Two-dimensional analysis of gas-particle two phase flow in pulsed MHD channel

Author(s):  
Hiroyuki Sugita ◽  
Tetsuji Matsuo ◽  
Yoshitaka Inui ◽  
Motoo Ishikawa
2000 ◽  
Vol 120 (3) ◽  
pp. 441-448
Author(s):  
Hiroyuki Sugita ◽  
Tetsuji Matsuo ◽  
Yoshitaka Inui ◽  
Motoo Ishikawa

2001 ◽  
Vol 123 (4) ◽  
pp. 811-818 ◽  
Author(s):  
Jun Ishimoto ◽  
Mamoru Oike ◽  
Kenjiro Kamijo

The two-dimensional characteristics of the vapor-liquid two-phase flow of liquid helium in a pipe are numerically investigated to realize the further development and high performance of new cryogenic engineering applications. First, the governing equations of the two-phase flow of liquid helium based on the unsteady thermal nonequilibrium multi-fluid model are presented and several flow characteristics are numerically calculated, taking into account the effect of superfluidity. Based on the numerical results, the two-dimensional structure of the two-phase flow of liquid helium is shown in detail, and it is also found that the phase transition of the normal fluid to the superfluid and the generation of superfluid counterflow against normal fluid flow are conspicuous in the large gas phase volume fraction region where the liquid to gas phase change actively occurs. Furthermore, it is clarified that the mechanism of the He I to He II phase transition caused by the temperature decrease is due to the deprivation of latent heat for vaporization from the liquid phase. According to these theoretical results, the fundamental characteristics of the cryogenic two-phase flow are predicted. The numerical results obtained should contribute to the realization of advanced cryogenic industrial applications.


Author(s):  
Yumi Yamada ◽  
Toyou Akashi ◽  
Minoru Takahashi

In a lead-bismuth alloy (45%Pb-55%Bi) cooled direct contact boiling water fast reactor (PBWFR), steam can be produced by direct contact of feed water with primary Pb-Bi coolant in the upper core plenum, and Pb-Bi coolant can be circulated by buoyancy forces of steam bubbles. As a basic study to investigate the two-phase flow characteristics in the chimneys of PBWFR, a two-dimensional two-phase flow was simulated by injecting argon gas into Pb-Bi pool in a rectangular vessel (400mm in length, 1500mm in height, 50mm in width), and bubble behaviors were investigated experimentally. Bubble sizes, bubble rising velocities and void fractions were measured using void probes. Argon gas was injected through five nozzles of 4mm in diameter into Pb-Bi at two locations. The experimental conditions are the pressure of atmospheric pressure, Pb-Bi temperatures of 443K, and the flow rate of injection Ar gas is 10, 20, and 30 NL/min. The measured bubble rising velocities were distributed in the range from 1 to 3 m/s. The average velocity was about 0.6 m/s. The measured bubble chord lengths were distributed from 1mm up to 30mm. The average chord length was about 7mm. An analysis was performed by two-dimensional and two-fluid model. The experimental results were compared with the analytical results to evaluate the validity of the analytical model. Although large diameter bubbles were observed in the experiment, the drag force model for spherical bubbles performed better for simulation of the experimental result because of high surface tension force of Pb-Bi.


2013 ◽  
Vol 54 (7) ◽  
Author(s):  
Simo A. Mäkiharju ◽  
Celine Gabillet ◽  
Bu-Geun Paik ◽  
Natasha A. Chang ◽  
Marc Perlin ◽  
...  

2019 ◽  
Vol 52 (1) ◽  
pp. 015501
Author(s):  
Shirin Najafizadeh ◽  
Afshin Ahmadi Nadooshan ◽  
Morteza Bayareh

Author(s):  
Yota Suzuki ◽  
Yusei Tanaka ◽  
Taku Sakka ◽  
Akinori Sato ◽  
Kazuyuki Takase ◽  
...  

Clarifying thermal-hydraulic characteristics in a nuclear reactor core is important in particular to enhance the thermo-fluid safety of nuclear reactors. Spacers installed in subchannels of fuel assemblies have the role of keeping the interval between adjacent fuel rods constantly. Similarly, in case of PWR the spacer has also the role as the turbulence promoter. When the transient event occurs, two-phase flow is generated by boiling of water due to heating of fuel rods. Therefore, it is important to confirm the two-phase flow behavior around the spacer. So, the effect of the spacer affecting the two-phase flow was investigated experimentally at forced convective flow condition. Furthermore, in order to improve the thermal safety of current light water reactors, it is necessary to clarify the two-phase flow behavior in the subchannels at the stagnant flow condition. So, the bubbly flow data around a simulated fuel rod were obtained experimentally at the stagnant flow condition. A wire-mesh sensor was used to obtain a detailed two-dimensional void fraction distribution around the simulated spacer and fuel rod. As a result of this research, the bubbly behavior around the simulated spacer and fuel rod was qualitatively revealed and also bubble dynamics in the sub-channels at the conditions of forced convective and stagnant flows were evaluated. The present experimental data are very useful for verifying the detailed three-dimensional two-phase flow analysis codes.


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