Calculation of Dose Distribution in Neutron Brachytherapy Using 252-Cf Source Through the Monte Carlo Simulation and Comparison with Experimental Data

Detailed recognition of dose distribution around the brachytherapy sources in order to create appropriate plans for treatment of cancer is very important. In this study, with calculation of the dosimetric parameters of clinical 252Cf source based on TG-43U1 protocol and utilizing different tallies of dose calculation in MCNPX code [F4 (Fluence Tally), F6 (Kerma Tally) and *F8 (Dose Tally)], the dose rate at different directions and distances from the source center has been determined and compared with other experimental and simulation results. By comparing the results of this study with the experimental measurements and observing the good adaptation of the results, it was observed that the dose rate of clinical 252Cf source has its largest value at the direction of longitudinal axis of the source, which is the reason for more expansion of radioactive material distribution in this direction, in comparsion with other directions and consequently higher neutron flux in this direction, based on the angular dependance of dose rate to geometric function according to TG-43 protocol. It was also found that the F4 and F6 tally results in neutron dosimetry calculations are more accurate than the *F8 tally results. The dosimetry calculations performed in this study has provided a preliminary dosimetry characterization of 252Cf neutron sources for usage in treatment plans in the country.