Tally F6 was used to obtain MCNP-4C simulation data. This tally
calculates energy deposition in any cell (dosimetric volume) for
only one gamma photon that enters the cell and deposits its energy
in that cell. Output is represented in MeV/g that g denotes the mass
of material enclosed in cell. In this research regarding to geometry
approximations considered in Stankovic et al. (2010), a small cylinder
of air with 2 cm in diameter and 0.1 cm in thickness were considered
as dosimetric volume and set behind sub-cylinders in the
50 cm far from origin on Z axis. Air density (12.05104 g cm3) is
chemically specified in accordance with the recommendations of
ICRU Report 37 (Brice and David, 1984). Fig. 1 shows modeled concrete
samples, collimated beam source energy, source position and
dosimetric volume in this simulated geometry.