Calculation of the CANON experiment using the TRAC code. [PWR; BWR]
Conference
·
OSTI ID:5424400
In order to focus on the ability of the TRAC code to predict critical discharge flowrates and water subcooling in a loss-of-coolant simulation, the TRAC-P1A version of the code has been applied to calculate the CANON experiments (Riegel and Marechal, 1977). The CANON experimental assembly consists of a horizontal 4.4 meter long pipe (10 cm I.D.) filled with pressurized (32 bar) subcooled water. Both ends of the pipe are closed off. The temperature of the water varied from 200 to 230/sup 0/C. The experiment consisted in suddenly opening one end of the pipe and recording the pressure as a function of time at several locations along the pipe and measuring the void fraction as a function of time at a location of 1.5 meters from the dead end. The void fraction measurements were done using a neutron scattering technique. A diaphragm is located at the broken end of the pipe with a variable diameter (30, 50, 70, 100mm) to slow down the ejection of water to about 5 seconds from 1 second (depending on the opening) in duration.
- Research Organization:
- Brookhaven National Lab., Upton, NY (USA)
- DOE Contract Number:
- EY-76-C-02-0016
- OSTI ID:
- 5424400
- Report Number(s):
- BNL-NUREG-27191; CONF-800607-22
- Country of Publication:
- United States
- Language:
- English
Similar Records
Independent assessment of TRAC-P1A with SUPER-CANON blowdown tests. [PWR]
Water Reactor Safety Research Division: Quarterly Progress Report, July 1-September 30, 1980
Independent assessment of TRAC code with various blowdown experiments
Technical Report
·
Mon Dec 31 23:00:00 EST 1979
·
OSTI ID:6020259
Water Reactor Safety Research Division: Quarterly Progress Report, July 1-September 30, 1980
Technical Report
·
Fri Oct 31 23:00:00 EST 1980
·
OSTI ID:5450095
Independent assessment of TRAC code with various blowdown experiments
Conference
·
Wed Dec 31 23:00:00 EST 1980
·
OSTI ID:5998546
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BWR TYPE REACTORS
COMPARATIVE EVALUATIONS
COMPUTER CALCULATIONS
FLOW RATE
FLUID MECHANICS
HYDRAULICS
LOSS OF COOLANT
MECHANICS
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
SIMULATION
WATER COOLED REACTORS
WATER MODERATED REACTORS
210100 -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
BWR TYPE REACTORS
COMPARATIVE EVALUATIONS
COMPUTER CALCULATIONS
FLOW RATE
FLUID MECHANICS
HYDRAULICS
LOSS OF COOLANT
MECHANICS
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
SIMULATION
WATER COOLED REACTORS
WATER MODERATED REACTORS