Evaluation of rod bowing effect on critical heat flux in PWR fuel assemblies
Conference
·
OSTI ID:5744716
The purpose of the present study is to evaluate the effect of one of the fuel abnormalities, that of a bowed rod on a Critical Heat Flux (CHF) in Pressurized Water Reactor (PWR) type rod bundles. The results of three experimental programs to determine the effect of rod bowing on CHF conducted at the Heat Transfer Research Facility (HTRF) of Columbia University are presented in this paper. These tests were performed in 4X4 or 5X5 rod bundles simulating a section of a full sized PWR fuel assembly by electrically heating the test sections.
- OSTI ID:
- 5744716
- Report Number(s):
- CONF-861211-
- Country of Publication:
- United States
- Language:
- English
Similar Records
Effect of rod bow to partial closure on critical heat flux in PWR fuel assembly
Evaluations and modifications of the EPRI-1 correlation on PWR critical heat flux predictions under normal and abnormal fuel conditions
Rod bundle critical heat flux at low pressure
Journal Article
·
Fri Jul 01 00:00:00 EDT 1983
· Am. Soc. Mech. Eng., (Pap.); (United States)
·
OSTI ID:6987308
Evaluations and modifications of the EPRI-1 correlation on PWR critical heat flux predictions under normal and abnormal fuel conditions
Journal Article
·
Fri Oct 31 23:00:00 EST 1986
· Nucl. Technol.; (United States)
·
OSTI ID:5112137
Rod bundle critical heat flux at low pressure
Journal Article
·
Fri Nov 30 23:00:00 EST 1979
· Nucl. Technol.; (United States)
·
OSTI ID:5533006
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BENDING
CORRELATIONS
CRITICAL HEAT FLUX
DATA ANALYSIS
FUEL ASSEMBLIES
FUEL ELEMENTS
FUEL RODS
HEAT FLUX
HEATING
HIGH PRESSURE
PRESSURE EFFECTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
ROD BUNDLES
TEMPERATURE EFFECTS
TESTING
THERMODYNAMICS
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BENDING
CORRELATIONS
CRITICAL HEAT FLUX
DATA ANALYSIS
FUEL ASSEMBLIES
FUEL ELEMENTS
FUEL RODS
HEAT FLUX
HEATING
HIGH PRESSURE
PRESSURE EFFECTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
ROD BUNDLES
TEMPERATURE EFFECTS
TESTING
THERMODYNAMICS
WATER COOLED REACTORS
WATER MODERATED REACTORS