• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.13 no.1
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• pp.9-17
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• 2001

In this study, flow boiling experiments were performed using R-134a on a plain tube bundle. Tests were conducted for the following range of variables; quality from 0.1 to 0.9, mass flux from $8\;kg/m^2s$ to $26\;kg/m^2s$ and heat flux from $10\;kW/m^2s$ to $40\;kW/m^2s$. The heat transfer coefficients were strongly dependent on the heat flux. However, they were almost independent on the mass flux or quality. The data are compared with the modified Chen model, which satisfactorily () predicted the data. Original Chen model, however, did not adequately predict the effect of quality. The reason may be attributed to the flow pattern of the present test, where the bubbly flow prevailed for the entire test range. The heat transfer coefficients of the tube bundle were 6~40% higher than those of the single tube pool boiling.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.30 no.7 s.250
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• pp.612-621
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• 2006

Drawing is a mechanical operation attenuating material thickness to an appropriate level for the next processing or end usage. When the input material has a form of bundle or bundles made of very thin and long shaped wires or fibers, this attenuation operation is called 'bundle drawing' or 'drafting'. Bundle drawing is being used widely in manufacturing micro sized wires or staple yarns. However, the bundle processed by this operation has more or less defects in the evenness of linear density. Such irregularities cause many problems not only for the product quality but also for the efficiency of the next successive processes. In this research a mathematical model for the dynamic behavior of the bundle fluid is to be set up on the basis of general physical laws containing physical variables, i.e. linear density and velocity as the dynamic state variables of the bundle fluid. The governing equations resulting from the modeling show that they appear in a slightly different form from what they do in a continuum fluid. Then, the governing equations system is simplified in a steady state and the bundle dynamics is simulated, showing that the shape of the velocity profiles depends on two model parameters. Experiments confirm that the model parameters are to be well adjusted to show a coincidence with the theoretical analysis. The higher the drawing ratio and drawing speed we, the more sensitive becomes the bundle flow to exogenous disturbances.

• Proceedings of the Korean Society of Precision Engineering Conference
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• 2005.06a
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• pp.965-968
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• 2005

Roll drafting operation causes variations in the linear density of bundles because the bundle flow cannot be controlled completely by roll pairs. Defects occurring in this operation bring about many problems successively in the next processes. In this paper, we attempt to analyze the draft dynamics and the linear density irregularity based on the governing equation of a bundle motion that has been suggested in our previous studies. For analyzing the dynamic characteristics of the roll drafting operation, it is indispensable to investigate a transient state in time domain before the bundle flux reaches a steady state. However, since governing equations of bundle flow consisting of continuity and motion equations turn out to be nonlinear, and coupled between variables, the solutions for a transient state cannot be obtained by an analytical method. Therefore, we use the Finite Difference Method(FDM), particularly, the FTBS(Forward-Time Backward-Space) difference method. Then, the total equations system yields to an algebraic equations system and is solved under given initial and boundary conditions in an iterative fashion. From the simulation results, we confirm that state variables show different behavior in the transient state; e.g., the velocity distribution in the flow field changes more quickly the linear density distribution. During a transient flow in a drafting zone, the output irregularity is influenced differently by the disturbances, e.g., the variation in input bundle thickness, the drafting speed, and the draft ratio.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.6 no.1
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• pp.65-71
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• 2010

The plugging rate of reheater tubes of Wolsung unit 1 nuclear power plant has been increased by corrosion and erosion since 1990. As the dimensions of the new first stage reheater bundle tubes which were supplied by Hanjung company to replace were different from old one, numerical calculations are carried out for flow and heat transfer in the reheater bundle tubes of the N.P.P. Numerical calculations consists of thermal performance, drain line pressure drop, flow change by pressure drop of line, stress analysis of finned tubes and analysis of flow induced vibration. Computational analysis using heat transfer research institute program is adopted to verify the results of the numerical calculations. It contains the evalution of performance in the system with view to location of the new reheater bundle and it shows the differences between the numerical calculation results and heat transfer research institute program output.

• 한국전산유체공학회:학술대회논문집
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• 2003.10a
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• pp.71-73
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• 2003

A CFD analysis has been made for fully developed turbulent flows in a triangular bare rod bundle with pitch to diameter ratio (P/D) of 1.123. The nonlinear turbulence models predicted the turbulence­driven secondary flow in the triangular subchannel. The nonlinear quadratic $\kappa-\omega$ models by Speziale and Myong-Kasagi predicted turbulence structure in the rod bundle fairly well. The nonlinear quadratic and cubic $\kappa-\omega$ models by Shih et al. and Craft et al. showed somewhat weaker anisotropic turbulence. The differential Reynolds stress model appeared to overpredict the turbulence anisotropy in the rod bundle.

