A theoretical prediction of critical heat flux in saturated pool boiling during power transients
Understanding and predicting critical heat flux (CHF) behavior during steady-state and transient conditions is of fundamental interest in the design, operation, and safety of boiling and two-phase flow devices. Presented within this paper are the results of a comprehensive theoretical study specifically conducted to model transient CHF behavior in saturated pool boiling. Thermal energy conduction within a heating element and its influence on the CHF are also discussed. The resultant theory provides new insight into the basic physics of the CHF phenomenon and indicates favorable agreement with the experimental data from cylindrical heaters with small radii. However, the flat-ribbon heater data compared poorly with the present theory, although the general trend was predicted. Finally, various factors that affect the discrepency between the data and the theory are listed.
- Research Organization:
- Los Alamos National Laboratory (LANL), Los Alamos, NM (United States); University of Central Florida, Orlando (USA). Coll. of Engineering
- DOE Contract Number:
- W-7405-ENG-36
- OSTI ID:
- 6291531
- Report Number(s):
- LA-UR-87-117; CONF-870816-4; ON: DE87005109
- Resource Relation:
- Conference: 24. national heat transfer conference and exhibition, Pittsburgh, PA, USA, 9 Aug 1987
- Country of Publication:
- United States
- Language:
- English
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Critical heat flux modeling in pool boiling for steady-state and power transients
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Related Subjects
22 GENERAL STUDIES OF NUCLEAR REACTORS
HEAT FLUX
MATHEMATICAL MODELS
POOL BOILING
EXPERIMENTAL DATA
HEAT TRANSFER
HYDRODYNAMIC MODEL
INSTABILITY
PWR TYPE REACTORS
REACTIVITY INSERTIONS
STEADY-STATE CONDITIONS
TRANSIENTS
TWO-PHASE FLOW
BOILING
DATA
ENERGY TRANSFER
FLUID FLOW
INFORMATION
NUMERICAL DATA
PARTICLE MODELS
PHASE TRANSFORMATIONS
REACTIVITY
REACTORS
STATISTICAL MODELS
THERMODYNAMIC MODEL
WATER COOLED REACTORS
WATER MODERATED REACTORS
420400* - Engineering- Heat Transfer & Fluid Flow
220900 - Nuclear Reactor Technology- Reactor Safety