Given a core power distribution and fluid boundary conditions at the core inlet and/or outlet, a core thermal-hydraulics code predicts the three-dimensional distributions of coolant velocity, coolant pressure and coolant enthalpy within the core. The distributions of other subsidiary quantities such as coolant temperature, pin surface temperature, coolant density, mass velocity, thermodynamic quality, void fraction (for LWRs), departure from nucleate boiling ratio (for PWRs) and critical power ratio (for BWRs) may also be calculated. The predictions can be either for a single statepoint (i. e. for a single instant in time) or for multiple statepoints (i. e. for an evolution with time of the core). Steady-state codes assume equilibrium conditions for each statepoint, whereas transient codes accurately model the time variations in the fluid conditions (given appropriate time-varying core power distributions and fluid boundary conditions).

For simple analyses, the core power distributions and fluid boundary conditions are typically obtained from technical specifications. For more detailed analyses, the core power distributions and fluid boundary conditions are generally taken from the output of a whole core neutronics code and a system thermal-hydraulics code, respectively.

The code predictions are generated by solving the mass, momentum and energy conservation equations that govern the fluid flow. This is typically achieved by discretising the core into a number of notional ‘subchannels’, i. e. interconnected parallel flow channels, which extend across the full length of the core (in which case the code is often known as a ‘subchannel code’). The subchannels are further discretised in the axial direction into subchannel nodes. The mass, momentum and energy transfer between nodes can then be evaluated in a manageable way by the solution of the appropriate matrix equations (with various approximations). In the hot regions of the core, where the thermal-hydraulic conditions are of most interest, subchannel boundaries are typically delineated by the loci between the centres of adjacent pins. In the colder regions, where conditions are less important, subchannels can be larger, with boundaries encompassing multiple fuel pins and/ or assemblies.

With respect to water-cooled reactors, older core thermal-hydraulics codes tend to solve the mass, momentum and energy conservation equations for a homogeneous fluid, with empirical models to determine void fraction as a function of fluid quality (the fraction of the total mass flow rate that is vapour in a vapour-liquid two-phase flow) and to calculate liquid-vapour slip (relative motion of the two phases). In contrast, newer codes tend to solve the two-fluid equations (i. e. two sets of equations, one for each of the liquid and vapour phases), which is a more accurate, but also a more time-consuming, approach, which negates the need for the empirical models just described. However, the newer codes still require some empirical models, in particular, to take account of subcooled boiling.