• Journal of the Korean Society of Manufacturing Technology Engineers
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• v.24 no.3
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• pp.263-269
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• 2015

Growing concerns regarding air pollution have recently increased the demand for liquefied natural gas (LNG) fueled ships. LNG-fueled ships are equipped with an explosion relief valve in the crankcase to relieve excessive pressure and stop flames from emitting from the crankcase. In this study, a finite element analysis was conducted to evaluate the crankcase relief valve disk spring design using an ANSYS Workbench, v.15. The setting pressure, leak and explosion test performed by european standard EN14797 to evaluate function and mechanical integrity of crankcase relief valve. The tests results indicate that the pressure of the crankcase relief valve is 3.05 bar, with no air leakage at 2.97 bar. Finally, the mechanical integrity of the crankcase relief valve was confirmed through an explosion test in which the valve plate assembly, flame arrester, and other parts were safe from fracturing.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.1
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• pp.179-187
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• 2002

Finite element analysis was performed to evaluate the integrity of the tubes in the fixed tubesheet of horizontal type heat exchanger under operating condition. For the finite element analysis of the heat exchanger, tubes and tubesheets were equivalently modeled with concentroidal hexagonal columns and solid plates having equivalent properties for the convenience of finite element modeling, respectively. Load combination of tube pressure and thermal expansion most likely to precipitate possible failure of the tubes was selected and applied to the finite element analysis. The compressive stresses of the tubes were calculated based on displacements of each tube, which were obtained from anile element analysis. Finally, the maximum tube stress was compared with the design criterion of ASME Boiler and Pressure Vessel Code Section VIII.

• Journal of Mechanical Science and Technology
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• v.15 no.1
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• pp.10-20
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• 2001

The leak before break(LBB) concept is difficult to apply to a structure with a thin tube that is immersed in a water environment. A heat exchanger in a nuclear power plant is such a structure. The present paper addresses an application of the LBB concept to a heat exchanger in a nuclear power plant. The minimum leaked coolant amount(approximately 37.9 liters) containing the radioactive material which can activate the radiation detector device installed in near the heat exchanger is assumed. A postulated initial flaw size that can not grow to a critical flaw size within the time period to activate the radiation detector is justified. In this case, the radiation detector can activate the warning signal caused by coolant leakage from initially postulated flaws of the heat exchanger. The nuclear plant can safely shutdown when this occurs. Since the postulated initial flaw size can not grow to the critical flaw size, the structural integrity of the heat exchanger is not impeded. Particularly the informational scenario presented in this paper discusses an actual nuclear plant.

• KSII Transactions on Internet and Information Systems (TIIS)
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• v.11 no.1
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• pp.1-17
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• 2017

Mobile Cloud Computing has become a promising computing platform. It moves users' data to the centralized large data centers for users' mobile devices to conveniently access. Since the data storage service may not be fully trusted, many public verification algorithms are proposed to check the data integrity. However, these algorithms hardly consider the huge computational burden for the verifiers with resource-constrained mobile devices to execute the verification tasks. We propose an energy-efficient algorithm for assigning verification tasks (EEAVT) to optimize the energy consumption and assign the verification tasks by elastic and customizable ways. The algorithm prioritizes verification tasks according to the expected finish time of the verification, and assigns the number of checked blocks referring to devices' residual energy and available operation time. Theoretical analysis and experiment evaluation show that our algorithm not only shortens the verification finish time, but also decreases energy consumption, thus improving the efficiency and reliability of the verification.

• Korean Journal of Hydrosciences
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• v.8
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• pp.31-40
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• 1997

Flow carrying capacity of water distribution systems is getting reduced by deterioration of pipes in the systems. The objective of this paper is to present a managerial decision-making model for the rehabilitation of water distribution systems with a mininum cost. The decisions made by the model also satisfy the requirements for discharge and pressure at demanding nodes in the systems. Replacement cost, pipe break repair cost, and pumping cost are considered in the economic evaluation of the decision along with the break rate and the interest rate to determine the optimal replacement time for each pipe. Then, the hydraulic integrity of the water distribution system is checked for the decision by a pipe network simulator, KYPIPE, if discharge and pressure requirements are satisfied. In case the system does not satisfy the hydraulic requirements, the decision made for the optimal replacement time is revised until the requirments are satisfied. The model is well applied to an existing water distribution system, the Seoul Metropolitan Water Supply System (1st Phase). The results show that the decisions for the replacement time determined by the economic analysis are accepted as optimal and hydraulic integrity of the system is in good condition.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.1
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• pp.17-26
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• 2011

CANDU reactor core is composed a few hundreds pressure tubes, which support and locate the nuclear fuels in the reactor. Each pressure tube provides pressure boundary and flow path of primary heat transport system in the core region. In order to guarantee the structural integrity of pressure tube flaws which can be found by in-service inspection, crack growth and fracture initiation assessment have to be performed. Stress intensity factors are important and basic information for structural integrity assessment of planar and laminar flaws (e. g. crack). This paper reviews and confirms the stress intensity factor of axial crack, proposed in CSA N285.8-05, which is an fitness-for-service evaluation code for pressure tubes in CANDU nuclear reactors. The stress intensity factors in CSA N285.8-05 were compared with stress intensity factors calculated by three methods (finite element results, API 579-1/ASME FFS-1 2007 Fitness-For-Service and ASME Boiler and Pressure Vessel Code Section XI). The effects of Poisson's ratio and anisotropic elastic modulus on stress intensity factors were also discussed.

