Annular burnout data from rod-bundle experiments. [PWR]

Conference · Sat Jan 01 00:00:00 EST 1983

OSTI ID:6546118

Yoder, G L; Morris, D G; Mullins, C B

Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident.

• Research Organization: Oak Ridge National Lab., TN (USA)
• DOE Contract Number: W-7405-ENG-26
• OSTI ID: 6546118
• Report Number(s): CONF-830103-28; ON: DE83006137
• Resource Relation: Conference: 2. international topical meeting on nuclear reactor thermal hydraulics, Santa Barbara, CA, USA, 11 Jan 1983; Other Information: Portions are illegible in microfiche products
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FUEL RODS
BURNOUT
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
PWR TYPE REACTORS
FLOW RATE
FUEL ASSEMBLIES
PRESSURE GRADIENTS
REACTOR SAFETY
TEST FACILITIES
ACCIDENTS
ENERGY TRANSFER
FLUID MECHANICS
FUEL ELEMENTS
MECHANICS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
SAFETY
WATER COOLED REACTORS
WATER MODERATED REACTORS
220900* - Nuclear Reactor Technology- Reactor Safety
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled