scholarly journals A Derivation of the Nonlocal Volume-Averaged Equations for Two-Phase Flow Transport

2012 ◽  
Vol 2012 ◽  
pp. 1-8 ◽  
Author(s):  
Gilberto Espinosa-Paredes
Keyword(s):  
Nuclear Reactor ◽  
Two Phase Flow ◽  
Solid Phase ◽  
Volume Averaging ◽  
Phase Flow ◽  
Two Phase ◽  
Averaged Equations ◽  
Non Local ◽  
General Transport ◽  
Volume Method

In this paper a detailed derivation of the general transport equations for two-phase systems using a method based onnonlocalvolume averaging is presented. Thelocalvolume averaging equations are commonly applied in nuclear reactor system for optimal design and safe operation. Unfortunately, these equations are limited to length-scale restriction and according with the theory of the averaging volume method, these fail in transition of the flow patterns and boundaries between two-phase flow and solid, which produce rapid changes in the physical properties and void fraction. Thenon-localvolume averaging equations derived in this work contain new terms related withnon-localtransport effects due to accumulation, convection diffusion and transport properties for two-phase flow; for instance, they can be applied in the boundary between a two-phase flow and a solid phase, or in the boundary of the transition region of two-phase flows where thelocalvolume averaging equations fail.

The influence of the Stokes and Archimedes forces on the stability of a two-phase flow

2003 ◽  
Vol 3 ◽  
pp. 266-270
Author(s):  
B.H. Khudjuyerov ◽  
I.A. Chuliev
Keyword(s):  
Spectral Method ◽  
Stability Region ◽  
Gas Flow ◽  
Two Phase Flow ◽  
Solid Phase ◽  
Stability Characteristic ◽  
Phase Flow ◽  
Two Phase ◽  
The Stability

The problem of the stability of a two-phase flow is considered. The solution of the stability equations is performed by the spectral method using polynomials of Chebyshev. A decrease in the stability region gas flow with the addition of particles of the solid phase. The analysis influence on the stability characteristic of Stokes and Archimedes forces.

Calculating Method for the Solid Phase Effect Coefficient of the Solid-Fluid Two Phase Flow Centrifugal Pump

2010 ◽  
Vol 29-32 ◽  
pp. 658-663
Author(s):  
Chuan Jun Li ◽  
Gang Yu ◽  
Xin Wang
Keyword(s):  
Differential Equation ◽  
Centrifugal Pump ◽  
Two Phase Flow ◽  
Solid Phase ◽  
High Accuracy ◽  
Phase Flow ◽  
Phase Effect ◽  
Two Phase ◽  
Error Band ◽  
Calculating Method

In order to acquire the solid-fluid two phase flow centrifugal pump’s slurry head according with the fact, the solid phase effect coefficient must be calculated precisely. By analysising the acting forces on the solid particle, its moving differential equation was established. And the calculating formula of the solid phase effect coefficient was deduced based on the equation. For the sake of verifying its validity, a test of contrast and comparison on the calculating slurry heads by some ways was carried out. The results shows that the relative error values of the slurry head calculated are less than 2.00% with a small and stably error band. The method has the advantage of simple calculating process, high accuracy, low randomness and good practicability.

Comparison of Drift-Flux Models for Void Fraction Prediction in Sub-Channel of Vertical Rod Bundles

2018 ◽  
Author(s):  
Quanyao Ren ◽  
Liangming Pan ◽  
Wenxiong Zhou ◽  
Tingpu Ye ◽  
Hang Liu ◽  
...  
Keyword(s):  
Void Fraction ◽  
Drift Velocity ◽  
Nuclear Reactor ◽  
Two Phase Flow ◽  
Rectangular Channel ◽  
Phase Flow ◽  
Rod Bundles ◽  
Two Phase ◽  
Drift Flux Model ◽  
Drift Flux

In order to simulate the transfer of mass, momentum and energy in the gas-liquid two-phase flow system, tremendous work focused on the phenomenon, mechanisms and models for two-phase flow in different channels, such as circular pipe, rectangular channel, rod bundle and annulus. Drift-flux model is one of the widely used models for its simplicity and good accuracy, especially for the reactor safety analysis codes (RELAP5 and TRAC et al.) and sub-channel analysis code (COBRA, SILFEED and NASCA et al.). Most of the adopted drift-flux models in these codes were developed based on the void fraction measured in pipe and annulus, which were different with the actual nuclear reactor. Although some drift-flux models were developed for rod bundles, they were based on the void fraction on the whole cross-section not in subchannel in rod bundles due to the lack of effective measuring methods. A novel sub-channel impedance void meter (SCIVM) has been developed to measure the void fraction in sub-channel of 5 × 5 rod bundles, which is adopted to evaluate these existing drift-flux models for rod bundles. By comparison, the values of drift-flux parameters have large differences among different correlations, which are suggested to be reconsidered. Based on the experimental data and physical laws, Lellouche-Zolotar and Chexal-Lellouche correlations show a better performance for drift velocity. If the predicting error of void fraction is the only concerned parameter, Chen-Liu, Ishizuka-Inoue and Chexal-Lellouche correlations are recommended for averaged relative error less than 30%. More experiments are suggested to focus on the distribution parameter and drift velocity through their definition.

