Evaluation of the neutron flux distribution in an AmBe irradiator using the MCNP-4C code

Abstract

The present work shows the evaluation of the flux and absorbed dose rate of neutrons in a 241 AmBe Irradiator at IPEN facilities using the MCNP-4C transport code. The geometry of the 241 AmBe source as well as the Irradiator design, constituted of 2 neutrons sources, were modeled. In addition, four and eight sources of 241 AmBe were also considered for checking the viability to increase the Irradiator performance. The results show an increase for both flux and absorbed dose rate, mainly for the fast configuration.

Effectiveness of Serpentine Concrete as Shielding Material for Neutron Source Facility Using Monte Carlo Code

“…Te stability of the irradiator's neutron fux is a major advantage. However, the low magnitude of this fux compared to the one generated in a particle accelerator or nuclear reactor requires that large samples are irradiated for good sensitivity [4,15].…”

“…Te low neutron fux levels of this facility allow for good sensitivity when used for neutron activation analysis (NAA). Having a stable neutron fux eliminates the need for standard material in the measurement of induced activity in a sample, thus having an advantage over nuclear reactors [4].…”