A 1-dimensional diffusion neutron transport solver. Solves the multi-group diffusion equation for neutron transport for a 1-D slab problem.
- user specificies an array of material cells with the required material properties.
- The input is automatically expanded into a series of slabs with the desired mesh spacing.
- The non-fission matrix(H) and an initial flux guess are generated.
- The fission matrix(F) is generated based on the flux guess
- The problem HΦ=FΦ is inverted to solve for Φ, the flux.
- This process is iterated until the flux converges.