• Nuclear Engineering and Technology
• /
• v.23 no.3
• /
• pp.328-339
• /
• 1991

A multi-purpose research reactor called KMRR has been developed by Korea Atomic Energy Research Institute(KAERI) to generate a maximum thermal output of 30 MW. As a part of thermal hydraulics study, pressure drop characteristics of the longitudinally finned fuel rod bundles were experimentally investigated in a recirculating water test loop. The present study is focused on the investigation of fin effects on pressure drop and the development of pressure drop correlation for the finned rod bundles in a wide range of flow conditions. Friction factor correlations for each design of the finned rod bundles are developed. The value of friction factor for the finned rod bundles was higher than the analytical solution (64/Re) of laminar circular channel new but became lower than the Blasius equation as Reynolds number was increased.

• Proceedings of the SAREK Conference
• /
• 2005.11a
• /
• pp.300-305
• /
• 2005

The present experimental study investigates single-phase heat transfer coefficients downstream of support grid in $6{\times}6$ rod bundles. Support grid with split mixing vanes enhance heat transfer in rod bundles by generating it make turbulence. But this turbulence is confined to short distance. Support grid with LSVF mixing vanes enhanced heat transfer to longer distance. The corresponding Reynolds number investigated in the present study is Re=30,000. The heat transfer coefficients are measured using heated and unheated copper sensor.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
• /
• v.2 no.4
• /
• pp.295-302
• /
• 1990

The purpose of this study is augmentation of heat transfer without additional power in two-dimensional impinging air jet. The technique of heat transfer augmentation used in this experiment is to place rod bundles in front of the flat heated surface. The effects of rod diameter, nozzle-to-target plate distance and the nozzle exit velocity on heat transfer have been investigated. The main conclusions obtained from this experiment are as follows. High heat transfer augmentation is achieved by means of flow acceleration and thinning of boundary layer by placing rod bundles in front of the flat plate. Average heat transfer coefficient becomes maximum in the case of H/B=10,D=4mm. For H/B=2,D=4mm, maximum heat transfer augmentation has been determined to be about 1.5 times larger than that of the flat plate. Heat transfer augmentation by placing the rod bundles at 12m/s is to be about 2 times more than increasing nozzle exit velocity from 12m/s to 18m/s.

• Nuclear Engineering and Technology
• /
• v.56 no.2
• /
• pp.410-418
• /
• 2024

Study of flow behavior within rod bundles has been an active topic. Surface modification technologies are important parts of the design of the fourth generation reactor, which can increase the strength of the secondary flow within the rod bundle lattices. Quasi-periodic large-scale vortex structure (QLVS) is introduced by arranging micro ribs on the surface of rod bundles, which enhanced the scale of the secondary flow between the rod bundle lattices. Using computational fluid dynamics (CFD) and water experiments, the flow field distribution and drag coefficient of the rod-bundle lattices are studied. The secondary flow between the micro-ribbed rod-bundle lattice is significantly enhanced compared to the standard rod-bundle lattice. The numerical simulation results agree well with the experimental results.

• Proceedings of the SAREK Conference
• /
• 2006.06a
• /
• pp.188-193
• /
• 2006

In present study, the thermal-hydraulic characteristics of the subchannels are investigated as measuring single-phase heat transfer coefficients and the cross sectional velocity field using LDV in the downstream of support grid in $6{\times}6$ rod bundles. Support grid with mixing vanes make enhancing heat transfer in rod bundles by generating turbulent flow. But this turbulent flow only is reserved in a short distance. Support grid with LSVF mixing vanes keep the turbulent flow a long distance. The experiments are performed at the nominal Reynolds number 30,000 and 50,000. The heat transfer coefficients are measured using heated and unheated copper sensor. In this study, the comparison of local heat transfer coefficients for LSVF mixing vane and split mixing vane is represented.

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

• Nuclear Engineering and Technology
• /
• v.52 no.7
• /
• pp.1386-1395
• /
• 2020

High fidelity nuclear reactor fuel assembly simulation using CFD method is an effective way for the structure design and optimization. The validated models and user practice guidelines play critical roles in achieving reliable results in CFD simulations. In this paper, the international benchmark MATiS-H is studied carefully and the best user practice guideline is achieved for the rod bundles simulation. Then a 5 × 5 rod bundles model in the advanced pressurized water reactor (PWR) is established and the detailed three-dimensional thermal-hydraulic characteristics are investigated. The influence of spacer grids and mixing vanes on the flow and hear transfer in rod bundles is revealed. As the coolant flows through the spacer grids and mixing vanes in the rod bundles, the drastic lateral flow would be induced and the pressure drop increases significantly. In addition, the heat transfer is enhanced remarkably due to the strong mixing effects. The calculation results could provide meaningful guidelines for the design of advanced PWR fuel assembly.

• Nuclear Engineering and Technology
• /
• v.23 no.3
• /
• pp.275-284
• /
• 1991

It is important to predict the main feature of fully developed turbulent secondary flow through infinite triangular arrays of parallel rod bundles. One-equation turbulence model which include anisotropic eddy viscosity model was applied to predict the exact velocity field. For a constant properties, Reynolds equations were solved by the finite element method. Mean axial velocity near the wall is simulated by the law of the wall. The numerical results showed good agreement with avaiable experimental data. The strength of the secondary flow increased with Reynolds number but decreased with rod spacing, P/D (pitch-to-diameter). The secondary flow affects remarkably the distribution of the axial velocity, wall shear stress and turbulent kinetic energy in the closely packed rod array bundles.

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

• Nuclear Engineering and Technology
• /
• v.24 no.3
• /
• pp.263-273
• /
• 1992

The study on the velocity distribution and the pressure drop characteristic of the nuclear fuel assembly is of importance for the thermal hydraulic design and safety analysis. The purpose of this experimental study is to investigate the hydraulic mixing behind the different kinds of spacer grids in the now or rod bundles. In this study, the detailed hydraulic characteristics in subchannels of 5$\times$5 PWR(Pressurized Water Reactor) rod bundles were measured using one-component He-Ne LDV(Laser Doppler Velocimeter). Measurements of the axial velocity, turbulent intensities and pressure drops were peformed Lateral velocity, turbulent intensities and Reynolds shear stress were also measured by adjust-ing LDV alignment. Friction factors in rod bundles and loss coefficients for spacer grids were evaluated from the measured pressure drops. Hydraulic mixing performance for different kinds of spacer grids could be investigated by estimating the turbulent cross-flow mixing rates between neighboring subchannels.