Numerical Studies on Dynamic Behavior of Air-Water Cross Flow Between Two Circular Interconnected Channels
スポンサーリンク
概要
- 論文の詳細を見る
The objective of this paper is to investigate the dynamic behavior of air-water fluid transfer across gap regions between fuel rods in LWR bundles, referred to as cross flow. A computational multifluid dynamic simulation of the cross flow is conducted by using an analysis code based on a three-dimensional two-fluid formulation plus an interface-tracking method. This code is applied to two-phase flows inside a single vertical cylindrical tube in order to perform wall friction sensitivity studies with a model that accounts for velocity gradients across boundary layers flowing on walls. Numerical analyses of two-phase cross flow across the gap region of two circular interconnected channels showed that the calculated turbulent mixing rates are in relatively good agreement with the results of experimental tests in the case of increased wall friction at low void fractions. On the other hand, at high void fraction, the agreement is not as good. Spectrum analyses of the calculated mixing velocities of both air and water across the gap region and the calculated differential pressure between channels showed a strong correlation between them. The present numerical study could demonstrate the following observations shown in previous experimental studies. (1) The fluctuation in differential pressure between channels is attributed to the passing of large bubbles through a channel. (2) The fluctuation in differential pressure primarily induces turbulent mixing phenomena.
- 2009-09-01
著者
-
NINOKATA Hisashi
Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology
-
KANEKO Junichi
Nuclear Eng. Dept., Tokyo Institute of Technology
-
MINATO Akihiko
Advancesoft Corporation
-
Ninokata Hisashi
Research Lab. For Nuclear Reactors Tokyo Institute Of Technology
-
Kaneko Junichi
Nuclear Eng. Dept. Tokyo Institute Of Technology
関連論文
- Study of the Self-Controllability for the Fast Reactor Core with High-Thermal-Conductivity Fuel
- Numerical Studies on Dynamic Behavior of Air-Water Cross Flow Between Two Circular Interconnected Channels
- Study of the Self-Controllability for the Fast Reactor Core with High-Thermal-Conductivity Fuel
- Computational Fluid Dynamics Study of Liquid Droplet Impingement Erosion in the Inner Wall of a Bent Pipe
- PRA-Based SMA : the First Tool toward a Risk-Informed Approach to the Seismic Design of the IRIS
- Prediction of Release Rate of Burnt Sodium as Aerosol
- Analysis of an Out-of-pile Experiment for Materials Redistribution under Core Disruptive Accident Condition of Fast Breeder Reactors
- Prediction of the Equilibrium Two-Phase Flow Distributions in Inter-Connected Subchannel Systems
- Concept of Erbium Doped Uranium Oxide Fuel Cycle in Light Water Reactors
- Second International Symposium on Global Environment and Nuclear Energy Systems
- Numerical Method for Simulation of Fluid Flow and Heat Transfer in Geometrically Disturbed Rod Bundles
- Numerical Study on Observed Decay Ratio of Coupled Neutronic-Thermal Hydraulic Instability in Ringhals Unit 1 under Random Noise Excitation
- An Algorithm for Attenuation of Turbulence in Particulate Flow Linked to the Fluid-dynamic Code COMMIX-M
- Development of Parallel Coupling System between Three-Dimensional Nodal Kinetic Code ENTREE and Two-Fluid Plant Simulator TRAC/BF1
- The Multi-Fluid Multi-Phase Subchannel Analysis Code KAMUI for Subassembly Accident Analysis of an LMFR
- Nonlinear Iterative Nodal Method Applied to Neutron Flux Modal Analysis