Simulation of Void Fraction Profile Evolution in Subcooled Nucleate Boiling in a Vertical Annulus with a Bubble-Tracking Model

2004 ◽  
Author(s):  
Ivo Kljenak ◽  
Borut Mavko
Keyword(s):  
Void Fraction ◽  
Nucleate Boiling ◽  
Tracking Model ◽  
Vertical Annulus

Simulation of Subcooled Nucleate Boiling in a Vertical Annulus With Coupling of Bubble-Tracking and Two-Fluid Models: Further Comparison With Experimental Results

2006 ◽  
Author(s):  
Ivo Kljenak ◽  
Bosˇtjan Koncˇar ◽  
Borut Mavko
Keyword(s):  
Fluid Model ◽  
Three Dimensional ◽  
Vertical Channel ◽  
Nucleate Boiling ◽  
Bubble Nucleation ◽  
Experimental Results ◽  
Tracking Model ◽  
Two Fluid Model ◽  
Vertical Annulus ◽  
Two Fluid

A model of subcooled nucleate boiling flow in a vertical channel at low pressure conditions that was developed earlier is further validated by comparison of simulations with experimental results from Purdue University (USA). The model consists of a three-dimensional bubble-tracking model and a two-dimensional two-fluid model which are coupled off-line. By taking into account dynamic phenomena (liquid flow, bubble motion and interaction) and thermal phenomena (liquid heating, bubble nucleation and condensation), the model is able to simulate the gradual evolution of void fraction profiles along the boiling channel.

Downcomer Boiling During LBLOCA with Direct Vessel Injection of APR1400

2006 ◽  
Author(s):  
Eu Hwak Lee ◽  
Hee Cheon No ◽  
Dong Won Lee ◽  
Chul Hwa Lee
Keyword(s):  
Heat Flux ◽  
Void Fraction ◽  
Nucleate Boiling ◽  
Injection System ◽  
Heated Wall ◽  
Local Phase ◽  
Transfer Coefficients ◽  
Phase Velocities ◽  
Drift Flux ◽  
Nucleate Boiling Heat Transfer

An experimental study has been performed to investigate thermal-hydraulic phenomena in the downcomer during LBLOCA with Direct Vessel Injection (DVI), which is a new Safety Injection System (SIS) of the Advanced Power Reactor 1400 MW (APR1400). In order to understand the downcomer boiling phenomena experimentally, the KAIST Downcomer Boiling experiment has been performed with the same as the thickness of the reactor vessel of APR1400 (21 cm) and a height of 150 cm to observe downcomer boiling phenomena and to measure local parameters such as local phase velocities, local void fraction, collapsed water level and heat flux from the heated wall. From the KAIST Downcomer Boiling experiment, we visually observed strong liquid recirculation and vapor jetting near the heated wall due to the axial migration of voids only in the thin layer (about 4 cm) of the heated wall but little bubble migration out of the bubble region. Local phase velocities and local void fraction were measured to estimate the drift-flux parameters in the downcomer channel. Heat flux from the heated side was back-calculated to find the CHF and to estimate nucleate boiling heat transfer coefficients.

Nucleate Boiling Flow Simulation With Coupling of Eulerian and Lagrangian Methods

2005 ◽  
Author(s):  
Bosˇtjan Koncˇar ◽  
Ivo Kljenak ◽  
Borut Mavko
Keyword(s):  
Model Simulation ◽  
Fluid Model ◽  
Nucleate Boiling ◽  
Homogeneous Distribution ◽  
Lagrangian Simulation ◽  
Radial Profiles ◽  
Boiling Flow ◽  
Tracking Model ◽  
Two Fluid Model ◽  
Two Fluid

Subcooled boiling flow was simulated by combining the two-fluid model of the CFX-4.4 code and a Lagrangian bubble-tracking model. At present, both models are coupled “off-line” via the local bubble Sauter diameter. The two-fluid model simulation with the CFX-4.4 code provides local values of turbulent kinetic energy field of the liquid phase, which is used as an input for the bubble-tracking model. In the bubble-tracking model, vapour is distributed in the liquid in the form of individually tracked bubbles. The result of the Lagrangian simulation is a non-homogeneous distribution of local Sauter diameter, which is used in the two-fluid model to predict the interfacial forces and interfacial transfer rates of mass and heat transfer. The coupled approach requires a few iterations to obtain a converged solution. The results of the proposed approach were validated against boiling flow experiments from the literature. A good agreement between measured and calculated radial profiles of void fraction and bubble diameter was obtained.

