• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회A
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• pp.1803-1806
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• 2007

The General roles of a spacer grid(SG) are providing a lateral and vertical support for fuel rods, promoting a mixing of coolant and keeping guide tubes straight so as not to impede a control rod insertion under any normal or accidental conditions. To evaluate the impact characteristics of a SG such as impact velocity, critical buckling strength and duration time, a few types of impact tests for SGs have been conducted. In a previous study, a new welding method, a through-welding method, was proposed to increase critical buckling strength of a SG without any design change or material change and was verified by impact tests with $7{\times}7$ partial SG specimens.In this paper, the effect of through-welding method in case of a $16{\times}16$ full-size SG is investigated by pendulum impact tests with $16{\times}16$ SG specimens. And the increase of critical buckling strength for full-size SGs is measured by comparison with impact results of spot-welded and through-welded SGs.

• Steel and Composite Structures
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• 제8권3호
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• pp.179-200
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• 2008

This paper reports the results of a recent experimental study into the behavior of welded fin-plate connections to both hollow and concrete filled tubular (CFT) columns under shear. Experiments have been performed at both ambient and elevated temperatures with the aid of an electric kiln. The observed failure modes include fracture of the fin plate and tearing out of the tube around the welds. By considering the results of previously published research, the current design method for similar connections under purely tensile load, in CIDECT Guide 9, based on a deformation limit of 3% of the tube width is shown to be inadequate when evaluating the ultimate strength of such connections. By comparing the results from the current test program which failed in the fin-plate with Eurocode guidance for failure of a fin-plate alone under shear and bending load it is shown that the column face influences the overall connection strength regardless of failure mode. Concrete in-fill is observed to significantly increase the strength of connections over empty specimens, and circular column specimens were observed to exhibit greater strength than similarly proportioned square columns. A finite element (F.E.) model, developed using ABAQUS, is presented and validated against the experimental results in order that extensive parametric tests may be subsequently performed. When validating the model against elevated temperature tests it was found that using reduction factors suggested in published research for the specific steel grades improved results over applying the generic Eurocode elevated temperature steel strength reduction factors.

• 국제초고층학회논문집
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• 제3권3호
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• pp.215-221
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• 2014

Ultra-high strength concrete and high tensile steel are becoming very attractive materials for high-rise buildings because of the need to reduce member size and structural self-weight. However, limited test data and design guidelines are available to support the applications of high strength materials for building constructions. This paper presents significant findings from comprehensive experimental investigations on the behaviour of tubular columns in-filled with ultra-high strength concrete at ambient and elevated temperatures. A series of tests was conducted to investigate the basic mechanical properties of the high strength materials, and structural behaviour of stub columns under concentric compression, beams under moment and slender beam-columns under concentric and eccentric compression. High tensile steel with yield strength up to 780 MPa and ultra-high strength concrete with compressive cylinder strength up to 180 MPa were used to construct the test specimens. The test results were compared with the predictions using a modified Eurocode 4 approach. In addition, more than 2000 test data samples collected from literature on concrete filled steel tubes with normal and high strength materials were also analysed to formulate the design guide for implementation in practice.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2005년도 학술대회 논문집 정보 및 제어부문
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• pp.225-227
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• 2005

The in-core neutron Flux Mapping System in a pressurized water reactor yields information on the neutron flux distribution in the reactor core at selected core locations by means of movable detectors. The obtained data are used to verify the reactor core design parameters. The detector cables run through guide tubes(thimbles), and typically thirty-six to fifty-eight thimbles are allocated in the reactor depending on the number of fuel assemblies. These thimbles are inserted into nuclear fuel assemblies through conduits connected from the bottom of the reactor vessel to a seal table. During the plant refueling outage period, the thimbles are withdrawn up to 4m from the seal table, the height of a nuclear fuel. In spite of their importance, however, the thimble handling work has been performed by only human operators. In addition, its efficiency is very low due to narrow working environments on the seal table, thereby resulting in the excessive radiation exposure of maintenance personnel. To solve these problems, a new thimble handling equipment for in-core flux mapping system was developed, and we confirmed its effectiveness through experiments.

• International Journal of Fluid Machinery and Systems
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• 제2권4호
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• pp.375-382
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• 2009

Hydraulic instability associated with pressure fluctuations is a serious problem in hydraulic machinery. Pressure fluctuations are usually a result of a strong vortex created in the centre of a flow at the outlet of a runner. At every radial turbine and also at every single regulating axial turbine, the draft tube vortex appears at part-load operating regimes. The consequences of the vortex developed in the draft tube are very unpleasant pressure pulsation, axial and radial forces and torque fluctuation as well as turbine structure vibration. The consequences of the vortex are transferred upstream and downstream with amplitude and frequency modulation in respect of the turbine operating regime, cavitation conditions and air admitted content. Numerical prediction of the vortex appearance in the design stage is a very important task. The amplitude of the pressure pulsation is different for each operating regime therefore the main goal of this research was to numerically predict pressure pulsation amplitude versus different guide vane openings and to compare the results with experimental ones. For the numerical flow analysis of a complete Francis turbine (FT), the computer code ANSYS-CFX11 has been used.

