TRAC-PF1/MOD1 computer code and developmental assessment
The Transient Reactor Analysis Code (TRAC) is being developed at Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized-water reactors and for many thermal-hydraulic experimental facilities. The code features either one- or three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid, nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences from normal operating conditions to severe transients. A new numerical algorithm is used in the one-dimensional hydrodynamics which permits this portion of the fluid dynamics to violate the material Courant condition. The technique permits large time steps and hence reduced running time for slow transients. Included in the article are the results of development assessment calculations performed with TRAC-PF1/MOD1 prior to its public release. The assessment set consists of six integral effects calculations in the Loss-of-Fluid Test and Semiscale facilities. Computer run times required to predict each test also are reported.
- Research Organization:
- Los Alamos National Lab., NM
- OSTI ID:
- 7053599
- Journal Information:
- Nucl. Saf.; (United States), Journal Name: Nucl. Saf.; (United States) Vol. 26:4; ISSN NUSAA
- Country of Publication:
- United States
- Language:
- English
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ACCIDENTS
ALGORITHMS
BORON
COMPUTER CODES
COMPUTERIZED SIMULATION
CONTAINERS
CONVECTION
COOLING SYSTEMS
ECCS
ELEMENTS
ENERGY SYSTEMS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HIGH PRESSURE COOLANT INJECTION
HYDRAULICS
KINETICS
LASL
LOSS OF COOLANT
MASS TRANSFER
MATHEMATICAL LOGIC
MECHANICS
NATIONAL ORGANIZATIONS
NATURAL CONVECTION
ONE-DIMENSIONAL CALCULATIONS
PRESSURE VESSELS
PRIMARY COOLANT CIRCUITS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR KINETICS
REACTOR PROTECTION SYSTEMS
REACTOR SAFETY
REACTOR SAFETY EXPERIMENTS
REACTORS
SAFETY
SECONDARY COOLANT CIRCUITS
SEMIMETALS
SIMULATION
T CODES
THREE-DIMENSIONAL CALCULATIONS
TRANSIENTS
TWO-PHASE FLOW
US AEC
US DOE
US ERDA
US NRC
US ORGANIZATIONS
WATER COOLED REACTORS
WATER MODERATED REACTORS