Verification of STACS-VOF Based Two-Phase Flow Model for Interfacial Flows

2012 ◽  
Vol 212-213 ◽  
pp. 1098-1102
Author(s):  
Bin Deng ◽  
Chang Bo Jiang ◽  
Zhi Xin Guan ◽  
Chao Shen
Keyword(s):  
Flow Model ◽  
Two Phase Flow ◽  
Three Dimensional ◽  
Test Cases ◽  
Phase Flow ◽  
Interfacial Flows ◽  
Governing Equations ◽  
Two Phase ◽  
Resolution Scheme ◽  
Volume Method

The numerical calculation and simulation of gas-liquid two-phase flows with interfacial deformations have nowadays become more and more popular issues in various scientific and industrial fields. In this study, a three-dimensional gas-liquid two-phase flow numerical model is presented for investigating interfacial flows. The finite volume method was used to discretize the governing equations. A High-resolution scheme of VOF method (STACS) is applied to capture the free surface. The paper outlines the methodology of STACS and its validation against three typical test cases used to verify its accuracy. The results show the STACS-VOF gives very satisfactory results for three-dimensional two-phase interfacial flows problem, and this scheme performs more accurate and less diffusive preserving interface sharpness and boundedness.

Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration: (13) Rising Bubble Motion Under Horizontal Vibration

2014 ◽  
Author(s):  
Rie Arai ◽  
Akiko Kaneko ◽  
Hideaki Monji ◽  
Yutaka Abe ◽  
Hiroyuki Yoshida ◽  
...  
Keyword(s):  
Void Fraction ◽  
Nuclear Reactor ◽  
Two Phase Flow ◽  
Liquid Velocity ◽  
Nuclear Reactors ◽  
Bubbly Flow ◽  
Structural Safety ◽  
Phase Flow ◽  
Two Phase ◽  
Safety Operation

An earthquake is one of the most serious phenomena for the safety of a nuclear reactor in Japan. Therefore, structural safety of nuclear reactors has been studied and nuclear reactors ware contracted with structural safety for a big earthquake. However, it is not enough for safety operation of nuclear reactors because thermal-fluid safety is not confirmed under the earthquake. For instance, behavior of gas-liquid two-phase flow is unknown under the earthquake conditions. Especially, fluctuation of void fraction is an important factor for the safety operation of the nuclear reactor. In the previous work, fluctuation of void faction in bubbly flow was studied experimentally and theoretically, to investigate the stability of the bubbly flow. In such studies, flow rate or void fraction fluctuations were given to the steady bubbly flow. In the case of the earthquake, the fluctuation is not only the flow rate, but also a body force on the two-phase flow and a shear force through a pipe wall. Interactions of gas and liquid through their interface also act on the behavior of the two-phase flow. The fluctuation of the void fraction is not clear for such complicated situation under the earthquake. Therefore, in this research project, the behavior of gas-liquid two-phase flow is investigated experimentally and numerically in the series of study. In this study, to investigate the effects of vibration on bubbly flow in the components and construct an experimental database for validation, we performed visualization experiments of vertical bubbly flow in a rectangular water tank on which a sine wave vibration was applied. In this paper, results of visualized experiment evaluated by the visualization techniques, including positions of bubbles, shapes of bubbles and liquid velocity distributions around bubbles, were shown. And liquid velocity distribution around bubbles by the PIV measurement was also shown. In the results, bubble behaviors were affected by oscillation. And the cycle of the bubble tilt angle was almost same as the cycle of oscillation table velocity.

scholarly journals RECENT IMPROVEMENTS IN THE CUPID CODE FOR A MULTI-DIMENSIONAL TWO-PHASE FLOW ANALYSIS OF NUCLEAR REACTOR COMPONENTS

2014 ◽  
Vol 46 (5) ◽  
pp. 655-666 ◽  
Author(s):  
HAN YOUNG YOON ◽  
JAE RYONG LEE ◽  
HYUNGRAE KIM ◽  
IK KYU PARK ◽  
CHUL-HWA SONG ◽  
...  
Keyword(s):  
Nuclear Reactor ◽  
Two Phase Flow ◽  
Flow Analysis ◽  
Phase Flow ◽  
Two Phase
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