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COBRA/TRAC - A Thermal-Hydraulics Code for Transient Analysis of Nuclear Reactor Vessels and Primary Coolant Systems. Volume 1. Equations and Constitutive Models.

NUREGCR3046V1

Publication Date	1983
Personal Author	Thurgood, M. J.; Kelly, J. M.; Guidotti, T. E.; Kohrt, R. J.; Crowell, K. R.
Page Count	186
Abstract	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 accumulators. The heat generation rate of the core is specified by input and no reactor kinetics calculations are included in the solution. This volume describes the conservation equations and physical models used in the vessel module.
Keywords	Nuclear reactor accidents Two phase flow Hydraulics Thermal analysis Mathematical models Computerized simulation Reactor vessels Primary coolant circuits COBRA/TRAC computer program Loss of coolant Transients
Source Agency	Technical Information Center Oak Ridge Tennessee Nuclear Regulatory Commission
NTIS Subject Category	77H - Reactor Engineering & Nuclear Power Plants
Corporate Authors	Battelle Pacific Northwest Labs., Richland, WA.; Nuclear Regulatory Commission, Washington, DC.
Supplemental Notes	See also NUREG/CR-3046-V3.
Document Type	Technical Report
NTIS Issue Number	198315

COBRA/TRAC - A Thermal-Hydraulics Code for Transient Analysis of Nuclear Reactor Vessels and Primary Coolant Systems. Volume 1. Equations and Constitutive Models.

COBRA/TRAC - A Thermal-Hydraulics Code for Transient Analysis of Nuclear Reactor Vessels and Primary Coolant Systems. Volume 1. Equations and Constitutive Models.
NUREGCR3046V1

Publication Date	1983
Personal Author	Thurgood, M. J.; Kelly, J. M.; Guidotti, T. E.; Kohrt, R. J.; Crowell, K. R.
Page Count	186
Abstract	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 accumulators. The heat generation rate of the core is specified by input and no reactor kinetics calculations are included in the solution. This volume describes the conservation equations and physical models used in the vessel module.
Keywords	Nuclear reactor accidents Two phase flow Hydraulics Thermal analysis Mathematical models Computerized simulation Reactor vessels Primary coolant circuits COBRA/TRAC computer program Loss of coolant Transients
Source Agency	Technical Information Center Oak Ridge Tennessee Nuclear Regulatory Commission
NTIS Subject Category	77H - Reactor Engineering & Nuclear Power Plants
Corporate Authors	Battelle Pacific Northwest Labs., Richland, WA.; Nuclear Regulatory Commission, Washington, DC.
Supplemental Notes	See also NUREG/CR-3046-V3.
Document Type	Technical Report
NTIS Issue Number	198315