A general unified non-equilibrium model for predicting saturated and subcooled critical two-phase flow rates through short and long tubes

Technical Report · Fri Sep 01 04:00:00 EDT 1995

DOI:https://doi.org/10.2172/106995· OSTI ID:106995

Fraser, D. W.H. ^[1]; Abdelmessih, A. H. ^[2]

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1. Univ. of British Columbia (Canada)
2. Univ. of Toronto, Ontario (Canada)

A general unified model is developed to predict one-component critical two-phase pipe flow. Modelling of the two-phase flow is accomplished by describing the evolution of the flow between the location of flashing inception and the exit (critical) plane. The model approximates the nonequilibrium phase change process via thermodynamic equilibrium paths. Included are the relative effects of varying the location of flashing inception, pipe geometry, fluid properties and length to diameter ratio. The model predicts that a range of critical mass fluxes exist and is bound by a maximum and minimum value for a given thermodynamic state. This range is more pronounced at lower subcooled stagnation states and can be attributed to the variation in the location of flashing inception. The model is based on the results of an experimental study of the critical two-phase flow of saturated and subcooled water through long tubes. In that study, the location of flashing inception was accurately controlled and adjusted through the use of a new device. The data obtained revealed that for fixed stagnation conditions, the maximum critical mass flux occurred with flashing inception located near the pipe exit; while minimum critical mass fluxes occurred with the flashing front located further upstream. Available data since 1970 for both short and long tubes over a wide range of conditions are compared with the model predictions. This includes test section L/D ratios from 25 to 300 and covers a temperature and pressure range of 110 to 280{degrees}C and 0.16 to 6.9 MPa. respectively. The predicted maximum and minimum critical mass fluxes show an excellent agreement with the range observed in the experimental data.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology; American Nuclear Society, La Grange Park, IL (United States); American Inst. of Chemical Engineers, New York, NY (United States); American Society of Mechanical Engineers, New York, NY (United States); Canadian Nuclear Society, Toronto, ON (Canada); Japan Society of Multiphase Flow, Kyoto (Japan)
• OSTI ID: 106995
• Report Number(s): NUREG/CP--0142-Vol.1; CONF-950904--Vol.1; ON: TI95017077; CNN: Grant OGP-0004172
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
42 ENGINEERING
COMPUTERIZED SIMULATION
CRITICAL FLOW
EXPERIMENTAL DATA
FLOW MODELS
HYDRAULICS
SATURATION
SUBCOOLING
THEORETICAL DATA
TUBES
TWO-PHASE FLOW