The thermal-hydraulic behavior of the reactor coolant system is simulated with a fully non-equilibrium two fluid model (three conservation equations for the vapor phase and three conservation equations for the liquid phase) in one-dimensional approximation.  

The KORSAR code also provides for prediction of changes in the soluble poison (boric acid) concentration, dissolution of noncondensable gases in the liquid phase, their transport in the both phases, degassing under certain liquid phase conditions, and effect of noncondensables on the heat transfer behavior. 

The unsteady state heat transfer in heat conduction structures is predicted in one- and two-dimensional approximations.

Unsteady-state calculations of neutron processes in pressurized water reactor cores are carried out in a quasi-spatial approximation using a point reactor neutron kinetics model with an arbitrary number of delayed neutron precursor groups, or such calculations are performed using the 3D spatial distribution of neutron fluxes in the core with a two-group  diffusion approximation.

Lumped parameter models are mainly used for calculating the processes in plant equipment (centrifugal pump, accumulator, etc.).

Dedicated mathematical models were developed and implemented into the code to describe the following specific effects:

  • critical break flow,
  • countercurrent flow limitation of water and water vapor [5],
  • critical heat flux in steam generating pipelines and fuel rod bundles,
  • reflooding,
  • two-fluid flow stratification in horizontal pipes [4],
  • radiative heat transfer.