Two-Phase Cross Flow Between Subchannels in a Tight-Lattice Rod Bundle

Author(s):  
Weizhong Zhang ◽  
Hiroyuki Yoshida ◽  
Kazuyuki Takase

In relation to the thermal-hydraulic design of an innovative Flexible-fuel-cycle Water Reactor (FLWR), this study investigates inter-subchannel cross flow phenomena in a tight-lattice rod bundle. Numerical simulations of cross flow using advance interface tracking method were performed and the results were analyzed by a statistical method to clarify the characteristics of inter-subchannel two phase cross flow in the FLWR reactor core. It was revealed that strong correlation exists between differential pressure and gas/liquid mixing coefficients, and cross flow results mechanistically from differential pressure between subchannels. An approximate model is presented which permits the prediction in detail of the components of the inter-subchannel fluctuation differential pressure. The instantaneous fluctuation of differential pressure between two subchannels in gas-liquid slug flow regime is deemed as a result of the intermittent nature of slug flow in each subchannel. The model is based on the detailed numerical simulation results that pressure drop occurs mainly in liquid slug region and in the bubble region it is negligibly small. The instantaneous fluctuation of differential pressure is associated with pressure gradient in the liquid slug for each channel. In addition to a hydrostatic gradient, acceleration and frictional gradients are taken into account to predict pressure gradient in the liquid slug. This model used in conjunction with the numerical simulation code works satisfactorily to reproduce numerical simulation results for instantaneous fluctuation of differential pressure between two modeled subchannels.

2008 ◽  
Vol 131 (2) ◽  
Author(s):  
Weizhong Zhang ◽  
Hiroyuki Yoshida ◽  
Kazuyuki Takase

To explore the mechanism of differential pressure fluctuation inducing cross flow between subchannels in the tight-lattice rod bundle, an evaluation method is presented, which permits the prediction in detail of the unsteady differential pressure fluctuation behavior between subchannels. The instantaneous fluctuation of differential pressure between two subchannels in gas-liquid slug flow regime is deemed as a result of the intermittent nature of slug flow in each subchannel. The method is based on the detailed numerical simulation result of two-phase flow that pressure drop occurs mainly in the liquid slug region and it is, however, negligibly small in the bubble region. The instantaneous fluctuation of differential pressure between two subchannels is associated with pressure gradient in the liquid slug for each channel. In addition to a hydrostatic gradient, acceleration and frictional gradients are taken into account to predict pressure gradient in the liquid slug. This method used in conjunction with the numerical simulation code works satisfactorily to reproduce numerical simulation results for instantaneous fluctuation of differential pressure between two modeled subchannels. It is shown that the static head, acceleration, and frictional pressure drops in the liquid slug are main contributions to the fluctuation of differential pressure between subchannels.


Author(s):  
Weizhong Zhang ◽  
Hiroyuki Yoshida ◽  
Kazuyuki Takase

An approximate model is presented which permits the prediction in detail of the unsteady differential pressure fluctuation behavior between subchannels in the nuclear reactor core. The instantaneous fluctuation of differential pressure between two subchannels in gas-liquid slug flow regime is deemed as a result of the intermittent nature slug flow in each subchannel. The model is based on the detailed numerical simulation result of two-phase flow that pressure drop occurs mainly in liquid slug region and in the bubble region it is negligibly small. The instantaneous fluctuation of differential pressure between the two subchannels is associated with pressure gradient in the liquid slug for each channel. In addition to a hydrostatic gradient, acceleration and frictional gradients are taken into account to predict pressure gradient in the liquid slug. This model temporarily used in conjunction with the numerical simulation code works satisfactorily to reproduce numerical simulation results for instantaneous fluctuation of differential pressure between two modeled subchannels.


2007 ◽  
Vol 2007.60 (0) ◽  
pp. 219-220
Author(s):  
Tatsuya HIGUCHI ◽  
Shinji MATSUNAGA ◽  
Michio SADATOMI ◽  
Akimaro KAWAHARA

Author(s):  
Hidesada Tamai ◽  
Akira Ohnuki ◽  
Hajime Akimoto

Evaluation of a critical heat flux is one of the most important issues for design of an advanced water-cooled reactor core. Since it becomes difficult to perform full-scale experiments due to a larger scale of the advanced reactor cores, an analytical approach has been widely noticed in the core design. To predict the critical heat flux in high accuracy, it is required to correctly understand a horizontal distribution of a two-phase flow in the rod bundles. In this study, the two-phase flow characteristics through narrow gaps in the tight-lattice 37-rod bundle experiment at JAERI were investigated using the subchannel analysis code, NASCA. At the center of the bundle, liquid flowed toward the periphery due to the diversion cross-flow at the elevation where boiling started and the turbulent mixing and the void drift were not influential as they can be neglected. On the periphery of the bundle, the flow mixings due to the diversion cross flow, turbulent mixing and void drift were almost the same order. Gas flowed in the same way with the liquid phase due to the diversion cross-flow, and the turbulent mixing and the void drift moved the gas in the opposite way of the liquid phase migration. An amount of the diversion cross-flow for the liquid phase increased in proportion to the square of the mass velocity. The characteristics of cross flow were almost the same in the different local power peaking and in the different gap widths in the present model.


Author(s):  
Tatsuya Higuchi ◽  
Akimaro Kawahara ◽  
Michio Sadatomi ◽  
Hiroyuki Kudo

Single- and two-phase diversion cross-flows arising from the pressure difference between tight lattice subchannels are our concern in this study. In order to obtain a correlation of the diversion cross-flow, we conducted adiabatic experiments using a vertical multiple-channel with two subchannels simplifying the triangle tight lattice rod bundle for air-water flows at room temperature and atmospheric pressure. In the experiments, data were obtained on the axial variations in the pressure difference between the subchannels, the ratio of flow rate in one subchannel to the whole channel, the void fraction in each subchannel for slug-churn and annular flows in two-phase flow case. These data were analyzed by use of a lateral momentum equation based on a two-fluid model to determine both the cross-flow resistance coefficient between liquid phase and channel wall and the gas-liquid interfacial friction coefficient. The resulting coefficients have been correlated in a way similar to that developed for square lattice subchannel case by Kano et al. (2002); the cross-flow resistance coefficient data can be well correlated with a ratio of the lateral velocity due to the cross-flow to the axial one irrespective of single- and two-phase flows; the interfacial friction coefficient data were well correlated with a Reynolds number, which is based on the relative velocity between gas and liquid cross-flows as the characteristic velocity.


2006 ◽  
Vol 2006.3 (0) ◽  
pp. 3-4
Author(s):  
Hiroyuki KUDO ◽  
Shinji MATSUNAGA ◽  
Tatsuya HIGUCHI ◽  
Akimaro KAWAHARA ◽  
Michio SADATOMI

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