COBRA-NC: a thermal hydraulics code for transient analysis of nuclear reactor components. Volume 2. COBRA-NC numerical solution methods

Technical Report · Tue Apr 01 00:00:00 EST 1986

OSTI ID:5396848

Thurgood, M J; George, T L; Wheeler, C L

The COBRA-NC computer program has been developed to predict the thermal-hydraulic response of nuclear reactor components to thermal-hydraulic transients. The code solves the multicomponent, compressible three-dimensional, two-fluid, three-field equations for two-phase flow. The three fields are the vapor 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. This volume describes the finite-volume equations and the numerical solution methods used to solve these equations. It is directed toward the user who is interested in gaining a more complete understanding of the numerical methods used to obtain a solution to the hydrodynamic equations.

• Research Organization: Pacific Northwest Lab., Richland, WA (USA)
• DOE Contract Number: AC06-76RL01830
• OSTI ID: 5396848
• Report Number(s): NUREG/CR-3262-Vol.2; PNL-5515-Vol.2; ON: TI86010442
• Country of Publication: United States
• Language: English

Related Subjects

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