HOTROD: a computer program for subchannel analysis of coolant flow in rod bundles (LWBR Development Program)
Technical Report
·
OSTI ID:7010383
A digital computer program is described for the steady-state thermal and hydraulic subchannel analysis of coolant flow in rod bundle reactor cores and heated-rod experimental test sections. The program allows for the transfer of two-phase fluid properties in three dimensions and predicts the local fluid conditions, fuel rod surface temperature and critical heat flux. The history of subchannel analysis in rod bundles is discussed. A detailed description is presented of the HOTROD analytical model, the numerical solution procedures and the program input formats. 29 references. (NSA 29: 17559)
- Research Organization:
- Bettis Atomic Power Lab., Pittsburgh, PA (USA)
- DOE Contract Number:
- AT(11-1)-GEN-14
- OSTI ID:
- 7010383
- Report Number(s):
- WAPD-TM-1070
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210500* -- Power Reactors
Breeding
BREEDER REACTORS
COMPUTER CODES
CRITICAL HEAT FLUX
ENERGY TRANSFER
FLUID FLOW
FLUID MECHANICS
FUEL ASSEMBLIES
FUEL CHANNELS
FUEL ELEMENT CLUSTERS
H CODES
HEAT FLUX
HEAT TRANSFER
HYDRAULICS
LWBR TYPE REACTORS
MECHANICS
REACTOR CHANNELS
REACTOR COMPONENTS
REACTORS
ROD BUNDLES
THERMAL REACTORS
TWO-PHASE FLOW
WATER COOLED REACTORS
WATER MODERATED REACTORS
210500* -- Power Reactors
Breeding
BREEDER REACTORS
COMPUTER CODES
CRITICAL HEAT FLUX
ENERGY TRANSFER
FLUID FLOW
FLUID MECHANICS
FUEL ASSEMBLIES
FUEL CHANNELS
FUEL ELEMENT CLUSTERS
H CODES
HEAT FLUX
HEAT TRANSFER
HYDRAULICS
LWBR TYPE REACTORS
MECHANICS
REACTOR CHANNELS
REACTOR COMPONENTS
REACTORS
ROD BUNDLES
THERMAL REACTORS
TWO-PHASE FLOW
WATER COOLED REACTORS
WATER MODERATED REACTORS