• 한국산학기술학회논문지
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• 제12권1호
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• pp.7-14
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• 2011

연질 동관은 조선, 자동차, 플랜드, 냉동공조 기기 등에 많이 사용되는 대표적인 파이프 재료 중의 하나이다. 그러나 이러한 재료들은 벤딩가공시 외부의 주름, 스프링백, 관이 얇아지는 현상들이 자주 발생하여 문제가 된다. 본 논문에서는 연성재질을 벤딩시 맨드럴이 슬라이딩 가이드 레일을 따라서 파이프를 누르는 벤딩방식을 선택한 연성재질 전용 벤딩머시인을 제작하여 이와 같은 현상을 줄였다. 이의 성능을 확인하는 과정에서 재료로 사용한 동관의 직경이 작을수록 스프링백 현상은 증가하였고 또한 굽힘각도가 클수록 컸다. 또한 동관의 벤딩시 주름이 거의 발생하지 않는 성형품을 제작할 수 있었다.

• Nuclear Engineering and Technology
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• 제50권2호
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• pp.292-296
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• 2018

Recent events in the nuclear industry have resulted in a movement towards increased seismic and LOCA excitations and requirements that challenge current fuel designs. AREVA NP's GAIA fuel design introduces unique and robust characteristics to resist the effects of seismic and LOCA excitations. For demanding seismic and LOCA scenarios, fuel assembly spacer grids can undergo plastic deformations. These plastic deformations must not prohibit the complete insertion of the control rod assemblies and the cooling of the fuel rods after the accident. The specific structure of the GAIA spacer grid produces a unique and stable compressive deformation mode which maintains the regular array of the fuel rods and guide tubes. The stability of the spacer grid allows it to absorb a significant amount of energy without a loss of load-carrying capacity. The GAIA-specific grid behavior is in contrast to the typical spacer grid, which is characterized by a buckling instability. The increased mechanical robustness of the GAIA spacer grid is advantageous in meeting the increased seismic and LOCA loadings and the associated safety requirements. The unique GAIA spacer grid behavior will be incorporated into AREVA NP's licensed methodologies to take full benefit of the increased mechanical robustness.

• Nuclear Engineering and Technology
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• 제50권2호
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• pp.297-305
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• 2018

Fuel assembly (FA) bow in pressurized water reactor (PWR) cores is considered to be a complex process with a large number of influencing mechanisms and several unknowns. Uncertainty and sensitivity analyses are a common way to assess the predictability of such complex phenomena. To perform such analyses, a structural model of a row of 15 FAs in the reactor core is implemented with the finite-element code ANSYS Mechanical APDL. The distribution of lateral hydraulic forces within the core row is estimated based on a two-dimensional Computational Fluid Dynamics model with porous media, assuming symmetric or asymmetric core inlet and outlet flow profiles. The influence of the creep rate on the bow amplitude is tested based on different creep models for guide tubes and fuel rods. Different FA initial states are considered: fresh FAs or FAs with higher burnup, which may be initially straight or exhibit an initial bow from previous cycles. The simulation results over one reactor cycle demonstrate that changes in the creep rate and the hydraulic conditions may have a considerable impact on the bow amplitudes and the bow patterns. A good knowledge of the specific creep behavior and the hydraulic conditions is therefore crucial for making reliable predictions.

• Journal of Welding and Joining
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• 제26권5호
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• pp.66-73
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• 2008

Welding work in pre-erection or erection stage of shipbuilding construction to be carried out in flat and vertical upward position mostly and Electrogas welding(EGW) is actively applied especially for vertical butt joint of thicker steel plate recently. In this study considered how to develope and improve mechanical properties of weld metal and HAZ in high heat input welding processes such as EGW and Electroslag welding(ESW) with its welding equipment in order to extend the application range to the longitudinal members and hatch coaming parts of container ship. Some components of welding system and parameters were modified to get the faster travel speed and reduce weld heat input, and also by adding additional filler rods or tubes increase the amount of deposited weld metal. With the test get some good date can apply to actual fabrication work and recommend items to manufacture welding materials make better. Above all things it's a fruition that to prepare the possibility of application of ESW to shipbuilding construction which fill up the gap of stoppage days of more than 20 years.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3264-3274
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• 2021

In a severe accident of light water reactor (LWR), molten core material (corium) can be released into the wet cavity, and a fuel-coolant interaction (FCI) can occur. The molten jet with high speed is broken and fragmented into small debris, which may cause a steam explosion or a molten core concrete interaction (MCCI). Since the premixing stage where the jet breakup occurs has a large impact on the severe accident progression, the understanding and evaluation of the jet breakup phenomenon are highly important. Therefore, in this study, the jet breakup simulations were performed using the Smoothed Particle Hydrodynamics (SPH) method which is a particle-based Lagrangian numerical method. For the multi-fluid system, the normalized density approach and improved surface tension model (CSF) were applied to the in-house SPH code (single GPU-based SOPHIA code) to improve the calculation accuracy at the interface of fluids. The jet breakup simulations were conducted in two cases: (1) jet breakup without structures, and (2) jet breakup with structures (control rod guide tubes). The penetration depth of the jet and jet breakup length were compared with those of the reference experiments, and these SPH simulation results are qualitatively and quantitatively consistent with the experiments.