OpenMC simulation of dose distributions with an 241AmBe neutron source

Abstract

The Philippine Nuclear Research Institute (PNRI) recently acquired an ²⁴¹AmBe neutron source for its recently commissioned Philippine Research Reactor-1 Subcritical Assembly for Training, Education and Research (PRR-1 SATER). Initial characterization of the ²⁴¹AmBe source was performed through Monte Carlo simulations to serve as basis for experiments to be conducted with the source. This paper applies OpenMC, an open-source Monte Carlo particle transport code, to calculate the ambient dose from the neutrons and secondary photons emitted by the neutron source. Results obtained from OpenMC were compared with those obtained from the well-validated MCNP code. Dose maps around the modeled ²⁴¹AmBe source were shown to have the same distribution pattern as simulated by both transport codes. Radiation doses at phantoms 50 cm and 100 cm away from the source were also calculated to further compare results obtained from the two codes. Neutron doses produced by both OpenMC and MCNP are in good agreement with a percentage difference of 1.02% and 0.92% for 50 cm and 100 cm, respectively, while the photon doses averages at ~5.63% percentage difference. The larger deviation obtained for the lattter is attributed to the different physics models applied for photon transport by the two codes.