COBRA/TRAC simulation of a semiscale small break LOCA. [PWR]
Conference
·
OSTI ID:5624402
The COBRA/TRAC computer program has been developed to predict the thermal-hydraulic response of nuclear reactor primary coolant systems to small and large break loss-of-coolant accidents and other anticipated transients. The code solves the compressible three-dimensional, two-fluid, three-field equations for two-phase flow in the reactor vessel. The three fields are the vapor field, the continuous liquid field, and the liquid drop field. A five-equation drift flux model is used to model fluid flow in the primary system piping, pressurizer, pumps, and accumulator. In the code modeling of Semiscale Test S-UT-2, the intact and broken loops, and the upper head injection (UHI) systems are represented by one-dimensional components,. The pressure vessel and two steam generators are modeled using the three-dimensional VESSEL component. The results from the COBRA/TRAC calculation give a reasonable match with the measured data.
- Research Organization:
- Pacific Northwest Lab., Richland, WA (USA)
- DOE Contract Number:
- AC06-76RL01830
- OSTI ID:
- 5624402
- Report Number(s):
- PNL-SA-11420; CONF-831047-96; ON: DE84002589
- 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
ACCIDENTS
COMPUTER CALCULATIONS
DATA
ENERGY TRANSFER
EXPERIMENTAL DATA
FLOW RATE
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
INFORMATION
LOSS OF COOLANT
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NUMERICAL DATA
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
TEST FACILITIES
THEORETICAL DATA
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900* -- Nuclear Reactor Technology-- Reactor Safety
ACCIDENTS
COMPUTER CALCULATIONS
DATA
ENERGY TRANSFER
EXPERIMENTAL DATA
FLOW RATE
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
INFORMATION
LOSS OF COOLANT
MECHANICS
NUMERICAL DATA
PRESSURE GRADIENTS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR SAFETY
REACTORS
SAFETY
TEST FACILITIES
THEORETICAL DATA
WATER COOLED REACTORS
WATER MODERATED REACTORS