|Program name||Package id||Status||Status date|
|Package ID||Orig. computer||Test computer|
|NEA-1614/001||IBM PC||PC Pentium III,Linux-based PC,DEC ALPHA W.S.,SGI W.S.,HP W.S.|
COBRA-EN is an upgraded version of the COBRA-3C/MIT for thermal-hydraulic transient analysis of reactor cores. Starting from a steady-state condition in a LWR core or fuel element, the code allows to simulate the thermal-hydraulic transient response to user-supplied changes of the total power, of the outlet pressure and of the inlet enthalpy and mass flowrate.
The thermal-hydraulic homogeneous model of COBRA-EN is based on three partial differential equations that, using what is known as "subchannel approximation", describe the conservation of mass, energy and momentum vector in axial and lateral directions for the water liquid/vapor mixture and the interaction of the two-phase coolant with the system structures. Optionally, a fourth equation can be added which tracks the vapor mass separately and which, along with the correlations for vapor generation and slip ratio, replaces the subcooled quality and quality/void fraction correlations, needed to extend the capabilities of the essentially homogeneous three-equation model.
If the computational domain is subdivided into a number of axial intervals, the control volume for mass, energy and axial momentum is a segment of subchannel while the control volume for the lateral momentum is a segment of the somewhat arbitrary region which straddles the two adjoining subchannels around a lateral gap. In each control volume, the flow equations as well as the one-dimensional (r) heat conduction equations in the fuel rods are approximated by finite differences. The resulting equations for the hydrodynamic phenomena form a system of coupled nonlinear equations that are solved either by an implicit iterative scheme based on the calculation of the pressure gradients in the axial direction or by a Newton-Raphson iteration procedure. The heat conduction equations in the solid structures are treated implicitly. Moreover, the rod-to-coolant heat transfer model is featured by a full boiling curve, comprising the basic heat-transfer regimes (forced convection, nucleate boiling, transition and film boiling), each represented by a set of optional correlations for the heat-transfer coefficient.
COBRA-EN was also improved with the VIPRE features and with some EPRI correlations.
COBRA-EN was widely used to verify the SBWR and AP600 design in the safety studies relating to the reactivity transient accidents.
Each of sample problem 1 thermal-hydraulic transient analyses with 546 assembly-sized (*20 cm) nodes in a PWR core octant and 150 time steps (each 0.01 s long ) requires about half an hour of CP time on a PC-486/100 but other typical problems concerning actual large LWR cores may require some thousands of nodes and some hundred or thousand time steps. Therefore, the CP times can range from a few to some tens of minutes for a steady-state and rise to some hours for a transient.
|Package ID||Status date||Status|
|NEA-1614/001||11-AUG-2001||Tested at NEADB|
E. Salina, G. Alloggio, E. Brega,
"QUARK: a Computer Code for the Neutronic and Thermal-Hydraulic Space- and Time-Dependent Analysis of Light Water Reactor Cores by Advanced Nodal Techniques", Synthesis Srl, rep. 1034/1 prepared for ENEL-ATN/GNUM, Milan, September 1994
E. Salina, E. Brega,
"The NORMA Program for Simulating the Long-Term Neutronic and Thermal-Hydraulic Behavior of Large LWR's by Three-Dimensional Coarse-Mesh Diffusion Methods", Synthesis Srl, rep. 1034/2 prepared for ENEL-ATN/GNUM, Milan, July 1995
E. Brega, R. Fontana, E. Salina,
"The NORMA-FP Program to Perform a Subchannel Analysis from Converged Coarse-Mesh Nodal Solutions (Rev. 3)", ENEL-DSR-CRTN-N5/91/05/MI, Milan, September 1991
|Package ID||Computer language|
Tested on a PC-486/100 provided with Microsoft DOS 6.0 and FORTRAN Power Station Compiler version 1.0.
Almost all in FORTRAN-77 except a few non-standard statements such as INCLUDE for the Common Blocks and CPU time, clock and date routines which are according to MS FORTRAN Power Station Compiler.
Keywords: fuel rods, heat transfer, light-water reactors, reactor cores, reactor kinetics, reactor safety, thermodynamics, three-dimensional, transients, two-phase flow.