A heterogeneous coarse mesh radiation transport method for neutronic analysis of prismatic reactors

► A new hybrid transport method is described for whole core calculations in 3-D hexagonal geometry. ► A deterministic procedure uses stochastically-generated information to solve core problems without low-order or homogenization approximations. ► The method is evaluated using comparisons with full core Monte Carlo reference solutions of gas-cooled, graphite-moderated reactor core designs. ► This new method enables highly accurate determination of core eigenvalues and flux shapes in hexagonal cores. ► The efficiency far exceeds that of other transport methods.