Evaluation of the PBF LOFT lead rod test results concerning surface thermocouple perturbation effects

Technical Report · Fri Feb 08 04:00:00 EST 1980

DOI:https://doi.org/10.2172/5603747· OSTI ID:5603747

Carboneau, M. L.; Tolman, E. L.

The purpose of the PBF LOFT Lead Rod (LLR) Test program was to provide experimental data to characterize the mechanical behavior of LOFT type nuclear fuel rods under loss of coolant accident (LOCA) conditions, simulating the test conditions expected for the LOFT Power Ascension (L2) Test series. Although the LLR tests were not explicitly designed to evaluate cladding surface thermocouple perturbation effects, comparison of the Linear Variable Differential Transformer (LVDT) data for rods instrumented with and without cladding thermocouples provided pertinent information concerning the effects of cladding thermocouples on the time to DNB and time to quench data. Documentation and review of this data is presented.

• Research Organization: EG and G Idaho, Inc., Idaho Falls (USA)
• DOE Contract Number: EY-76-C-07-1570
• OSTI ID: 5603747
• Report Number(s): EGG/LTR-LO-00-79-108
• Country of Publication: United States
• Language: English

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
ACCIDENTS
ACCURACY
ENERGY TRANSFER
FLUID MECHANICS
FUEL ELEMENTS
FUEL RODS
HEAT TRANSFER
HYDRAULICS
LOFT REACTOR
LOSS OF COOLANT
MEASURING INSTRUMENTS
MECHANICS
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR SAFETY
REACTORS
RESEARCH AND TEST REACTORS
SAFETY
TANK TYPE REACTORS
TEMPERATURE GRADIENTS
TEST REACTORS
THERMOCOUPLES
WATER COOLED REACTORS
WATER MODERATED REACTORS