Two-phase flow interfacial drag for once-through steam generators
- Texas A and M Univ., Dept. of Nuclear Engineering, College Station, TX (US)
This paper reports on the current version of the RELAP5/MOD2 computer code which underpredicts the degree of superheat in the secondary side of the steam generator bundles. Many studies have concluded that this is due to overprediction of the interphase drag force. New interphase drag correlations have been developed for the bubbly and slug regimes. These correlations were implemented in the current version of the RELAP5/MOD2 computer code. Steady-state conditions for 65, 75, and 100% power loads of 30-tube once-through steam generator tests are simulated. The calculated primary- and secondary-side temperature profiles show that the new interphase drag correlations achieve closer agreement with experimental data than the temperature profiles of the original code.
- OSTI ID:
- 5081494
- Journal Information:
- Nuclear Technology; (United States), Journal Name: Nuclear Technology; (United States) Vol. 95:1; ISSN 0029-5450; ISSN NUTYB
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
220200 -- Nuclear Reactor Technology-- Components & Accessories
220400* -- Nuclear Reactor Technology-- Control Systems
220900 -- Nuclear Reactor Technology-- Reactor Safety
42 ENGINEERING
420400 -- Engineering-- Heat Transfer & Fluid Flow
ANALOG SYSTEMS
BOILERS
CALCULATION METHODS
COMPUTER CODES
COOLING SYSTEMS
CORRECTIONS
DATA
DRAG
ENERGY SYSTEMS
EXPERIMENTAL DATA
FLUID FLOW
FUNCTIONAL MODELS
INFORMATION
NUMERICAL DATA
ONCE-THROUGH COOLING SYSTEMS
R CODES
REACTOR SAFETY EXPERIMENTS
REACTOR SIMULATORS
ROD BUNDLES
SIMULATORS
STEAM GENERATORS
TEMPERATURE GRADIENTS
TWO-PHASE FLOW
VAPOR GENERATORS