• Proceedings of KOSOMES biannual meeting
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• 2017.04a
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• pp.129-130
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• 2017

This paper suggests study of basic structure design and structural analysis for the twin car-ferries. The rules and methodology for the analysis of strength of medium and small high speed vessels with a length of more than 50m and a width / ratio of more than 12, such as car-ferries, have not been clarified yet. Therefore, in this paper, the scantling of the members is based on the Korea Classification standards, and the car-ferries standards were additionally applied to verify the structural strength of the design. The results of this study are expected to be useful as basic data related to structural design and structural analysis of high speed twin car-ferries.

• Structural Engineering and Mechanics
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• v.52 no.3
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• pp.613-632
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• 2014

In the early design stage of ships, the two most important structural analyses are performed to identify the structural capacity and safety. The first step is called global strength analysis (longitudinal strength analysis or hull girder strength analysis) and the second step is local buckling analysis (stiffened panel strength analysis). This paper deals with the ultimate strength performance of Arctic Sea Route-going commercial ships considering the effect of low temperature. In this study, two types of structural analyses are performed in Arctic sea conditions. Three types of ship namely oil tanker, bulk carrier and container ship with four different sizes (in total 12 vessels) are tested in four low temperatures (-20, -40, -60 and $-800^{\circ}C$), which are based on the Arctic environment and room temperature ($20^{\circ}C$). The ultimate strength performance is analysed with ALPS/HULL progressive hull collapse analysis code for ship hulls, then ALPS/ULSAP supersize finite element method for stiffened panels. The obtained results are summarised in terms of temperature, vessel type, vessel size, loading type and other effects. The important insights and outcomes are documented.

• Geomechanics and Engineering
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• v.5 no.4
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• pp.363-377
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• 2013

Any structure constructed on the earth is supported by the underlying soil. Foundation is an interfacing element between superstructure and the underlying soil that transmits the loads supported by the foundation including its self weight. Foundation design requires evaluation of safe bearing capacity along with both immediate and long term settlements. Weak and compressible soils are subjected to problems related to bearing capacity and settlement. The conventional method of design of footing requires sufficient safety against failure and the settlement must be kept within the allowable limit. These requirements are dependent on the bearing capacity of soil. Thus, the estimation of load carrying capacity of footing is the most important step in the design of foundation. A number of theoretical approaches, in-situ tests and laboratory model tests are available to find out the bearing capacity of footings. The reliability of any theory can be demonstrated by comparing it with the experimental results. Results from laboratory model tests on square footings resting on sand are presented in this paper. The variation of bearing capacity of sand below a model plate footing of square shape with variation in size, depth and the effect of permissible settlement are evaluated. A steel tank of size $900mm{\times}1200mm{\times}1000mm$ is used for conducting model tests. Bearing capacity factor $N_{\gamma}$ is evaluated and is compared with Terzaghi, Meyerhof, Hansen and Vesic's $N_{\gamma}$ values. From the experimental investigations it is found that, as the depth of sand cushion below the footing ($D_{sc}$) increases, ultimate bearing capacity and settlement values show an increasing trend up to a certain depth of sand cushion.

• Steel and Composite Structures
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• v.22 no.3
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• pp.537-565
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• 2016

Australian railway networks possess a large amount of aging timber components and need to replace them in excess of 280 thousands $m^3$ per year. The relatively high turnover of timber sleepers (crossties in a plain track), bearers (skeleton ties in a turnout), and transoms (bridge cross beams) is responsible for producing greenhouse gas emissions 6 times greater than an equivalent reinforced concrete counterparts. This paper presents an innovative solution for the replacement of aging timber transoms installed on existing railway bridges along with the incorporation of a continuous walkway platform, which is proven to provide environmental, safety and financial benefits. Recent developments for alternative composite materials to replace timber components in railway infrastructure construction and maintenance demonstrate some compatibility issues with track stiffness as well as structural and geometrical track systems. Structural concrete are generally used for new railway bridges where the comparatively thicker and heavier fixed slab track systems can be accommodated. This study firstly demonstrates a novel and resilient alterative by incorporating steel-concrete composite slab theory and combines the capabilities of being precast and modulated, in order to reduce the depth, weight and required installation time relative to conventional concrete direct-fixation track slab systems. Clear benefits of the new steel-concrete composites are the maintainability and constructability, especially for existing railway bridges (or brown fields). Critical considerations in the design and finite element modelling for performance benchmarking of composite structures and their failure modes are highlighted in this paper, altogether with risks, compatibilities and compliances.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.31 no.2 s.257
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• pp.277-284
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• 2007

The steam generator in a nuclear power plant is a large heat exchanger that uses heat from a reactor to generate steam to drive the turbine generator. Rupture of a steam generator tube can result in release of fission products to environment outside. Therefore, an accurate integrity assessment of the steam generator tubes with cracks is of great importance for maintaining the safety of a nuclear power plant. The steam generator tubes are supported at regular intervals by support plates and rotations of the tubes are restrained. Although it has been reported that the limit load for a circumferential crack is significantly affected by boundary condition of the tube, existing limit load solutions do not consider the restraining effect of support plate correctly. In addition, there are no limit load solutions for circumferential cracks in U-bend region with the effect of the support plate. This paper provides detailed limit load solutions for circumferential cracks in top of tube sheet and the U-bend regions of the steam generator tube with the actual boundary conditions to simulate the restraining effect of the support plate. Such solutions are developed based on three dimensional finite element analyses. The resulting limit load solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes.

