Group Title: Innovative Nuclear Space Power and Propulsion Institute informational brochures
Title: Computational fluid dynamics
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 Material Information
Title: Computational fluid dynamics
Series Title: Innovative Nuclear Space Power and Propulsion Institute informational brochures
Physical Description: Archival
Language: English
Creator: Innovative Nuclear Space Power and Propulsion Institute, University of Florida
Publisher: Innovative Nuclear Space Power and Propulsion Institute, University of Florida
Place of Publication: Gainesville, Fla.
 Record Information
Bibliographic ID: UF00091281
Volume ID: VID00004
Source Institution: University of Florida
Holding Location: University of Florida
Rights Management: All rights reserved by the source institution and holding location.


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"Space Exploration is the ultimate
investment in America's Future"
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An International Leader
in Space Applications

Contact Information
General: Ms. Lynne Schreiber,
Research: Dr. Travis Knight,
Academic: Ms. Ines Aviles-Spadoni,
P.O. Box 116502
Gainesville, FL 32611-6502
Phone: (352) 392-1427
FAX: (352) 392-8656



Schematic Representation of the Energy-Of-Flow
(EOF) Approach with Cell Mesh Layout for Simulation
of Two-Phase Flow and Boiling Heat Transfer






0 D.5 1

The Innovative Nuclear Space Power and
Propulsion Institute (INSPI) at the University of
Florida has a well-established academic CFD team
for advanced CFD and heat transfer research. The
strength of the CFD team is in two-phase flow and
boiling heat transfer in LWR hydraulics fields.
Currently, this team is developing an ultra fine
mesh two-phase computational fluid dynamic
model, Energy-Of-Flow (EOF), for LWR safety
The researchers of the CFD team also make
considerable use of commercially available CFD
software such as Fluent, CFX, STAR-CD, and
FLOW-3D to predict flow and temperature fields in
many complex conditions and to compare with the
results calculated by in-house CFD programs. The
other task for the CFD team is to justify the
empirical correlations used by well-known
computational thermal hydraulic system simulation
codes such as RELAP5, RETRAN, TRAC, and
CATHARE and subchannel analysis codes like VIPRE
and COBRA.







Single and


A critical issue in safety assessment is the thermal hydraulic
performance of a reactor core under normal and abnormal
operational transients. In particular, the numerical simulation of
micro-scale two-phase flow and heat transfer problems play an
increasingly important role in the safety analysis of water-
cooled nuclear reactors. Over the past two decades, several
algorithms have been developed to directly simulate two-phase
flow systems. These include the two-fluid model, the step-
function approach, the Volume of Flow (VOF) method, the Level
Set method, the Second gradient method, the Lattice Boltzmann
method, and the Front Tracking method. Due to a variety of
limitations, a high-efficiency and high-resolution two-phase
flow model is still not developed.

Film Boiling Bubble Growth and Collapse
under Sub-cooled Flow Condition

Two-Phase Flow
with Wall Heat
Transfer in
Vertical Tube

Post-CMF Film Boiling

Thermionic Fuel

The EOF model is a high-resolution two-phase computational
fluid dynamic approach for analysis of fluid flow and heat transfer
with phase change. This novel model is based on the three-
dimensional Navier-Stokes equations that are directly solved for
the entire two-phase field. The Boussinesq approximation is used
to generate conventional buoyancy force and body force for bulk
fluid. The Clausius-Clapeyron equation is utilized to describe the
relation between saturation pressure and temperature. A novel
energy-based transport equation is employed to track and
delineate the dynamic liquid-vapor interfacial boundary. The
need for temporal and spatial averaging is completely eliminated.
The geometrical void fraction in this formulation is replaced by a
dynamic vapor-phase fraction, which identifies the heat transfer
regimes in the two-phase flow system. Preliminary results have
demonstrated the computational efficiency and the applicability
of the CFD model to a variety of two-phase flow and heat transfer
problems of interest to light water reactor (LWR) safety, such as
Loss-of-Coolant Accident (LOCA), Critical Heat Flux (CHF),
Departure-from-Nucleate Boiling (DNB) and Dryout.
The significant advance made in the developing model is to
avoid a pseudo-singularity like behavior in the Navier-Stokes
equations which describes two-phase fluid and thermal
conditions. Due to the order of magnitude type change in flow
properties between the bubble and its surrounding water,
simulation of flow boiling is a challenging problem. In particular,
buoyancy force and surface tension which play a pivotal role in
the boiling nucleation make the governing equations rather
complicated. The EOF method adopted mathematical smoothing
techniques to deal with the effects of latent heat, buoyancy force,
surface tension and shear stress discontinuities at the water-
steam interface. This EOF approach achieves high-accuracy,
high-stability and easy-convergence for complicated two-phase
flow and heat transfer simulation.

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