Monte Carlo validation of dose mapping for the Tunisian Gamma Irradiation Facility using the MCNP6 code. (August 2020)

• Record Type: Journal Article
• Authors: Dridi, W.; Daoudi, M.; Farah, K.; Hosni, F.
• Is Part Of: Radiation physics and chemistry. Volume 173(2020:Aug.)
• Publication Date: 2020-08
• Subjects: Radiation chemistry -- Periodicals; Radiometry -- Periodicals; Radiation -- Periodicals; Chimie sous rayonnement -- Périodiques; 539.2
• DOI: 10.1016/j.radphyschem.2020.108942
• Languages: English
• ISSNs: 0969-806X

Abstract

Abstract: In this paper, we develop a three-dimensional configuration of the Tunisian Gamma Irradiation Facility (TGIF) to determine the spatial distribution of gamma absorbed dose rates in the irradiation area at the 60 Co source rack. The results are obtained by using the MCNP6 code. Experimental measurements of the gamma absorbed dose rate are performed in order to be compared with the calculated results. Two dosimetry systems are employed to perform measurements: red and amber perpex and cellulose triacetate (CTA). The results of MCNP6 simulations are compared to the absorbed dose measurements for the following experiments: (1) estimation of the 60 Co source rack position, (2) isotropy verification of gamma-ray emission, (3) isodose curves of the absorbed dose rate inside the irradiation room, (4) absorbed dose rate distribution in air on four vertical planes, (5) absorbed dose rate distribution in air on four horizontal planes versus increasing distance from the source axis. The root mean square error (RMSE) is employed to check the discrepancy between MCNP6 simulation and dose measurements. The values of RMSE for all experiments vary from 0.5 to 2.6%, showing good agreements between the measured and calculated results. This agreement reveals that the developed model of the TGIF is accurate. It inspires the possibility to use the model on an irradiation processing and a reload of the facility. Highlights: 3D model of the Tunisian Gamma Irradiation Facility was developed with the MCNP6 code. Red and amber perpex and cellulose triacetate dosimeters were used for measurements. The root mean square error (RMSE) was calculated for comparison of results. Good agreement between Monte Carlo simulations and experimental data was observed.