• Journal of the korean Society of Automotive Engineers
• /
• v.14 no.6
• /
• pp.99-104
• /
• 1992

The complex geometry of the engine room of a passenger car has been modelled two-dimensionally and the thermal and fluid flow therein have been analyzed by using a commercially available code, PATRAN/FLORAM$\mid$N. FLOTRAN adopts a finite element method with streamline upwind formulation for convective terms and the k-.epsilon. turbulence model to solve the three dimensional turbulent flow and heat transfer problems. Velocity vectors, pressure and temperature distributions have been obtained for various cases with different arrangements of license plate, underbody-covers and air dams. The results show that the numerical analysis using PATRAN/FLOTRAN can predict qualitatively well the practical phenomena.

• Journal of the Korean Society of Visualization
• /
• v.8 no.3
• /
• pp.49-55
• /
• 2010

Thermal flows inside a laser printer are affected by generated heat from a fuser and boards. Thus, the effect of fans has been investigated to control the thermal flows and behaviors of toners. In order to analyze the phenomena experimentally, a PIV (Particle Image Velocimetry) has been used, and then the flow inside the printer has been predicted by the CFD (Computational Fluid Dynamics) in this study to determine the efficient flow distribution by an optimum design of the printer. The determined optimum design has been confirmed by the developed PIV technique so that the efficiency of the laser printer can be improved by the proposed design.

• International Journal of Aeronautical and Space Sciences
• /
• v.1 no.1
• /
• pp.62-69
• /
• 2000

Present study deals with performance analysis of an inert gas generator (IGG) which can be used as effective means to suppress fire. The IGG uses a turbo-jet engine to generate inert gas for fire extinguishing. It is generally known that a less degree of oxygen content in the product of combustion will increase the effectiveness of fire extinguishing. An inert gas generator system with water injection has advantages of suffocating and cooling effects that are very important factors for fire extinguishing. Some aspects of influencing parameters, such as, air excess coefficient, compressor pressure ratio, air temperature before combustion chamber, gas temperature after combustion chamber, mass flow rate of water injection etc. on the performance of IGG system are investigated.

• Journal of the Korean Solar Energy Society
• /
• v.21 no.3
• /
• pp.61-68
• /
• 2001

This study represents numerical analysis on the thermal and fluid flow of the air layer in a solar collector. The boundary conditions was assumed that the top and bottom wall of the air layer have a heating and cooling surface, respectively, and this calculation model have a solid body with a cooling temperature of $20^{\circ}C$. As the results of simulations, the magnitudes of the velocity vectors and isotherms are increased proportionally to the tilt angles. As the tilt angle is increased, the mean Nusselt numbers are increased and the maximum value of the mean Nusselt number was appeared at tilt angle $\theta=75^{\circ}$.

• Proceedings of the KSME Conference
• /
• 2001.06d
• /
• pp.110-114
• /
• 2001

A numerical analysis has been perfonned to estimate the effect of turbulent penetration and thermal stratified flow in the branch lines piping. This phenomenon of thermal stratification are usually observed in the piping lines of the safety related systems and may be identified as the source of fatigue in the piping system due to the thermal stress loading which are associated with plant operating modes. The turbulent penetration length reaches to $1^{st}$ valve in safety injection piping from reactor coolant system (RCS) at normal operation for nuclear power plant when a coolant does not leak out through valve. At the time, therefore, the thermal stratification does not appear in the piping between RCS piping and $1^{st}$ valve of safety injection piping. When a coolant leak out through the $1^{st}$ valve by any damage, however, the thermal stratification can occur in the safety injection piping. At that time, the maximum temperature difference of fluid between top and bottom in the piping is estimated about $50^{\circ}C$.

