TRAC-PD2 independent assessment
Technical Report
·
OSTI ID:6069337
- comp.
The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in light-water reactors. The TRAC-PD2 program provides this analysis capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features a three-dimensional treatment of the pressure vessel and its associated internals, two-phase nonequilibrium hydrodynamic models, flow-regime-dependent constitutive relations, optional reflood=tracking capability for both bottom-reflood and falling-film quench fronts, and consistent treatment of entire accident sequences, including the generation of consistent steady-state conditions. The Los Alamos report, TRAC-PD2: An Advanced Best-Estimate Computer Program for Pressurized Water Reactor Loss-of-Coolant Accident Analysis, LA-8709-MS (NUREG/CR-2054), provides a detailed description of the code. This report documents the Los Alamos results of the second assessment phase, independent assessment, for TRAC-PD2.
- Research Organization:
- Los Alamos National Lab., NM (USA)
- DOE Contract Number:
- W-7405-ENG-36
- OSTI ID:
- 6069337
- Report Number(s):
- NUREG/CR-3866; LA-10166-MS; ON: TI85007373
- Country of Publication:
- United States
- Language:
- English
Similar Records
TRAC-PD2: advanced best-estimate computer program for pressurized water reactor loss-of-coolant accident analysis
TRAC-PD2 developmental assessment
TRAC-PD2 modeling of LOFT and PWR small cold-leg breaks
Technical Report
·
Tue Mar 31 23:00:00 EST 1981
·
OSTI ID:5391880
TRAC-PD2 developmental assessment
Technical Report
·
Mon Dec 31 23:00:00 EST 1984
·
OSTI ID:5817864
TRAC-PD2 modeling of LOFT and PWR small cold-leg breaks
Conference
·
Wed Dec 31 23:00:00 EST 1980
·
OSTI ID:6463579
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 CODES
CONTAINERS
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
FLOW MODELS
FLOW RATE
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MATHEMATICAL MODELS
MECHANICS
PRESSURE GRADIENTS
PRESSURE VESSELS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR SAFETY
REACTORS
SAFETY
SYSTEM FAILURE ANALYSIS
SYSTEMS ANALYSIS
TEMPERATURE GRADIENTS
TESTING
THERMAL ANALYSIS
TWO-PHASE FLOW
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 CODES
CONTAINERS
COOLING SYSTEMS
ENERGY SYSTEMS
ENERGY TRANSFER
FLOW MODELS
FLOW RATE
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MATHEMATICAL MODELS
MECHANICS
PRESSURE GRADIENTS
PRESSURE VESSELS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR SAFETY
REACTORS
SAFETY
SYSTEM FAILURE ANALYSIS
SYSTEMS ANALYSIS
TEMPERATURE GRADIENTS
TESTING
THERMAL ANALYSIS
TWO-PHASE FLOW
WATER COOLED REACTORS
WATER MODERATED REACTORS