Core thermal response during Semiscale Mod-1 blowdown heat transfer tests. [PWR]
Selected experimental data and results calculated from experimental data obtained from the Semiscale Mod-1 PWR blowdown heat transfer test series are analyzed. These tests were designed primarily to provide information on the core thermal response to a loss-of-coolant accident. The data are analyzed to determine the effect of core flow on the heater rod thermal response. The data are also analyzed to determine the effects of initial operating conditions on the rod cladding temperature behavior during the transient. The departure from nucleate boiling and rewetting characteristics of the rod surfaces are examined for radial and axial patterns in the response. Repeatability of core thermal response data is also investigated. The test data and the core thermal response calculated with the RELAP4 code are compared.
- Research Organization:
- Idaho National Engineering Lab., Idaho Falls (USA)
- DOE Contract Number:
- E(10-1)-1375
- OSTI ID:
- 7274438
- Report Number(s):
- ANCR-NUREG-1285
- Country of Publication:
- United States
- Language:
- English
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ACCIDENTS
BLOWDOWN
ENERGY TRANSFER
FUEL ELEMENTS
HEAT TRANSFER
LOSS OF COOLANT
MOCKUP
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORES
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STRUCTURAL MODELS
TEMPERATURE DISTRIBUTION
TRANSIENTS
WATER COOLED REACTORS
WATER MODERATED REACTORS