• Proceedings of the KSME Conference
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• 2004.11a
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• pp.1592-1597
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• 2004

Experimental studies were carried out to confirm the turbulent enhancement of the cooling system of nuclear reactor by large scale vortex generation in nuclear fuel rod bundle. The large scale vortex motions were generated by rearranging the inclination angles of mixing vanes to the coordinate directions. Experimental studies were carried out at Reynolds Number 60,000 with hydraulic condition. Normal variations of mean velocity and turbulent intensity in the rod bundle subchannel were measured by the 2-color LDV measurement system. The turbulence generated by split mixing vanes has small length scales so that they maintain only about 10DH after the spacer grid. On the other hand, the turbulences generated by the large scale vortex continue more and remain up $25D_{H}$ after the spacer grid.

• Transactions of the Korean Society of Mechanical Engineers
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• v.10 no.1
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• pp.7-14
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• 1986

To predict the fuel clad temperature during the reflooding phase of a LOCA, one may need a knowledge of reflood heat tranfer mechanism in a rod bundle. For this purpose reflooding experiments have been carried out with an electrically heated 3*3 rod bundle. Using the method for the determination of local heat transfer coefficient from the measured wall temperature the parametric effects of coolant flow rate, initial wall temperature, coolant subcooling and heat generation rate on the propagation of rewetting front were investigated. Prediction of the wall temperature histories for these experiments was discussed using REFLUX code with modification of the rewetting temperature correlation. Through this modification, better agreement between experiment and prediction was obtained.

• 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.52 no.11
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• pp.2479-2490
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• 2020

We performed experiments to measure a single-phase upward flow in a 5 × 5 rod bundle with spacer grids using a particle image velocimetry, focusing on the crossflow. The Reynolds number based on the hydraulic diameter and the bulk velocity is 10,000. The ratio of pitch between rods and rod diameter is 1.4 and spacer grid is installed periodically. The turbulence in the rod bundle results from the combination of a forced mixing and natural mixing. The forced mixing by the spacer grid persists up to 10D_h from the spacer grid, while the natural mixing is attributed to the crossflow between adjacent subchannels. The combined effects contribute to a sinusoidal distribution of the time-averaged stream-wise velocity along the lateral direction, which is relatively weak right behind the spacer grid as well as in the gap. The streamwise and lateral turbulence intensities are stronger right behind the spacer grid and in the gap. Based on these findings, we newly defined a modified mixing coefficient as the ratio of the lateral turbulence intensity to the time-averaged streamwise velocity, which shows a spatial variation. Finally, we compared the developed model with the measured data, which shows a good agreement with each other.

• Proceedings of the KSME Conference
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• 2005.11a
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• pp.97.1-97.1
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• 2005

• 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.

• Proceedings of the SAREK Conference
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• 2007.11a
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• pp.191-196
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• 2007

This paper presents the measuring technique for local heat transfer coefficients using a copper sensor in a rod bundle with mixing vanes. A copper sensor consists of a cartridge heater and four pieces of thermocouple. The Heater is located at the center of the copper sensor and thermocouples measure the surface temperature of the copper sensor. Unheated copper sensor and heated copper sensor are able to measure the local heat transfer coefficient at the position where the heated copper sensor is installed. However the entire region of a rod bundle is actually not heated, the decay of local heat transfer coefficients measured represents overestimated value rather than an actual value. The calibration curve for local heat transfer coefficients is presented using the correction factor calculated by CFD.

• Transactions of the Korean Society of Mechanical Engineers
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• v.19 no.10
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• pp.2630-2636
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• 1995

A theoretically based prediction for the determination of the subchannel friction factor at low pitch to the rod diameter ratio (P/D < 1.2) in the bare rod bundle flow has been developed. The present model assumes the validity of the Law of Wall over the entire flow area. The algebraic form of the Law of the Wall is integrated over the entire flow area and the local friction velocity variation along the rod periphery is considered in this study. The present method is applied to the rod bundles with P/D < 1.2, and the prediction results show good agreement with the available experimental data.

• Transactions of the Korean Society of Mechanical Engineers
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• v.16 no.8
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• pp.1545-1551
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• 1992

This work is concerned with the prediction of turbulent friction factors in rod bundles. In this study, the Law of the Wall is assumed to be valid over the entire flow area. The flow channel is divided into element channels, and the algebraic form of the "Law of the Wall" is integrated over each element channel to yield an analytic expression for the pressure drop in the rod bundle. The method is applied to rod bundles with 1.2

• Nuclear Engineering and Technology
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• v.28 no.2
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• pp.162-174
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• 1996

The local hydraulic characteristics in a single phase flow of a 6$\times$6 rod bundle with neighboring different spacer grids were measured by using a LDV(Laser Doppler Velocimeter) system. 6$\times$6 rod bundle is formed by two 3$\times$6 rod bundles with different spacer grids. The objective of this study in a rod bundle is to investigate the thermal-hydraulic interactions between different spacer grids with different configurations and resistance. By using a LDV system, the velocity and turbulent intensity in axial and horizontal directions ore measured. Pressure drop measurements ore also performed to evaluate the loss coefficient for the spacer grid and the friction factor for rod bundles. Implications concerning thermal mining due to spacer grids were investigated based on the hydraulic test results. Swirl factor, which is assumed as a qualitative criteria for DNB(departure from nucleate boiling), was defined and estimated from the horizontal velocity result.