (HRSG) and the steam turbine in a combined cycle power plant (CCPP). It is used to control the flow of steam and runs with repetitive operations such as startups, load changes, and shutdowns during its operation period. Therefore, it is necessary to evaluate the fatigue damage and the structural integrity under a large compressive thermal stress due to the temperature difference through the valve wall thickness during the startup operations. In this paper, the thermal-structural analysis and the fatigue life evaluation of a 16-inch PSGV, which is installed on the HP steam line, is performed according to the fatigue life assessment method described in the ASME B&PVC VIII-2; the method uses the equivalent stress from the elastic stress analysis.

• Journal of the Society of Disaster Information
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• v.8 no.2
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• pp.188-196
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• 2012

Dynamic parameters such as natural frequencies can provide global stiffness information of a structure, and thus be utilized in monitoring structural integrity of large structures such as a containment. To identify the dynamic parameters without interrupting normal operation, a modal analysis method based on ambient vibration measurements should be applied. In this study, dynamic parameters of the containment of Wolsong Unit 2 are identified using ambient vibration measurement data. The feasibility of the study is verified using a numerical model for the containment. From the modal analysis, dynamic parameters of the containment with acceptable correlation to analytical modes can be estimated.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.5 no.2
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• pp.155-169
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• 2007

A transport container for a 500 kCi tritium storage vessel was developed, which could be used for the transport of metal tritide from Wolsong TRF facility to a disposal site. The structural, thermal, shielding, and confinement analyses were performed for the container in a view of Type B. As a result of structural analysis, the developed container sustained its integrity under normal and accidental conditions. The maximum temperature increase of the inner storage vessel by radiation was evaluated at $134.8^{\circ}C at room temperature. In $800^{\circ}C$ fire test, The thermal barrier of container sustained the inner vessel at $405^{\circ}C after 30 min, which temperature was allowable for the container integrity since maximum design temperature of inner vessel was $550^{\circ}C. In the evaluation of the shielding, the activity of radiation was nearly zero on the outer surface of inner vessel. Consequently the transport container for a 500 kCi tritium was evaluated to pass all the safety tests including accidental condition, so it was concluded that the designed transport container is proper to be used.