A new computational model for steady-state, single-phase, thermal-hydraulic, multichannel analysis of fluid flow through nuclear reactor fuel elements is presented. The model accounts for the conservation of mass, energy, and momentum subject to pressure-drop boundary conditions and leads to a nonlinear multipoint boundary-value problem. The turbulent interchange, radial thermal conduction, and forced flow due to the wire-wrap or grid between the channels are explicitly taken into account. The temperature distribution of the coolant, cladding, and fuel, and the size of the central void of the oxide fuel after thermal restructuring are computed in the model. Three different thermal-hydraulic channel arrangements, i.e., square, hexagonal, and triangular, can be treated by the method presented here. Multipin analysis with transverse interactions or multiassembly calculations without transverse interactions between the channels can be performed. The most important features of this new computational model are: (a) that the effect of axial flow area variation has been incorporated into the derivation of governing equations, (b) that the cross-flow approximation has been improved so that the assumption of constant transverse momentum flux in the direction under consideration is removed, and (c) that partial flow blockage occurring anywhere along the flow path can be analyzed, and the effect on the inlet mass velocity redistribution can be taken into account.