mc-solver is a Monte Carlo solver for neutron transport in slab geometry. It supports groupwise cross sections and scores scalar fluxes by energy group.
From the project directory, do /scripts/run.sh /path/to/output/file to produce /path/to/output/file.mat, a Matlab file with each energy group's scalar flux. For now, the problem can be set up by editing src/mc-solver.cpp which I think is straightforward enough to edit for most material and geometry configurations.
From the project directory run make to build the executable mc-solver.
A fixed source, purely absorbing problem has been done using mc-solver with 100000000 histories. The test slab is 100 cm thick with vacuum conditions at the left boundary (x = 0 cm) and reflecting conditions at the right boundary (x = 100 cm). The macroscopic absoroption cross section is 0.1 cm-1. The mean free path is 10 cm and the scattering ratio is 0. A comparison to the analytic solution is shown below.
A fixed source, non-multiplying medium has been done using mc-solver with 100000000 histories. The test slab is 100 cm thick with vacuum conditions at the left boundary (x = 0 cm) and reflecting conditions at the right boundary (x = 100 cm). The mean free path was fixed at 10 cm. The scattering ratio Σs/Σt was varied between 0, 0.5, 0.9 and compared with a method of characteristics (MOC) and discrete ordinates (sn-solver-2) solver below.
A fixed source, two-group, multiplying medium has been modeled using mc-solver. The test slab is 87.5 cm thick with vacuum conditions at the left boundary (x = 0 cm) and reflecting conditions at the right boundary (x = 87.5 cm). The mean free path of fast and thermal neutrons were both 10 cm. Material properties of the slab are given below.
A comparison of mc-solver with sn-solver-2 is given below. The solution for mc-solver used 1000000 histories.
