• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2005.11a
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• pp.503-506
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• 2005

The PWR Nuclear Fuel assembly consists of more than 250 fuel rods that are supported by leaf springs in the cells of more than 10 Spacer Grids (SG) along the rod length. Since it is not easy to conduct mechanical tests on a full-scale model basis, the small-scaled rod bundle (5$\times$5) is generally used for various performance tests during the development stage. As one of the small-scaled tests, a flow test should be carried out in order to verify the performance of the spacer grid like the coolant mixing performance and to obtain the Flow-Induced Vibration (FIV) characteristics of the rod bundle over the specified flow range. A vibration test should be also performed to obtain the modal parameters of the bundle prior to the flow test. In this study, we want to develop the estimation procedure of the damping ratio for the small scaled test bundle. For the damping factor of the rod bundle and the grid case at the first vibration mode, as one of the vibration tests, a so-called pluck testing has been performed in air as a preliminary test prior to in-flow damping measurement test. Logarithmic decrement method is used for calculation of the damping ratio. Estimated damping ratio of the rod bundle is about 0.7% with reasonable error of 2% for the previous results. Nonlinear behavior of the rod bundle might be stem mainly Iron the rod-grid support configuration.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2873-2878
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• 2007

The mixing vanes attached to the spacer grid of rod bundles are used to improve the heat transfer in heat exchanger devices by controlling the characteristics of the flow structures and turbulence. In this study, velocity patterns induced by two types of mixing vane(split and swirl vane) are measured by the PIV technique to better understand how to effect on the cross and secondary vortex flow patterns in $5{\times}$ rod bundle simulating the fuel assembly of the nuclear reactor. A successful measurement of the lateral velocity patterns was conducted using a specially designed beam sheet generator and experimental loop at KAERI. As the result, we found that for the cross flow between subchannels, the split vane is more effective than the swirl vane, while for the secondary vortex flow in each subchannel, the swirl vane's one is larger and longer than split vane's one.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.05a
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• pp.137-137
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• 2003

• Transactions of the Korean Society of Mechanical Engineers B
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• v.30 no.1 s.244
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• pp.24-31
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• 2006

Mixing vanes have been installed in the space grid of nuclear fuel rod bundle to improve turbulent heat transfer. Split mixing vanes induce the vortex flow in the cooling water to swirl in sub-channel of fuel assembly. But, The swirling flow decays rapidly so that the heat transfer enhancing effect limited to short length after the mixing vane. In thi present study, the large scale vortex flow(LSVF) is generated by rearranging the mixing vanes to the coordinated directions. This LSVF mixing vanes generate the most strong secondary flow vortices which maintain about 35 $D_H$ after the spacer grid. The streamwise vorticity generated by LSVF sustain two times more than that split mixing vane. Heat transfer in the rod bundle occurs greatly at the same direction to cross flow, and maximum temperature at the surface of bundle drops about 1.5K

• Transactions of the Korean Society of Mechanical Engineers B
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• v.36 no.7
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• pp.689-695
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• 2012

The friction factor in a rod bundle and the loss coefficient at a spacer grid were examined. As a test section, 25 smooth rods, 9.5 mm in diameter and 2000 mm in length, were prepared and installed in a $5{\times}5$ square array in a square channel. In this case, the P/D (Pitch-to-Diameter ratio) was 1.35. In this work, plain (i.e., no mixing vanes), split-vane, and hybrid-vane spacer grids were tested. In a bare rod bundle (i.e., no spacer grid), the measured friction factors were in good agreement with the previous correlations. Among the spacer grids tested, the hybrid-vane spacer grid presented the largest friction factor in the rod bundle and loss coefficient. This may be because of the flow pattern change induced by large relative plugging of the flow cross section and mixing vane geometry. At Re=$5{\times}10^5$, the predicted loss coefficients of plain, splitvane, and hybrid-vane spacer grids were approximately 0.79, 0.80, and 0.88, respectively.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.337-338
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• 2005

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1914-1928
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• 2022

To improve the heat transfer efficiency of the reactor fuel assembly, it is necessary to accurately calculate the two-phase flow boiling characteristics and the critical heat flux (CHF) in the fuel assembly. In this paper, a Eulerian two-fluid model combined with the extended wall boiling model was used to numerically simulate the 5 × 5 fuel rod bundle with spacer grids (four sets of mixing vane grids and four sets of simple support grids without mixing vanes). We calculated and analyzed 11 experimental conditions under different pressure, inlet temperature, and mass flux. After comparing the CHF and the location of departure from the nucleate boiling obtained by the numerical simulation with the experimental results, we confirmed the reliability of computational fluid dynamic analysis for the prediction of the CHF of the rod bundle and the boiling characteristics of the two-phase flow. Subsequently, we analyzed the influence of the spacer grid and mixing vanes on the void fraction, liquid temperature, and secondary flow distribution. The research in this article provides theoretical support for the design of fuel assemblies.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2303-2308
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• 2007

Experiments were performed to determine the thermal (or turbulent) diffusion coefficient (TDC) and to investigate the critical heat flux (CHF) performance in the 5${\times}$5 rod bundle with 5 unheated rods which are supported by Hybrid Mixing Vane. In this study, HFC-134a fluid was used as working fluid and the fluid temperature were measured in the important subchannels. To determine the TDC value, the measured fluid temperatures were compared with the predicted values obtained from the MATRA code. The best optimized value of ${\beta}$ was found to be 0.02 by considering prediction statistics, i.e., average and standard deviations of the differences between the experimental results and code calculations. Using the best optimized value of ${\beta}$ as 0.02, the MATRA code predicts the test results of the fluid temperature within ${\pm}$1.0 % of error. According to the experimental results on CHF of 5 non-heating guide tubes, the case with non-heating guide tube showed a little good performance in terms of CHF.

• Nuclear Engineering and Technology
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• v.37 no.5
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• pp.457-468
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• 2005

An experimental study on post-CHF heat transfer has been performed with a $3\times3$ rod bundle using a vertical steam-water two-phase flow at low flow conditions. The effects of various parameters on the post-CHF heat transfer are investigated and the reasons for the parametric effects are discussed. As the heat transfer regime changes from CHF to post-CHF, the radial wall temperature distribution is changed depending on the pressure and the mass flux conditions. The superheat of the fluid increases considerably with an increase of the wall temperature (or heat flux) and with a decrease of the mass flux. This implies, indirectly, a strong thermal non-equilibrium at high wall temperature and low mass flux conditions. In order to improve the prediction accuracy of the existing post-CHF correlations, it is necessary to perform more experiments, particularly direct measurement of the vapor superheat, and to modify the correlation by considering a strong thermal non-equilibrium at low flow and low pressure conditions.

• Proceedings of the KSME Conference
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• 2005.11a
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• pp.173.1-173.1
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• 2005