• Nuclear Engineering and Technology
• /
• v.48 no.2
• /
• pp.376-385
• /
• 2016

A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and $60^{\circ}C$ (equivalent to Re ~ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

• Nuclear Engineering and Technology
• /
• v.53 no.9
• /
• pp.3112-3121
• /
• 2021

One of the major factors affecting the life span of a Reactor Pressure Vessel (RPV) is the Pressurised Thermal Shock (PTS). PTS is a thermo-mechanical load on the RPV wall due to steep temperature gradients and structural load created by internal pressure of the fluid within the RPV. Safe operating life of a nuclear power plant is ensured by carrying out fracture analysis of the RPV against thermal shock. Carrying out fracture tests on RPV/large scale components is not always feasible. Hence, studies on laboratory level specimens are necessary to validate and supplement the prototype results. This paper aims to study the fracture behaviour of standard Compact Tension [C(T)] specimens, made of RPV steel 20MnMoNi55, subjected to thermal shock through experimental and numerical investigations. Fracture tests have been carried out on the C(T) specimens subjected to thermal transient load and tensile load to quantify the effect of thermal shock. Crack resistance curves are obtained from the fracture tests as per ASTM E1820 and compared with those obtained numerically using XFEM and a good agreement was found. A quantitative study on the crack tip plastic zone, computed using cohesive segment approach, from the numerical analyses justified the experimental crack initiation toughness.

• International Journal of Aeronautical and Space Sciences
• /
• v.18 no.1
• /
• pp.70-81
• /
• 2017

Various components of an engine nozzle are modeled as flexible multi-body components that are operated under high temperature and pressure. In this paper, in order to predict complex behavior of an engine nozzle, thermo-fluid-flexible multi-body dynamics coupled analysis framework was developed. Temperature and pressure on the nozzle wall were obtained by the steady-state flow analysis for a two-dimensional nozzle. The pressure and temperature-dependent material properties were delivered to the flexible multi-body dynamics analysis. Then the deflection and strain distribution for a nozzle configuration was obtained. Heat conduction and thermal analyses were done using MSC.NASTRAN. The present framework was validated for a simple nozzle configuration by using a one-way coupled analysis. A two-way coupled analysis was also performed for the simple nozzle with an arbitrary joint clearance, and an asymmetric flow was observed. Finally, the total strain result for a realistic nozzle configuration was obtained using the one-way and two-way coupled analyses.

• Transactions of the Korean Society for Noise and Vibration Engineering
• /
• v.12 no.4
• /
• pp.261-271
• /
• 2002

The fluid-elastic instability analysis of the U-tube bundle inside the steam generator is very important not only for detailed design stage of the SG but also for the change of operating condition of the nuclear powerplant. However the calculation procedure for the fluid-elastic instability was so complicated that the consolidated computer program has not been developed until now. In this study, the numerical calculation procedure and the computer program to obtain the stability ratio were developed. The thermal-hydraulic data in the region of secondary side of steam generator was obtained from executing the ATHOS3 code. The distribution of the fluid density can be calculated by using the void fraction, enthalpy, and operating pressure. The effective mass distribution along the U-tube was required to calculate natural frequency and dynamic mode shape using the ANSYS ver. 5.6 code. Finally, stability ratios for selected tubes of the CE type steam generator were computed. We considered the YGN 3.4 nuclear powerplant as the model plant, and stability ratios were investigated at the flow exit region of the U-tube. From our results, stability ratios at the central and the outside region of the tube bundle are much higher than those of other region.

• Nuclear Engineering and Technology
• /
• v.51 no.4
• /
• pp.1008-1016
• /
• 2019

In the SG (steam generator) of PWR (pressurized water reactor) for a nuclear plant, hundreds of U-shaped tubes are used for the heat exchanger system. They interact with primary pressurized cooling water flow, generating flow-induced vibration in the secondary flow region. A simplified U-tube model is proposed in this study to apply for experiment and its counterpart computation. Using the commercial code, ANSYS-CFX, we first verified the Moody chart, comparing the straight pipe theory with the results derived from CFD (computational fluid dynamics) analysis. Considering the virtual mass of fluid, we computed the major modes with the low natural frequencies through the comparison with impact hammer test, and then investigated the effect of pump flow in the frequency domain using FFT (fast Fourier transform) analysis of the experimental data. Using two-way fluid-structure interaction module in the CFD code, we studied the influence on mean flow rate to generate the displacement data. A feasible CFD method has been setup in this research that could be applied potentially in the field of nuclear thermal-hydraulics.