Development of Analytical Procedures of Two-Phase Flow in Tight-Lattice Fuel Bundles for Innovative Water Reactor for Flexible Fuel Cycle

2008 ◽  
Vol 164 (1) ◽  
pp. 45-54 ◽  
Author(s):  
Hiroyuki Yoshida ◽  
Akira Ohnuki ◽  
Takeharu Misawa ◽  
Kazuyuki Takase ◽  
Hajime Akimoto
Keyword(s):  
Two Phase Flow ◽  
Fuel Cycle ◽  
Phase Flow ◽  
Two Phase ◽  
Water Reactor ◽  
Analytical Procedures ◽  
Tight Lattice

Development of Quantitative Analytical Procedures on Two-Phase Flow in Tight-Lattice Fuel Bundles for Reduced-Moderation Light-Water Reactors

2004 ◽  
Author(s):  
Kazuyuki Takae ◽  
Hiroyuki Yoshida ◽  
Masatoshi Kureta ◽  
Hidesada Tamai ◽  
Akira Ohnuki ◽  
...  
Keyword(s):  
Two Phase Flow ◽  
Conversion Ratio ◽  
Thermal Design ◽  
Phase Flow ◽  
Two Phase ◽  
Light Water ◽  
Water Reactor ◽  
Model Experiments ◽  
Analytical Procedures ◽  
Tight Lattice

R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) has started at Japan Atomic Energy Research Institute (JAERI) in collaboration with power companies, reactor vendors, universities since 2002. The RMWR is a light water reactor which a higher conversion ratio more than one can be expected. In order to attain this higher conversion ratio, triangular tight-lattice fuel bundles which gap spacing between each fuel rod is around 1 mm are required. As for the thermal design of the RMWR core, conventional analytical methods are no good because the conventional composition equations can not predict the RMWR core with high accuracy. Then, development of new quantitative analytical procedures was carried out. Those analytical procedures are constructed by model experiments and advanced two-phase flow analysis codes. This paper describes the results of the model experiments and analytical results with the developed analysis codes.

TWO-PHASE FLOW INSTABILITIES ANALYSES IN BOILING WATER REACTOR

2016 ◽  
Author(s):  
Antonella Lombardi Costa ◽  
WILMER ARUQUIPA COLOMA ◽  
Antonella Lombardi Costa ◽  
Claubia Pereira ◽  
Maria Veloso ◽  
...  
Keyword(s):  
Two Phase Flow ◽  
Boiling Water ◽  
Flow Instabilities ◽  
Boiling Water Reactor ◽  
Phase Flow ◽  
Two Phase ◽  
Water Reactor

Two-phase flow redistribution due to void drift between triangle tight lattice subchannels

2004 ◽  
Vol 2004.57 (0) ◽  
pp. 321-322 ◽  
Author(s):  
Yuichi SASAKI ◽  
Akimaro KAWAHARA ◽  
Michio SADATOMI ◽  
Keiko KANO
Keyword(s):  
Two Phase Flow ◽  
Phase Flow ◽  
Two Phase ◽  
Tight Lattice

A Large-Scale Numerical Simulation on Bubbly and Liquid Film Flows in Narrow Fuel Channels

2005 ◽  
Author(s):  
K. Takase ◽  
H. Yoshida ◽  
Y. Ose ◽  
H. Akimoto
Keyword(s):  
Large Scale ◽  
Two Phase Flow ◽  
Reactor Core ◽  
Light Water Reactor ◽  
Phase Flow ◽  
Analysis Method ◽  
Two Phase ◽  
Light Water ◽  
Water Reactor ◽  
Fuel Bundle

In order to predict the water-vapor two-phase flow structure in a fuel bundle of an advanced light-water reactor, large-scale numerical simulations were carried out using a newly developed two-phase flow analysis method and a highly parallel-vector supercomputer. Conventional analysis methods such as subchannel codes need composition equations based on many experimental data. Therefore, it is difficult to obtain highly prediction accuracy on the thermal design of the advanced light-water reactor core if the experimental data are insufficient. Then, a new analysis method using the large-scale direct numerical simulation of water-vapor two-phase flow was proposed. The coalescence and fragmentation of small bubbles were investigated numerically and the bubbly flow dynamics in narrow fuel channels were clarified. Moreover, the liquid film flow inside a tight-lattice fuel bundle which is used to the advanced light-water reactor core was analyzed and the water and vapor distributions around fuel rods and a spacer were estimated quantitatively.

Numerical Evaluation of Fluid Mixing Phenomena in Boiling Water Reactor Using Advanced Interface-Tracking Method

2007 ◽  
Author(s):  
Hiroyuki Yoshida ◽  
Takuji Nagayoshi ◽  
Kazuyuki Takase ◽  
Hajime Akimoto
Keyword(s):  
Two Phase Flow ◽  
Flow Simulation ◽  
Fluid Mixing ◽  
Interface Tracking ◽  
Phase Flow ◽  
Two Phase ◽  
Tracking Method ◽  
Water Reactor ◽  
Hydraulic Design ◽  
Actual Size

Thermal-hydraulic design of the current boiling water reactor (BWR) is performed by correlations with empirical results of actual-size tests. Then, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test that simulates its design is required to confirm or modify the correlations. Development of a method that enables the thermal-hydraulic design of nuclear rectors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology. In this study, detailed two-phase flow simulation code using advanced interface tracking method: TPFIT is developed to get the detailed information of the two-phase flow. In this paper, firstly, we tried to verify the TPFIT code comparing with the existing 2-channel air-water mixing experimental results. Secondary, the TPFIT code was applied to simulation of steamwater two-phase flow in modeled two subchannels of current BWRs rod bundle. The fluid mixing was observed at a gap between the subchannels. The existing two-phase flow correlation for fluid mixing is evaluated using detailed numerical simulation data. From the data, pressure difference between fluid channels is responsible for the fluid mixing, and effects of the time averaged and fluctuating pressure difference must be incorporated in the two-phase flow correlation for fluid mixing.

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