TRAC-PF1/MOD1 correlations and models
The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate simulations of real and postulated transients in pressurized light-water reactors and for many related thermal-hydraulic facilities. The TRAC-PF1/MOD1 program is the latest released version. The code features a one- and/or three-dimensional, two-fluid treatment for the thermal hydraulics, together with other necessary modeling capabilities to describe a reactor system. This report complements the TRAC-PF1/MOD1 code manual and provides much more detailed descriptions of the various correlations and models utilized by the code. These correlations and models in general provide the necessary closure relations required by the field equations and also provide additional models necessary for the transient simulations. The report is primarily an audit, or a detailed description, of the correlations and models as represented in the code. The report also provides the basis for each through references to original literature and/or a description of the development process; lists the assumptions made in the implementation, including the definitions of required parameters not normally calculated by the code; and describes other details of the implementation. A partial assessment has been provided of some of the more important, less well-founded correlations and models to indicate the inherent accuracy, although it is believed that a true measure of overall code accuracy must include assessment of the code against integral experiment data because of synergistic effects.
- Research Organization:
- Nuclear Regulatory Commission, Washington, DC (USA). Div. of Reactor Accident Analysis; Los Alamos National Lab., NM (USA)
- DOE Contract Number:
- W-7405-ENG-36
- OSTI ID:
- 6493147
- Report Number(s):
- NUREG/CR-5069; LA-11208-MS; ON: TI89005993
- Country of Publication:
- United States
- Language:
- English
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ACCIDENTS
COMPUTER CODES
COMPUTERIZED SIMULATION
COOLING SYSTEMS
CORRELATIONS
CRITICAL FLOW
ENERGY SYSTEMS
ENERGY TRANSFER
FLOW MODELS
FLUID FLOW
FLUID MECHANICS
HEAT TRANSFER
HYDRAULICS
KINETICS
LOSS OF COOLANT
MATHEMATICAL MODELS
MECHANICS
PUMPS
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR COOLING SYSTEMS
REACTOR KINETICS
REACTOR SAFETY
REACTORS
SAFETY
SIMULATION
SOLUBILITY
T CODES
TRANSIENTS
TWO-PHASE FLOW
WATER COOLED REACTORS
WATER MODERATED REACTORS