Analytical modeling of core hydraulics and flow management in breeder reactors
Conference
·
OSTI ID:5803769
An analytical model representing the hydraulic behavior of the primary system of fast breeder nuclear reactors is discussed. A computer code capable of detailing the core flow distribution and characterizing the flow and pressure drop in each reactor component is presented. Application of this method to the reactor core thermal-hydraulic design has allowed optimization of the flow management with consequent upgrading in performance, reduction of unnecessary conservatism and very substantial cost savings. Typical quantitative examples are presented.
- Research Organization:
- Westinghouse Electric Corp., Madison, PA (USA). Advanced Reactors Div.
- DOE Contract Number:
- EY-76-C-15-2395
- OSTI ID:
- 5803769
- Report Number(s):
- CONF-790934-1
- 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
COOLING SYSTEMS
EPITHERMAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FLUID MECHANICS
HYDRAULICS
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
MATHEMATICAL MODELS
MECHANICS
PRIMARY COOLANT CIRCUITS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR CORES
REACTORS
210500* -- Power Reactors
Breeding
BREEDER REACTORS
COMPUTER CODES
COOLING SYSTEMS
EPITHERMAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
FLUID MECHANICS
HYDRAULICS
LIQUID METAL COOLED REACTORS
LMFBR TYPE REACTORS
MATHEMATICAL MODELS
MECHANICS
PRIMARY COOLANT CIRCUITS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR CORES
REACTORS