THORAX pretest prediction of a sodium-boiling transient in a 19-pin simulated LMFBR driver bundle

Conference · Fri Jan 01 00:00:00 EST 1982

OSTI ID:6756920

Rose, S D

Experiments will be conducted in the Thermal-Hydraulic Out-of-Reactor Safety-Shutdown Heat Removal System (THORS-SHRS) Assembly 1 loop at Oak Ridge National Laboratory (ORNL) to model the behavior of a reactor during degraded decay heat removal conditions. The test section is to consist of two parallel 19-pin electrically-heated driver bundles, typical of U.S. Large Developmental Plant (LDP) Liquid Metal Fast Breeder Reactor (LMFBR) design. Analysis of these experiments will include using THORAX, a two-dimensional boiling model which assumes an equilibrium mixture two-phase flow (with slip). A THORAX prediction is presented for a single-bundle forced convection boiling-to-dryout transient at 15.8 kW/pin.

• Research Organization: Oak Ridge National Lab., TN (USA)
• DOE Contract Number: W-7405-ENG-26
• OSTI ID: 6756920
• Report Number(s): CONF-821068-1; ON: DE83001943
• Resource Relation: Conference: 10. liquid metal boiling working group meeting, Karlsruhe, F.R. Germany, 27 Oct 1982; Other Information: Portions of document are illegible
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
AFTER-HEAT REMOVAL
HEAT TRANSFER
HYDRAULICS
FUEL ASSEMBLIES
LMFBR TYPE REACTORS
LOSS OF FLOW
AFTER-HEAT
COMPUTER CALCULATIONS
REACTOR SAFETY
TEST FACILITIES
TWO-PHASE FLOW
ACCIDENTS
BREEDER REACTORS
ENERGY TRANSFER
EPITHERMAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FLUID FLOW
FLUID MECHANICS
LIQUID METAL COOLED REACTORS
MECHANICS
REACTOR ACCIDENTS
REACTORS
REMOVAL
SAFETY
220900* - Nuclear Reactor Technology- Reactor Safety
210500 - Power Reactors
Breeding