Generalized drift-flux correlation
Conference
·
· Transactions of the American Nuclear Society; (United States)
OSTI ID:5977840
- Westinghouse Electric Corp., Pittsburgh, PA (United States)
A one-dimensional drift-flux model with five conservation equations is frequently employed in major computer codes, such as TRAC-PD2, and in simulator codes. In this method, the relative velocity between liquid and vapor phases, or slip ratio, is given by correlations, rather than by direct solution of the phasic momentum equations, as in the case of the two-fluid model used in TRAC-PF1. The correlations for churn-turbulent bubbly flow and slug flow regimes were given in terms of drift velocities by Zuber and Findlay. For the annular flow regime, the drift velocity correlations were developed by Ishii et al., using interphasic force balances. Another approach is to define the drift velocity so that flooding and liquid hold-up conditions are properly simulated, as reported here. The generalized correlation is used to reanalyze the MB-2 test data for two-phase flow in a large-diameter pipe. The results are applied to the generalized drift flux velocity, whose relationship to the other correlations is discussed. Finally, the generalized drift flux correlation is implemented in TRAC-PD2. Flow reversal from countercurrent to cocurrent flow is computed in small-diameter U-shaped tubes and is compared with the flooding curve.
- OSTI ID:
- 5977840
- Report Number(s):
- CONF-910603--
- Conference Information:
- Journal Name: Transactions of the American Nuclear Society; (United States) Journal Volume: 63
- Country of Publication:
- United States
- Language:
- English
Similar Records
Generalized drift flux correlation for vertical flow
Improved Accuracy for Two-Phase Downflow Scenarios
Drift flux model as approximation of two fluid model for two phase dispersed and slug flow in tube
Journal Article
·
Thu Oct 01 00:00:00 EDT 1992
· Nuclear Science and Engineering; (United States)
·
OSTI ID:7024036
Improved Accuracy for Two-Phase Downflow Scenarios
Conference
·
Mon Oct 01 00:00:00 EDT 2012
·
OSTI ID:1060995
Drift flux model as approximation of two fluid model for two phase dispersed and slug flow in tube
Conference
·
Fri Sep 01 00:00:00 EDT 1995
·
OSTI ID:106989
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
42 ENGINEERING
420400 -- Engineering-- Heat Transfer & Fluid Flow
ACCIDENTS
BUBBLES
COMPUTER CODES
COMPUTERIZED SIMULATION
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
FLOW MODELS
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
INTERFACES
MATHEMATICAL MODELS
MECHANICS
ONE-DIMENSIONAL CALCULATIONS
POWER REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
SIMULATION
T CODES
TEST FACILITIES
THERMAL REACTORS
TWO-PHASE FLOW
VOID FRACTION
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
42 ENGINEERING
420400 -- Engineering-- Heat Transfer & Fluid Flow
ACCIDENTS
BUBBLES
COMPUTER CODES
COMPUTERIZED SIMULATION
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
FLOW MODELS
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
INTERFACES
MATHEMATICAL MODELS
MECHANICS
ONE-DIMENSIONAL CALCULATIONS
POWER REACTORS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
SIMULATION
T CODES
TEST FACILITIES
THERMAL REACTORS
TWO-PHASE FLOW
VOID FRACTION
WATER COOLED REACTORS
WATER MODERATED REACTORS