• Korean Journal of Ecology and Environment
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• v.43 no.1
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• pp.24-34
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• 2010

This study was carried out to analyze some influences on ecological health conditions, threaten by various stressors such as physical, chemical and biological parameters. We collected samples in 2008 from three zones of upstream, midstream and downstream, Gap Stream. We applied multi-metric fish assessment index (MFAI), based on biotic integrity model to the three zones along with habitat evaluations based on Qualitative Habitat Evaluation Index (QHEI). We also examined fish fauna and compositions, and analyzed relations with MFAI values, QHEI values, and various guild types. Chemical parameters such as oragnic matter (BOD, COD), nutrients (TP, $NH_3$-N), coli-form number (as MPN), and suspended solids (SS) were analyzed to identify the relationship among multiple stressor effects. Using the sentinel species of Zacco platypus, the population structures and condition factors were analyzed along with DNA damages related with genotoxicant effects by comet assay. This study using all these parameters showed that stream condition was degraded along the longitudinal gradient from upstream to downstream, and the downstream, especially, was impacted by nutrient enrichment and toxicant exposure from the point source, wastewater treatment plant. Overall results indicated that our approaches applying various parameters may be used as a cause-effect technique in the stream health assessments and also used as a pre-warning tool for diagnosis of ecological degradation.

• Nuclear Engineering and Technology
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• v.40 no.7
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• pp.607-614
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• 2008

Since pipes with wall-thinning defects can collapse at fluid pressure that are lower than expected, the collapse moment of wall-thinned pipes should be determined accurately for the safety of nuclear power plants. Wall-thinning defects, which are mostly found in pipe bends and elbows, are mainly caused by flow-accelerated corrosion. This lowers the failure pressure, load-carrying capacity, deformation ability, and fatigue resistance of pipe bends and elbows. This paper offers a support vector regression (SVR) model further enhanced with a fuzzy algorithm for calculation of the collapse moment and for evaluating the integrity of wall-thinned piping systems. The fuzzy support vector regression (FSVR) model is applied to numerical data obtained from finite element analyses of piping systems with wall-thinning defects. In this paper, three FSVR models are developed, respectively, for three data sets divided into extrados, intrados, and crown defects corresponding to three different defect locations. It is known that FSVR models are sufficiently accurate for an integrity evaluation of piping systems from laser or ultrasonic measurements of wall-thinning defects.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2001.04a
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• pp.481-488
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• 2001

The HANARO in-chimney bracket was proposed as a structure which supports the guide tubes of irradiation facilities at the irradiation sites of CT, IR and OR4/5 in HANARO core for the reduction of flow-induced vibration and seismic response of the irradiation facilities. For the evaluation of the structural integrity of the in-chimney bracket, its finite element model is developed. The seismic response analysis was performed for the in-chimney bracket and related reactor structures, under the response spectrum of OBE and SSE. The analysis results show that stress values of the in-chimney bracket and reactor structures for the seismic loads are within the ASME code limits. It is also confirmed that its fatigue usage factor is much less than 1.0. For the verification of the implementation effects of the in-chimney bracket, the vibration level of the guide tube of the instrumented fuel assembly, which is subjected to fluid-induced vibration, was measured and analyzed. The vibration analysis results demonstrate that the vibration level of the instrumented fuel assembly has been remarkably reduced after installing the in-chimney bracket. Therefore, when the in-chimney bracket is installed at the reactor chimney, any damage on the structural integrity is not expected.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.30 no.10 s.253
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• pp.1255-1260
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• 2006

To maintain the structural integrity of steam generator tubes, 40% of wall thickness plugging criterion has been developed. The approach is for the steam generator tube with single crack, so that the interaction effect of multiple cracks can not be considered. Although, recently, several approaches have been proposed to assess the integrity of steam generator tube with two identical cracks whilst actual multiple cracks reveal more complex shape. In this paper, the coalescence pressure of steam generator tube containing multiple cracks of different length is evaluated based on the detailed 3-dimensional (3-D) elastic-plastic finite element (FE) analyses. In terms of the crack shape, two collinear axial through-wall cracks with different length were considered. Furthermore, the resulting FE coalescence pressures are compared with FE coalescence pressures and experimental results for two identical collinear axial through-wall cracks to quantify the effect of crack length ratio on failure behavior of steam generator tube with multiple cracks. Finally, based on 3-D FE results, the coalescence evaluation diagrams were proposed.