Experimental studies of the air coolability of triga reactors following a loss-of-coolant accident
Journal Article
·
· Nucl. Technol.; (United States)
OSTI ID:6614009
To investigate the coolability of a uniformly heated tube by free convection of atmospheric air, heat transfer experiments were conducted using vertical open annuli with adiabatic outer walls. To examine the effect of the annulus ratio on the coolability of the heated tube, the experiments employed four annuli (diameter ratios of 1.155, 1.33, 1.63, and 12.0). The operating parameters included heat fluxes up to 1.38 W/cm/sup 2/ with a corresponding surface temperature of 856 K. The results, extrapolated to 1200 K, were used to provide a qualitative estimate of the coolability of multirod bundles, as a function of the equilibrium surface temperature and the pitch-to-diameter (P/D) ratio.
- Research Organization:
- The Univ. of New Mexico, Dept. of Chemical and Nuclear Engineering, Albuquerque, NM 87131
- OSTI ID:
- 6614009
- Journal Information:
- Nucl. Technol.; (United States), Vol. 76:3
- Country of Publication:
- United States
- Language:
- English
Similar Records
On the air coolability of TRIGA reactors following a loss-of-coolant accident
Air coolability of TRIGA reactors following a loss-of-coolant accident
ANALYTICAL STUDY OF HEAT TRANSFER RATES FOR PARALLEL FLOW OF LIQUID METALS THROUGH TUBE BUNDLES. PART I
Conference
·
Tue Jul 01 00:00:00 EDT 1986
·
OSTI ID:6614009
+3 more
Air coolability of TRIGA reactors following a loss-of-coolant accident
Conference
·
Wed Jan 01 00:00:00 EST 1986
·
OSTI ID:6614009
+3 more
ANALYTICAL STUDY OF HEAT TRANSFER RATES FOR PARALLEL FLOW OF LIQUID METALS THROUGH TUBE BUNDLES. PART I
Journal Article
·
Fri Jan 01 00:00:00 EST 1960
· Chem. Eng. Progr.
·
OSTI ID:6614009
Related Subjects
22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ROD BUNDLES
COOLING
TRIGA TYPE REACTORS
AFTER-HEAT REMOVAL
LOSS OF COOLANT
ADIABATIC PROCESSES
AIR
HEAT FLUX
HEAT TRANSFER
HIGH TEMPERATURE
INCLINATION
NATURAL CONVECTION
POWER RANGE 10-100 KW
RESEARCH PROGRAMS
SURFACE PROPERTIES
VERY HIGH TEMPERATURE
WALLS
ACCIDENTS
CONVECTION
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
FLUIDS
GASES
HOMOGENEOUS REACTORS
HYDRIDE MODERATED REACTORS
MASS TRANSFER
REACTOR ACCIDENTS
REACTORS
REMOVAL
RESEARCH AND TEST REACTORS
SOLID HOMOGENEOUS REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
ROD BUNDLES
COOLING
TRIGA TYPE REACTORS
AFTER-HEAT REMOVAL
LOSS OF COOLANT
ADIABATIC PROCESSES
AIR
HEAT FLUX
HEAT TRANSFER
HIGH TEMPERATURE
INCLINATION
NATURAL CONVECTION
POWER RANGE 10-100 KW
RESEARCH PROGRAMS
SURFACE PROPERTIES
VERY HIGH TEMPERATURE
WALLS
ACCIDENTS
CONVECTION
ENERGY TRANSFER
ENRICHED URANIUM REACTORS
FLUIDS
GASES
HOMOGENEOUS REACTORS
HYDRIDE MODERATED REACTORS
MASS TRANSFER
REACTOR ACCIDENTS
REACTORS
REMOVAL
RESEARCH AND TEST REACTORS
SOLID HOMOGENEOUS REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled