Effect of rod surface thermocouples on transient critical heat flux
Conference
·
OSTI ID:6558502
The purpose of the test program was to determine the effect of cladding surface thermocouples on time-to-critical heat flux (CHF) under blowdown conditions similar to those in the Loss-of-Fluid Test (LOFT) facility during a Loss-of-Coolant Experiment (LOCE). Previous steady state CHF tests indicated that cladding surface thermocouples reduce the CHF over the pressure range of 13.8 to 16.5 MPa, but have an insignificant effect on CHF near 11 MPa, where CHF would occur in pressurized water reactors (PWR) and in the LOFT facility during a blowdown. The LOFT facility is an integral nuclear reactor test facility which has been designed to simulate, as nearly as possible, all of the important effects that are anticipated to occur during a Loss-of-Coolant Accident (LOCA) in a PWR-type nuclear steam supply system. The LOFT facility utilizes special experimental instrumentation which has been located to measure the significant interacting nuclear and thermal-hydraulic process required to assess LOCA and Emergency Core Coolant System (ECCS) performance. The in-core experimental instrumentation includes thermocouples attached to the outside surfaces of the fuel rod cladding.
- Research Organization:
- EG and G Idaho, Inc., Idaho Falls (USA)
- DOE Contract Number:
- EY-76-C-07-1570
- OSTI ID:
- 6558502
- Report Number(s):
- CONF-781022-20
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
BLOWDOWN
CRITICAL HEAT FLUX
FUEL ASSEMBLIES
FUEL ELEMENT CLUSTERS
HEAT FLUX
LOFT REACTOR
MEASURING INSTRUMENTS
PWR TYPE REACTORS
REACTORS
RESEARCH AND TEST REACTORS
SIMULATION
TANK TYPE REACTORS
TEST REACTORS
THERMOCOUPLES
TRANSIENTS
WATER COOLED REACTORS
WATER MODERATED REACTORS
22 GENERAL STUDIES OF NUCLEAR REACTORS
220600 -- Nuclear Reactor Technology-- Research
Test & Experimental Reactors
220900* -- Nuclear Reactor Technology-- Reactor Safety
BLOWDOWN
CRITICAL HEAT FLUX
FUEL ASSEMBLIES
FUEL ELEMENT CLUSTERS
HEAT FLUX
LOFT REACTOR
MEASURING INSTRUMENTS
PWR TYPE REACTORS
REACTORS
RESEARCH AND TEST REACTORS
SIMULATION
TANK TYPE REACTORS
TEST REACTORS
THERMOCOUPLES
TRANSIENTS
WATER COOLED REACTORS
WATER MODERATED REACTORS