• Journal of Digital Convergence
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• v.15 no.7
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• pp.223-229
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• 2017

Europe, the US and Japan have acquired test results on impact coefficient for a long time and applied it to equipment design to secure safety of structures. However, Korean enterprises use the impact factor held by advanced business to design equipment as it is difficult for them to obtain it through tests. In this paper, NX/NASTRAN, was used to perform static load analysis and impact load analysis of a PCB Handler, semiconductor test equipment, and the result was employed to study how to calculate the impact coefficient with the finite element analysis. The calculation method was applied to the JIS(Japanese Industrial Standard), and the impact coefficient of the PCB handler was calculated as 1.27 for the sudden start or stop. The impact coefficient generated by the analysis is expected to make a great contribution to the industry as it can be used to improve the equipment structure and develop on existing equipment in the future.

• Journal of the Earthquake Engineering Society of Korea
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• v.19 no.1
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• pp.29-36
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• 2015

Base isolation is considered as a seismic protective system in the design of next generation Nuclear Power Plants (NPPs). If seismic isolation devices are installed in nuclear power plants then the safety under a seismic load of the power plant may be improved. However, with respect to some equipment, seismic risk may increase because displacement may become greater than before the installation of a seismic isolation device. Therefore, it is estimated to be necessary to select equipment in which the seismic risk increases due to an increase in the displacement by the installation of a seismic isolation device, and to perform research on the seismic performance of each piece of equipment. In this study, modified NRC-BNL benchmark models were used for seismic analysis. The numerical models include representations of isolation devices. In order to validate the numerical piping system model and to define the failure mode, a quasi-static loading test was conducted on the piping components before the analysis procedures. The fragility analysis was performed by using the results of the inelastic seismic response analysis. Inelastic seismic response analysis was carried out by using the shell finite element model of a piping system considering internal pressure. The implicit method was used for the direct integration time history analysis. In addition, the collapse load point was used for the failure mode for the fragility analysis.

• Corrosion Science and Technology
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• v.3 no.5
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• pp.216-221
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• 2004

The operating experience showed that the fatigue is one of the major piping failure mechanisms in nuclear power plants (NPPs). The pressure and/or temperature loading transients, the vibration, and the mechanical cyclic loading during the plant operation may induce the fatigue failure in the nuclear piping. Recently, many fatigue piping failure occurred at the socket weld area have been widely reported. Many failure cases showed that the gap requirement between the pipe and fitting in the socket weld was not satisfied though the ASME Code Sec. III requires 1/16 inch gap in the socket weld. The ASME Code OM also limits the vibration level of the piping system, but some failure cases showed the limitation was not satisfied during the plant operation. In this paper, the fatigue behavior of the socket weld in the nuclear piping was estimated by using the three dimensional finite element method. The results are as follows. (1) The socket weld is susceptible to the vibration if the vibration levels exceed the requirement in the ASME Code OM. (2) The effect of the pressure or temperature transient load on the socket weld in NPPs is not significant because of the very low frequency of the transient during the plant lifetime operation. (3) 'No gap' is very risky to the socket weld integrity for the specific systems having the vibration condition to exceed the requirement in the ASME OM Code and/or the transient loading condition. (4) The reduction of the weld leg size from $1.09*t_1$ to $0.75*t_1$ can affect severely on the socket weld integrity.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.166-177
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• 1993

A simplified thermal analysis method to evaluate the maximum temperature of the CANDU 37-element fuel bundle within a fuel basket in a given spent fuel dry storage canister has been presented along with the results of sample analyses performed to examine the parametric effects of the ambient conditions on the maximum fuel temperature within a canister. To solve the multi-dimensional heat transfer problem of the complex geometry of rod bundles within a canister where three modes of heat transfer are superimposed, the CANDU spent fuel bundles stored in the dry storage canister are first replaced by equivalent concentric fuel cylinders. The simplified axi-symmetric two-dimensional multi-mode heat transfer problem of the equivalent fuel cylinders is then analyzed with an existing computer code, HEATING5, using additional input data and heat transfer correlations. A comparison between the predicted temperature profile and the mock-up test results shows that the agreement is quite satisfactory.

• Nuclear Engineering and Technology
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• v.47 no.6
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• pp.712-718
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• 2015

Cast duplex stainless steels are susceptible to thermal aging during long-term service at temperatures ranging from $280^{\circ}C$ to $450^{\circ}C$. To analyze the effect of thermal aging on leak-before-break (LBB) behavior, three-dimensional finite element analysis models were built for circumferentially cracked pipes. Based on the elasticeplastic fracture mechanics theory, the detectable leakage crack length calculation and J-integral stability assessment diagram approach were carried out under different bending moments. The LBB curves and LBB assessment diagrams for unaged and thermally aged pipes were constructed. The results show that the detectable leakage crack length for thermally aged pipes increases with increasing bending moments, whereas the critical crack length decreases. The ligament instability line and critical crack length line for thermally aged pipes move downward and to the left, respectively, and unsafe LBB assessment results will be produced if thermal aging is not considered. If the applied bending moment is increased, the degree of safety decreases in the LBB assessment.