Analysis by Monte Carlo of thermal neutron flux from a 241Am/9Be source for a system of trace analysis in materials

Neutron techniques to characterize materials have a wide range of applications, one of the major developments being the identification of terrorist threats with chemical, biological, radiological, nuclear and explosives (CBRNE) materials. In this work, a thermal neutron irradiation system, using a241Am/9Be source of 111 GBq inside polyethylene cylindrical moderators, has been designed, built and tested. The geometry of moderator and the neutron source position were fixed trying to maximize the thermal neutrons flux emitted from the system. Therefore, the system is in fact a thermalized neutron source taking advantage of the backscattered neutrons, achieving thermal fluence rates of up to 5.3x102 cm−2 s−1, with dominantly thermal spectra. Samples can be placed there for several hours and thereafter be measured to identify their component elements by NAA (Neutron Activation Analysis). Through Monte Carlo techniques employing the MCNP6 code (Pelowitz et al., 2014), four different configurations with polyethylene cylinders were simulated to choose the most adequate geometry. The theoretical model was then replicated in the neutronics hall of the Neutron Measurements Laboratory of the Energy Engineering Department of Universidad Politécnica de Madrid (LMN-UPM), carrying out experimental measurements using a BF3 neutron detector. A high agreement between MCNP6 results and the experimental values measured was observed. Consequently, the system developed could be employed in future laboratory experiments, both for the identification of trace substances by NAA and for the calibration of neutron detection equipment. •Design of an enhanced thermal neutron source for use in trace detection in substances or materials, through NAA method.•The thermal neutron irradiation system is using a241Am/9Be source.•Different configurations of polyethylene cylinders were simulated to choose the one giving the highest thermal neutron flux.•The performance was evaluated with the MCNP6 code and validated with laboratory measurements with a miniature BF3 detector.