scholarly journals Technical assessment of two-phase flow aspects of nuclear reactor safety. Progress report, July 1, 1974--March 27, 1975

1975 ◽  
Author(s):  
G.B. Wallis
Keyword(s):  
Nuclear Reactor ◽  
Two Phase Flow ◽  
Progress Report ◽  
Phase Flow ◽  
Reactor Safety ◽  
Two Phase ◽  
Nuclear Reactor Safety ◽  
Technical Assessment

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.

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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