Transport Calculations of Radiation in Different Materials

Abstract

Transport calculations were made for radiation doses and fluxes in different materials for different ranges. The Monte Carlo multigroup experimental transport code MORSE-CG [1] was used for these calculations. Some modifications were introduced to this code. These modifications include the addition of subroutine called SOURCE, and the addition of a function called DIREC. The subroutine SOURCE allows the radiation source of particles to have certain geometry instead of the built-in point source. While the function DIREC makes the radiation to transport in a certain direction to reduce the time of calculations.The obtained results of calculations were compared with experimental referenced data which show a good agreement.