Specifying the flux and dose-equivalent buildup factors for infinite slabs irradiated by radionuclide neutron sources

In this work, flux and neutron dose-equivalent buildup parameters are calculated for six radionuclide point-like neutron sources having broad energy spectrum which irradiate infinite slab-like common shielding materials of beryllium, concrete, iron, graphite, water, and lead, employing the MCNPX code. The description of the buildup factor is made in a straightforward way which is analogous to that of gamma. The parameters are obtained for thicknesses of shield from 0.5 to 10 mean free paths (mfp). The achieved dose-equivalent buildup parameters are parameterized by polynomial expressions. Using this parameterization, one can determine these factors for the desired thickness of shield material and each neutron source.