CHF experiment and CFD analysis in a 2 × 3 rod bundle with mixing vane

Nuclear Engineering and Design

Volumn 239, Issue 5, 2009, Pages 899-912

  Shin, Byung Soo ^a   Chang, Soon Heung ^a  

^a Korea Advanced Institute of Science and Technology (KAIST)   (South Korea)

Author keywords

[No Author keywords available]

Indexed keywords

BOILING LIQUIDS; EXPERIMENTS; HEAT FLUX; PRESSURIZED WATER REACTORS; RELIABILITY ANALYSIS; SMALL NUCLEAR REACTORS; TABLE LOOKUP; TESTING; TWO PHASE FLOW;

AVERAGE VOID FRACTIONS; BOILING ANALYSIS; CFD ANALYSIS; CHF ENHANCEMENTS; CORE REGIONS; FLOW CHARACTERISTICS; FUEL RODS; HEATED LENGTHS; LATERAL MOTIONS; LATERAL VELOCITIES; LOW PRESSURES; LOWER PRESSURES; MASS FLUXES; MIXING VANES; NEAR WALLS; NEAR-WALL REGIONS; OPERATING CONDITIONS; PHASE PARAMETERS; PWR FUELS; ROD BUNDLES; SUBCOOLED BOILING FLOWS; SWIRL VANES; TEST SECTIONS; WORKING FLUIDS;

VOID FRACTION;

EID: 63349095485     PISSN: 00295493     EISSN: None     Source Type: Journal    
DOI: 10.1016/j.nucengdes.2009.01.011     Document Type: Article

References (18)

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