COBRA IIIcMIT-2 : a digital computer program for steady state and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements

The COBRA IIIc/MIT-2 computer program computes the flow and enthalpy in rod-bundle nuclear fuel element subchannels during both steady state and transient conditions. It uses a mathematical model which considers both turbulent and diversion crossflow mixing between adjacent subchannels. Each subchannel is assumed to contain one-dimensional, two-phase, separated, slip-flow. The two-phase flow structure is assumed to be fine enough to define the void fraction as a function of enthalpy, flow-rate, heat-flux, pressure, position and time. At the present time, steady-state two-phase flow correlations are assumed to apply to transients. The mathematical model neglects sonic velocity propagation; therefore, it is limited to transients where the transient times are greater than the time for a sonic wave to pass through the channel. The equations of the mathematical model are solved by using a semi-explicit finite difference scheme. This scheme also gives a boundary-value flow solution for both steady state and transients where the boundary conditions are the inlet enthalpy, inlet mass velocity, and exit pressure.