Analytic Solution to Verify Code Predictions of Two-Phase Flow in a Boiling Water Reactor Core Channel

1983 ◽  
Vol 62 (3) ◽  
pp. 361-363 ◽  
Author(s):  
Kuo-Fu Chen ◽  
C. A. Olson
Keyword(s):  
Analytic Solution ◽  
Two Phase Flow ◽  
Reactor Core ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Phase Flow ◽  
Two Phase ◽  
Water Reactor

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

Investigation of the two-phase flow in a boiling water reactor using neutron-noise technique

1979 ◽  
Vol 52 (3) ◽  
pp. 357-370 ◽  
Author(s):  
G. Kosály ◽  
Lj. Kostić ◽  
L. Miteff ◽  
G. Varadi ◽  
K. Behringer
Keyword(s):  
Two Phase Flow ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Phase Flow ◽  
Two Phase ◽  
Water Reactor ◽  
Neutron Noise

Analytical Procedure on Fluid Mixing Phenomena in Boiling Water Reactor Core With Advanced Interface Tracking Method

2007 ◽  
Author(s):  
Hiroyuki Yoshida ◽  
Takeharu Misawa ◽  
Kazuyuki Takase ◽  
Hajime Akimoto
Keyword(s):  
Two Phase Flow ◽  
Flow Simulation ◽  
Boiling Water ◽  
Fluid Mixing ◽  
Boiling Water Reactor ◽  
Phase Flow ◽  
Two Phase ◽  
Water Reactor ◽  
Hydraulic Design ◽  
Actual Size

Thermal-hydraulic design of a boiling water reactor (BWR) is performed by correlations with empirical results of actual-size tests. Then, when the reactor of a new design is developed, 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 BWRs using an innovative two-phase flow simulation technology. For this design method, we are developed an advanced interface tracking method that improves fluid volume conservation, to enable high accuracy prediction of two-phase flow fluid mixing phenomena in the fuel bundles. It was incorporated in the detailed two-phase flow simulation code: TPFIT. And the vectorization and parallelization of TPFIT code was conducted to analyze enormous amounts of data. In this study, to verify the TPFIT code performance, the TPFIT code was applied to the air-water and steam-water bubbly two-phase flow in various flow channels and the numerical results were compared with experimental results. Furthermore, the numerical results applied to the fluid mixing phenomena in boiling water reactor rod bundles are shown, and the existing correlations for the fluid mixing phenomena are evaluated by use of these results.

Prediction of Boiling Water Reactor Assembly Void Distributions Using a Two-Phase Computational Fluid Dynamics Model

2008 ◽  
Author(s):  
W. David Pointer ◽  
Adrian Tentner ◽  
Tanju Sofu ◽  
Simon Lo ◽  
Andrew Splawski
Keyword(s):  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
Fuel Assembly ◽  
Two Phase Flow ◽  
Boiling Water ◽  
Initial Assessment ◽  
Boiling Water Reactor ◽  
Phase Flow ◽  
Two Phase ◽  
Water Reactor

This paper presents recent results obtained as part of the on-going integral validation of an advanced Eulerian-Eulerian two-phase (E2P) computational fluid dynamics based boiling model that allows the detailed analysis of the two-phase flow and heat transfer phenomena in a Boiling Water Reactor (BWR) fuel assembly. The code is being developed as a customized module built on the foundation of the commercial CFD-code STAR-CD which provides general two-phase flow modeling capabilities. Simulations of a prototypic BWR fuel assembly experiment have been completed as an initial assessment of the applicability of the E2P model to realistic BWR geometries and conditions. Initial validation has focused on comparison with measured sub-channel averaged data to enable the benchmarking of the accuracy of the E2P against the current predictive capabilities of the sub-channel methods. The paper will discuss the effects of modeling assumptions, assumed coefficient values and the computational mesh structure used to describe the fuel assembly geometry on the accuracy of the sub-channel averaged void fraction.

Development and Validation of a Computational Fluid Dynamics Model for the Simulation of Two-Phase Flow Phenomena in a Boiling Water Reactor Fuel Assembly

2009 ◽  
Author(s):  
Adrian Tentner ◽  
W. David Pointer ◽  
Simon Lo ◽  
Andrew Splawski
Keyword(s):  
Fluid Dynamics ◽  
Computational Fluid Dynamics ◽  
Two Phase Flow ◽  
Operating Conditions ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Phase Flow ◽  
Two Phase ◽  
Water Reactor ◽  
Development And Validation

This paper presents the current status in the development and validation of an advanced Computational Fluid Dynamics (CFD) model, CFD-BWR, which allows the detailed analysis of the two-phase flow and heat transfer phenomena in Boiling Water Reactor (BWR) fuel assemblies under various operating conditions. The CFD-BWR model uses an Eulerian Two-Phase (E2P) approach, and is also referred to as the E2P modeling framework. It is being developed as a customized module built on the foundation of the commercial CFD-code STAR-CD which provides general two-phase flow modeling capabilities. The integral validation efforts have focused on the analysis of the NUPEC Full-Size Boiling Water Reactor Test (BFBT) within the framework of the OECD/NRC benchmark exercise. The paper reviews the two-phase models implemented in the CFD-BWR code, and emphasizes recently implemented models of inter-phase and coolant-cladding momentum and energy exchanges. Results of recent BFBT experiment simulations using these models are presented and the effects of the new models on the calculated void distribution are discussed. The paper concludes with a discussion of future model development and validation plans.

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.

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