DNB Prediction Method Using Local Void Fraction Applied to a PWR Rod Bundle With Mixing Vane Grid

2012 ◽  
Author(s):  
Tatsuya Sasakawa ◽  
Tsutomu Ikeno ◽  
Isao Kataoka
Keyword(s):  
Void Fraction ◽  
New Technology ◽  
Prediction Method ◽  
Nucleate Boiling ◽  
Two Phase ◽  
Average Value ◽  
Rod Bundle ◽  
Wide Range ◽  
Channel Analysis ◽  
Reasonable Prediction

A new code coupled between a sub-channel analysis code and a computational multi-fluid dynamics (CMFD) code was applied to a PWR rod bundle with mixing vane grid. The code was developed to predict the departure from nucleate boiling (DNB). This is a new technology of CMFD based on abundant experience and models developed for sub-channel analysis: CMFD computed void distribution to fit the average value calculated by two-phase models in sub-channel analysis code. A new source term represented centripetal motion of small bubbles in the wake behind a rising vapor slug. In order to apply the code to a PWR rod bundle, effects of local mass flux, pressure and mixing vane grid was modeled. The present method was applied to the analysis for DNB tests of simulated PWR fuel assembly with mixing vane grids. The result showed that, using only a critical void fraction, reasonable prediction was achieved in a wide range of flow condition and a variety of flow regimes.

Modeling of Two-Phase Flow Structure Evolution in Subcooled Nucleate Convective Boiling With Coupling of Bubble-Tracking and Two-Fluid Models

2008 ◽  
Author(s):  
Ivo Kljenak ◽  
Bosˇtjan Koncˇar ◽  
Luka Sˇtrubelj ◽  
Borut Mavko
Keyword(s):  
Fluid Model ◽  
Three Dimensional ◽  
Vertical Channel ◽  
Nucleate Boiling ◽  
Bubble Nucleation ◽  
Structure Evolution ◽  
Two Phase ◽  
Convective Boiling ◽  
Tracking Model ◽  
Two Fluid

A model of subcooled nucleate boiling flow in a vertical channel at low-pressure conditions is proposed. The model consists of a three-dimensional bubble-tracking model and a two-dimensional two-fluid model which are coupled off-line. By taking into account dynamic phenomena (liquid flow, bubble motion and interaction) and thermal phenomena (liquid heating, bubble nucleation and condensation), the model is able to simulate the gradual evolution of void fraction profiles along the boiling channel. The model is assessed by simulating experiments that were performed at Purdue University (USA) at atmospheric pressure in a vertical annulus with a central heating rod.

Optical Diagnostics of Nucleate Boiling From Transparent Heated Surfaces

2003 ◽  
Author(s):  
Anton Zimmermann ◽  
Adrian M. Holland ◽  
Colin P. Garner
Keyword(s):  
Void Fraction ◽  
Indium Tin Oxide ◽  
High Speed ◽  
Imaging System ◽  
Nucleate Boiling ◽  
Nucleation Site ◽  
Heat Fluxes ◽  
Diagnostic Methods ◽  
Frame Rate ◽  
Site Density

Indium Tin Oxide (ITO) coated glass was used to provide transparent heated surfaces with heat fluxes high enough to generate nucleate boiling in water. The technique enables extended horizontal surfaces exhibiting nucleate boiling to be analysed with novel optical diagnostic methods. A horizontal glass substrate coated with an ITO layer on its top surface was immersed in demineralised water of temperatures between 70 and 80°C. A direct electrical current was passed through the ITO to heat the water. A high-speed imaging system comprising an infrared laser and CCD camera was used to analyse the resulting nucleate pool boiling from the ITO surface. This system was operated at up to 1 kHz frame rate and the bubbles analysed in terms of size and shape. Statistical data regarding bubble size and nucleation site density were obtained for heat fluxes ranging from 63 to 105 kW.m−2. Nucleation site densities were found to be up to 35 000 sites.m−2. Furthermore, non-intrusive cross-sectional void fraction measurements were made, and ranged from zero to 14% of surface area. The increase in both site density and void fraction with increasing heat flux was found to be in good agreement with published literature.

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