This code solves the neutron transport equation in slab geometry for two groups using the deterministic Sn method.
Two test problems (Tests A and B) that simulate interaction of uranium and MOX fuel assemblies. There are two half-assemblies next to each other with reflective boundary conditions on the outside. A MOX fuel half-assembly is on the left and a uranium half-assembly is on the right (see Figure 1). Each assembly contains only one kind of fuel pin cells, with 8 pin cells per assembly.