COBRA-NC: a thermal-hydraulic code for transient analysis of nuclear reactor components. Equations and constitutive models. Volume 1

Technical Report · Thu May 01 04:00:00 EDT 1986

OSTI ID:5839159

Wheeler, C L; Thurgood, M J; Guidotti, T E; DeBellis, D E

COBRA-NC is a digital computer program written in FORTRAN IV that simulates the response of nuclear reactor components and systems to thermal-hydraulic transients. The code solves the multicomponent, compressible, three-dimensional, two-fluid, three-field equations for two-phase flow. The three velocity fields are the vapor/gas field, the continuous liquid field, and the liquid drop field. The code has been used to model flow and heat transfer within the reactor core, the reactor vessel, the steam generators, and in the nuclear containment. The conservation equations, equations of state, and physical models that are common to all applications are presented in this volume of the code documentation.

• Research Organization: Pacific Northwest Lab., Richland, WA (USA)
• DOE Contract Number: AC06-76RL01830
• OSTI ID: 5839159
• Report Number(s): NUREG/CR-3262-Vol.1; PNL-5515-Vol.1; ON: TI86010417
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220200 -- Nuclear Reactor Technology-- Components & Accessories
220900* -- Nuclear Reactor Technology-- Reactor Safety
42 ENGINEERING
420400 -- Engineering-- Heat Transfer & Fluid Flow
ACCIDENTS
BOILERS
C CODES
COMPUTER CODES
COMPUTERIZED SIMULATION
CONTAINERS
CONTAINMENT
CONTAINMENT SYSTEMS
ENERGY TRANSFER
ENGINEERED SAFETY SYSTEMS
FLUID FLOW
FLUID MECHANICS
FORTRAN
HEAT TRANSFER
HYDRAULICS
LOSS OF COOLANT
MATHEMATICAL MODELS
MECHANICS
NUCLEAR FACILITIES
NUCLEAR POWER PLANTS
OPERATION
POWER PLANTS
PROGRAMMING LANGUAGES
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTOR CORES
REACTOR OPERATION
REACTOR VESSELS
SIMULATION
STEAM GENERATORS
THERMAL POWER PLANTS
TRANSIENTS
VAPOR GENERATORS