“…MCNP is a radiation transport code that can be used for neutron, photon and electron transport calculations in 3-D configurations (X-5 Monte Carlo Team, 2008). Since complex geometries can be modeled, the program is widely used in the petroleum industry and well logging problems (Wang et al, 2017;van der Hoeven et al, 2017;Ortega et al, 2014;Mendoza et al, 2007;Wielopolski et al, 2005;Forster et al, 1990;Preeg and Scott, 1986).…”