• Nuclear Engineering and Technology
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• v.37 no.6
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• pp.565-574
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• 2005

The CANFLEX Mk-V bundle is designed to improve upon the critical heat flux (CHF) characteristics of the CANFLEX Mk-IV bundle. The main difference between these two bundles is an increase in bearing pad height of about 0.3 mm in the CANFLEX Mk-IV bundle. This change in bearing pad height leads to an increase in gap flow at the bottom of the bundle, primarily eliminating the localized narrow-gap effect that limits the CHF of the CANFLEX Mk-IV bundle. The objective of this paper is to examine the effects of bearing pad height and pressure tube creep on the sheath-temperature distribution, dryout power, and dryout location, as observed ken full-scale bundle tests, between CANFLEX Mk-IV and Mk-V bundles In uncrept and crept channels. A comparison of surface-temperature differences between the top and bottom elements of the bundles showed that increasing the bearing pad height has led to a more homogeneous enthalpy distribution in subchannels of the bundle. Initial dryout locations of the CANFLEX Mk-V bundle were mainly observed at the mid-spacer plane of either the $10^{th}$ (about $80\%$) or $11^{th}$ ($20\%$) bundle in the 12-bundle string, as compared to the mid-spacer and downstream-button planes for the CANFLEX Mk-IV bundle. Dryout power and boiling-length-average (BLA) CHF values exhibit consistent trends and little scatter with varying flow conditions for both types of CANFLEX bundles in uncrept and crept channels. An increase in pressure tube creep has led to a reduction in dryout power (about $20\%$ far the $3.3\%$ crept channel and $27\%$ for the $5.1\%$ crept channel as compared to dryout powers for the uncrept channel). Increasing the bearing pad height of the CANFLEX bundle has led to an increase in the dryout power. Overall, the dryout power of the CANFLEX Mk-V bundle is 7 to $10\%$ higher than that of the CANFLEX Mk-IV bundle at the inlet temperature range of interest (i.e., between 243 and $290^{\circ}C$).

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3625-3634
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• 2021

The crossflow is the key phenomenon in turbulent flow inside rod bundles. In order to establish confidence on application of computational fluid dynamics (CFD) to simulate the crossflow in rod bundles, three Reynolds-Averaged Navier Stokes (RANS) models i.e. the realizable k-ε model, the k-ω SST model and the Reynolds stress model (RSM), and the Large Eddy simulations (LES) with the Wall-Adapting Local Eddy-viscosity (WALE) model are validated based on the Particle Image Velocimetry (PIV) flow measurement experiment in a 5 × 5 rod bundle. In order to investigate effects of periodic boundary condition in the gap, the numerical results obtained with four inner subchannels are compared with that obtained with the whole 5 × 5 rod bundle. The results show that periodic boundaries in the gaps produce strong errors far downstream of the spacer grid, and therefore the full 5 × 5 rod bundle should be simulated. Furthermore, it can be concluded, that the realizable k-ε model can only provide reasonable results very close to the spacer grid, while the other investigated models are in good agreement with the experimental data in the whole downstream flow in the rod bundle. The LES approach shows superiority to the RANS models.

• Transactions of the Korean Society of Mechanical Engineers
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• v.16 no.11
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• pp.2111-2118
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• 1992

In this study, an analytic model is developed to predict Nusselt numbers for turbulent flow in a rod bundle. Flow channel area is divided into several element channels, and simple algebraic equations of universal velocity and temperature profiles are integrated over each element channel. The integral equations are then added to yield an analytic expression for the nusselt number of a rod bundle. The analytic model reasonably predicts the available heat transfer data.

• Nuclear Engineering and Technology
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• v.30 no.4
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• pp.328-341
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• 1998

This paper presents in detail the hydraulic characteristic measurements using LDV(Laser Doppler Velocimetry) in subchannels of a HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids and has a cylindrical configuration. Axial velocity and turbulent intensity were measured. The effects of the spacer grids on the turbulent flow were investigated using the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of the fuel bundle and the loss coefficients for the spacer grids were estimated from the measured pressure drops. The turbulent thermal mixing phenomena were discussed.

• Journal of computational fluids engineering
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• v.21 no.4
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• pp.71-77
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• 2016

The CUPID code is a transient, three-dimensional, two-fluid, thermal-hydraulic code designed for a component-scale analysis of nuclear reactor components. The primary objective of this study is to assess the applicability of CUPID to single-phase turbulent flow analyses of $2{\times}2$ rod bundle subchannel. The bulk velocity at the inlet varies from 1.0 m/s up to 2.0 m/s which is equivalent to the fully turbulent flow with the range of Re=12,500 to 25,000. Adiabatic single-phase flow is assumed. The velocity profile at the exit region is quantitatively compared with both experimental measurement and commercial CFD tool. Three different boundary conditions are simulated and quantitatively compared each other. The calculation results of CUPID code shows a good agreement with the experimental data. It is concluded that the CUPID code has capability to reproduce the turbulent flow behavior for the $2{\times}2$ rod bundle geometry.