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Coupled nuclear-thermal hydraulic code for prediction of core behavior during steam line break transient. [PWR]

Conference · · Trans. Am. Nucl. Soc.; (United States)
OSTI ID:6877199
;
None
Research Organization:
Westinghouse Electric Corp., Pittsburgh, PA
OSTI ID:
6877199
Report Number(s):
CONF-800607-
Conference Information:
Journal Name: Trans. Am. Nucl. Soc.; (United States) Journal Volume: 34
Country of Publication:
United States
Language:
English

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Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
COMPUTER CALCULATIONS
ENERGY TRANSFER
FAILURES
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
MATHEMATICAL MODELS
MECHANICS
PWR TYPE REACTORS
REACTOR SAFETY
REACTORS
SAFETY
STEAM LINES
TRANSIENTS
WATER COOLED REACTORS
WATER MODERATED REACTORS