Post critical heat transfer predictions using a modified RELAP5/MOD2 computer code
Journal Article
·
· Nuclear Science and Engineering; (USA)
OSTI ID:6958435
- Texas A and M Univ., College Station, TX (USA). Dept. of Nuclear Engineering
A new modified version of the RELAP5/MOD2 computer code for the analysis of the reflood phase after a hypothetical large-break loss-of-coolant accident is developed. Various rewetting correlations are examined and compared with full-length emergency core heat transfer separate-effects and system-effect test 9FLECHT-SEASET0 experimental reflood data. The RELAP5 prediction of vapor temperatures is low in comparison with the data. The use of a new interfacial heat transfer between droplets and steam results in a reasonable prediction of vapor superheats. A revised dispersed flow film boiling correlation which accounts for the enhancement of steam convective cooling by droplet-induced turbulence, is incorporated in the code. Comparison of the current results with data shows significant improvement in the prediction of clad temperature time histories over previous RELAP5 calculations.
- OSTI ID:
- 6958435
- Journal Information:
- Nuclear Science and Engineering; (USA), Journal Name: Nuclear Science and Engineering; (USA) Vol. 103:1; ISSN 0029-5639; ISSN NSENA
- Country of Publication:
- United States
- Language:
- English
Similar Records
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Conference
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·
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·
OSTI ID:5716804
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Conference
·
Tue Dec 31 23:00:00 EST 1985
· Trans. Am. Nucl. Soc.; (United States)
·
OSTI ID:6907166
Related Subjects
22 GENERAL STUDIES OF NUCLEAR REACTORS
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
ACCIDENTS
COMPUTER CODES
CONVECTION
DATA
ENERGY TRANSFER
EXPERIMENTAL DATA
HEAT TRANSFER
HEATING
INFORMATION
LOSS OF COOLANT
MASS TRANSFER
NUCLEAR SUPERHEATING
NUMERICAL DATA
R CODES
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORES
REACTOR SAFETY EXPERIMENTS
SUPERHEATING
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
ACCIDENTS
COMPUTER CODES
CONVECTION
DATA
ENERGY TRANSFER
EXPERIMENTAL DATA
HEAT TRANSFER
HEATING
INFORMATION
LOSS OF COOLANT
MASS TRANSFER
NUCLEAR SUPERHEATING
NUMERICAL DATA
R CODES
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORES
REACTOR SAFETY EXPERIMENTS
SUPERHEATING