TY  - CONF
TI  - Analysis by Monte Carlo of Thermal Neutron Flux from a 241Am/9Be source for a system of trace analysis in materials
ID  - upm55088
T2  - XVIII INTERNATIONAL SYMPOSIUM ON SOLID STATE DOSIMETRY (ISSSD 2018)
SP  - 70
AV  - restricted
UR  - http://www.smid.org.mx/eng.htm
M2  - Oaxaca, México
PB  - Mexican Society of Irradiation and Dosimetry
Y1  - 2018///
EP  - 85
KW  - MCNP6; 241Am/9Be source; traces detection; NAA
N2  - Neutron techniques to characterize materials have a wide range of applications, although the major development has been in the identification of terrorist threats with chemical, biological, radiological, nuclear and explosive, (CBRNE) materials. The main advantage is the high penetration of neutrons in matter and therefore threats as explosives, drugs or landmines can be detected even when they are hidden under layers of earth, or elements of high density. Through Monte Carlo techniques employing the MCNP6 code, three different configurations with polyethylene cylinders were simulated to choose the most optimal geometry of this part of a detection device. The thermal neutron irradiation system is based in a 241Am/9Be source inside of polyethylene cylinders moderators with different sizes. The designed system provides an irradiation chamber allowing take advantage of the backscattering neutrons, multiplied by the source strength (6.64 s-1 measured on 1969, date UPM certificate), achieving thermal fluence rates until 530 cm-2 s -1 . Once the geometry and configuration of the system has been optimized through different simulations, the theoretical model was replicated in the neutronic hall of Energy Engineering Department of Universidad Politécnica de Madrid (EED-UPM), carried out several experimental measures using a neutron detector based in BF3. The main conclusion is a high agreement between MCNP results and experimental values measured at the UPM neutronics hall. Consequently, the optimized model could be employed in future laboratory experiments, both in the identification of substances and in the calibration of neutron dosimetry equipment. The final aim of this work has been to design the geometry of moderator and to select the neutron source position, to optimize the thermal neutrons flux in the area where the sample to analyze must be placed. This optimal configuration can be employed hereafter in detection of trace in substances or materials, through the NAA (Neutron Activation Analysis) method.
A1  - Cevallos Robalino, Lenin Estuardo
A1  - García Fernández, Gonzalo
A1  - Lorente Fillol, Alfredo
A1  - Gallego Díaz, Eduardo F.
A1  - Vega-Carrillo, Héctor René
A1  - Guzmán-García, Karen Arlete
ER  -