• Proceedings of the Korean Society for Quality Management Conference
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• 2010.04a
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• pp.254-261
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• 2010

The increase in hazard level is attributed to the industrial hazard environment; complete national environmental hazards to human health include climate change. The damage level in Korea from 1993 to 2009 has exceeded the Increase In adverse environmental conditions. Priority areas of concern will include those risks that are most likely to occur and are expensive when they do take place such as accident or injury at a community pool. Therefore, in this paper, we suggest the System Engineering method for application to the railway RAMS. Recently, the requirement of high-integrity level of infrastructure has been deemed important. The systems level approach is defined through the assessment of the RAMS interactions between elements of complex system applications.

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.26 no.5
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• pp.594-601
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• 2016

Mass estimation analysis of loose-parts in pressure vessel is necessary for the structural integrity assessment of pressure boundary in nuclear power plants. Mass of loose-parts can be generally estimated from the peak values and the center frequency of impact signals. Magnitude of impact signals is, however, inevitably attenuated according to the traveling distance of the signals and depending on the frequencies. Attenuation rate must be therefore carefully compensated for the precise estimation of loose-part mass. This paper proposes a new compensation method for the attenuation rate based on Bessel function instead of Hankel function in conventional method which has a limitation of usage in near the impact location. It was verified that the suggested compensating equation based on the Bessel function can be applied to the attenuation rate calculation without any limitation.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.28 no.3
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• pp.311-317
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• 2004

This paper proposes engineering estimation equations for the maximum deflection of a cylinder subject to bending under elastic-plastic and elastic-creep conditions. Being based on the reference stress approach, the proposed equations are simple to use and can accommodate general tensile and creep behaviours. Validation against detailed 3-D FE results using actual stress-strain data and realistic creep-deformation data shows excellent agreement, which provides confidence in the use of the proposed equation. Based on the proposed equations, together with information on in-service inspection data, discussion is given how to estimate future time-dependent and time-independent deflection of the CANDU pressure tube. Thus the present result would be valuable information for integrity assessment of the CANDU pressure tube.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.430-434
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• 2001

To perform the integrity evaluation of RPV more realistically, it is necessary to evaluate the metallurgical microstructure and residual stress considering more real phenomena such as multi-pass welding process and PWHT. Accordingly, firstly, this paper proposes the integrated assessment methodology systematically developed for residual stress on weldment of RPV by using thermodynamics, diffusion theory, finite element method and validation experiment. Also, the residual stress on circumferential weldment of reactor pressure vessel is calculated considering multi-pass welding process by the commercial finite element package, ABAQUS.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2011.10a
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• pp.221-226
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• 2011

The structural integrity of APR1400 reactor vessel internals has been being assessed referring the US Nuclear Regulatory Commission regulatory guide 1.20 comprehensive vibration assessment program. The program is composed of a vibration and stress analysis, a limited vibration measurement, and an inspection. This paper covers the vibration and stress analysis on the reactor vessel internals due to the pulsation of reactor coolant pump. 3-dimensional models to calculate the hydraulic loads and structural responses were built and the pressure distributions and the structural responses were predicted using ANSYS. The peak stress of the reactor vessel internals is much lower than the acceptance limit.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2011.10a
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• pp.212-218
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• 2011

APR1400 RVI CVAP using the non-prototype category is being conducted to verify integrity of the RVI design and to secure the CVAP technology. The measurement programs are to confirm vibration analysis results for reactor internals during preoperational and initial startup testing and to detemine the safety margin. One of the important basis for the measurement programs is test acceptance criteria. Therefore, this paper is on establishment of response instrumentation test acceptance criteria for APR1400 RVI CVAP. The established acceptance criteria show that the stress criteria of APR1400 RVI are more conservative values than those of the valid prototype plant(Palo Verde unit 1) and, the displacement criteria of the IBA and the UGS were established to 0.03 in and 0.01 in, respectively.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2012.10a
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• pp.306-311
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• 2012

A reactor vessel internals comprehensive vibration assessment program (RVI CVAP) of an advanced power reactor 1400 (APR1400) is being performed as a non-prototype category-2 type of reactor based on the US Nuclear Regulatory Commission Regulatory Guide (NRC RG) 1.20. The aim of this paper is to present the results of structural response analysis and measuring locations of a upper guide structure (UGS) assembly of the APR1400 reactor. The analysis results of the UGS assembly results show that meet the specified integrity levels of the design acceptance criteria. Also, the measuring locations are set by the analysis results of the UGS assembly and selection criteria of measuring locations prior to this study. These analysis results and measuring locations will be used as fundamental materials to design a measurement system for the APR1400 RVI CVAP.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05d
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• pp.283-288
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• 1996

Core support barrel (CSB) assembly is one of the most important reactor internals structures supporting and protecting the nuclear core during normal operation and faulted events. For Yonggwang 3 and 4 (YGN 3&4), the adequacy of the analytical response prediction of reactor internals for flow induced vibration was demonstrated through the comprehensive vibration assessment program (CVAP) performed during hot functional test. Besides, the vibration characteristics of the CSB of operating nuclear power plant can be examined via the excore neutron noise monitoring signal. In this paper data from YGN 4 analyses, CVAP, and neutron noise monitoring system are compared and evaluated. In general, the results are comparable each other and conservative enough to ensure sufficient design margin and structural integrity. Further investigations on the modelling and analyses procedure are recommended to utilize the experimental results to the maximum extent. And collection of the neutron noise data is desired to serve as a baseline information.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.2
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• pp.1-8
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• 2010

Nuclear steam generator tubes have experienced wear degradation at tube support structure. Morphology of wear scar was analyzed by using eddy current signal. A burst test facility for steam generator tubes was established and tubes with 3 types of defects were tested. The burst test results show that the depth of wear scar is the main factor influencing the burst pressure of tubes, meanwhile, both the longitudinal length and the angle also have effect on the burst pressure. Based on test results, the burst pressure equation for wear degradation was proposed.

• Proceedings of the KSR Conference
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• 2001.05a
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• pp.230-237
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• 2001

For the structural integrity of large and complex structures such as railway vehicle, the in-field diagnosis of mechanical properties of the structures is needed, and especially, the mechanical characteristics of the weldment must be carefully evaluated. But, conventional standard testing methods having destructive procedures are not applicable to in-field assessment of mechanical property variations within weldment because they needs the limitations of specimen size and geometry. In this paper, to overcome this problems, the advanced indentation technique (AIS) is introduced for simple and non-destructive/in-field testing of weldment of industrial structures. This test measures indentation load-depth curve during indentation and analyzes the mechanical properties related to deformation and fracture. First of all, flow properties such as yield strength, tensile strength and work hardening index can be evaluated through the analysis of the deformation behavior beneath the spherical indenter. Additionally, case studies of advanced indentation techniques are introduced.