Monte Carlo study of detection systems of explosives and illicit substances using a D-D neutron generator

Prompt gamma neutron activation analysis (PGNAA) using a Deuterium-Deuterium (D-D) neutron generator, is one of the most promising method to detect explosives and illicit substances. Using Monte Carlo methods with MCNP6 code, a model of a DD Generator (DD-110 Adelphi Technology [1]) has been built with one or two NaI(Tl) detectors of 1.5" x 1" size in order to determine an optimal configuration. Several calculations have been performed to analyse the detection power for samples of different explosives: RDX, Ammonium Nitrate, land mines, military and homemade explosives, have been selected. Different models have been made varying moderator and shielding thicknesses (polyethylene and lead) and we conclude that the configuration with 10 cm thickness of polyethylene, is the best configuration to obtain a neutron thermal flux appropriate for detection of these simulated samples of explosive

Passive neutron area monitor with pairs of TLDs as neutron detector

A passive neutron area monitor has been designed using Monte Carlo methods; the monitor is a polyethylene cylinder with pairs of thermoluminescent dosimeters (TLD600 and TLD700) as thermal neutron detector. The monitor was calibrated with a bare and a thermalzed 241AmBe neutron sources and its performance was evaluated measuring the ambient dose equivalent due to photoneutrons produced by a 15 MV linear accelerator for radiotherapy and the neutrons in the output of a TRIGA Mark III radial beam port

Neutron area monitor with passive detector

Using Monte Carlo methods the responses of a passive neutron monitor area has been calculated. To detect thermal neutrons the monitor has a gold foil that is located at the center of a polyethylene cylinder. Impinging neutrons are moderated by polyethylene nuclei reaching the gold foil with the energy to induce activation through the reaction 197Au (n,γ) 198Au. The198Au decays emitting 0.411 MeV gamma rays with a half life of 2.7 days. The induced activity is intended to be measured with a gamma-ray spectrometer with a 3”Ø×3” NaI(Tl) scintillator and the saturation activity is correlated to the ambient dose equivalent. The response was calculated for 47 monoenergetic neutron sources ranging from 1×10−9 to 20 MeV. Calculated fluence response was compared with the response of neutron monitor area LB 6411. = Se utilizó el método de Monte Carlo para calcular las respuestas de un monitor de neutrones pasivo. Para detectar los neutrones térmicos el monitor tiene una lámina de oro que se encuentra en el centro de un cilindro de polietileno. Los neutrones que inciden son moderados por los núcleos de polietileno que llega a la lámina de oro con la energía para inducir la activación a través de la 197Au reacción (n,γ)198 Au. El 197Au decae emitiendo rayos gamma de 0.411 MeV con una vida media de 2.7 días. La actividad inducida se destina a medir con un espectrómetro de rayos gamma con un detector de centelleo de 3”Ø×3” NaI (Tl) y la actividad de saturación se correlaciona con la dosis equivalente ambiental. La respuesta se calculó para 47 fuentes de neutrones monoenergéticos desde 1×10−9 a 20 MeV. La respuesta a la fluencia se comparó con la respuesta del monitor de área LB 6411

Neutron detectors of 10B+ZnS(Ag) as an alternative to 3He-based installed in the Radiation Portal Monitors in homeland Security

Using Monte Carlo method features of 10B+ZnS(Ag) neutron detector under real operational conditions has been calculated, this results were compared against measurements. This detector is a viable substitute for 3He detectors installed in custom borders to prevent the illicit traffic of special nuclear materials, SNM

Passive neutron area monitor with TLD pairs

The response of a passive neutron area monitor with pairs of thermoluminescent dosimeters has been calculated using the Monte Carlo code MCNP5. The response was calculated for one TLD 600 located at the center of a polyethylene moderator. The response was calculated for 47 monoenergetic neutron sources ranging from 1E(-9) to 20 MeV. Response was calculated using two irradiation geometries, one with an upper source and another with a lateral source. For both irradiation schemes the response was calculated with the TLD in two positions, one parallel to the source and another perpendicular to the source. The advantage of this passive neutron monitor area is that can be used in locations with intense, pulsed and mixed radiation fields like those in radiotherapy vault rooms with linear accelerators. La respuesta de un monitor de área pasivo para neutrones con pares de dosímetros termoluminiscentes TLDs ha sido calculada mediante métodos Montecarlo con el código MCNP5. La respuesta fue calculada para un TLD 600 localizado en el centro del moderador cilíndrico de polietileno. La respuesta se calculó para 47 fuentes mono energéticas de neutrones con energías de 1E (-9) a 20 MeV. La respuesta se calculó para dos geometrías de irradiación, una con una fuente superior y la otra con una fuente lateral, para ambas geometrías la respuesta se obtuvo con el TLD en dos posiciones respecto a la fuente, una perpendicular y la otra paralela. La ventaja del monitor pasivo es que puede ser usado en instalaciones con campos de radiación intensos, pulsados y mixtos como los que se producen en los bunkers de radioterapia con aceleradores lineales

Photon Shielding Features of Quarry Tuff

Cantera is a quarry tuff widely used in the building industry; in this work the shielding features of cantera were determined. The shielding characteristics were calculated using XCOM and MCNP5 codes for 0.03, 0.07, 0.1, 0.3, 0.662, 1, 2, and 3 MeV photons. With XCOM the mass interaction coefficients, and the total mass attenuation coefficients, were calculated. With the MCNP5 code a transmission experiment was modelled using a point-like source located 42 cm apart from a point-like detector. Between the source and the detector, cantera pieces with different thickness, ranging from 0 to 40 cm were included. The collided and uncollided photon fluence, the Kerma in air and the Ambient dose equivalent were estimated. With the uncollided fluence the linear attenuation coefficients were determined and compared with those calculated with XCOM. The linear attenuation coefficient for 0.662 MeV photons was compared with the coefficient measured with a NaI(Tl)-based -ray spectrometer and a 137Cs source

Photon Shielding Features of Quarry Tuff

Abstract. Cantera is a quarry tuff widely used in the building industry; in this work the shielding features of cantera were determined. The shielding characteristics were calculated using XCOM and MCNP5 codes for 0.03, 0.07, 0.1, 0.3, 0.662, 1, 2, and 3 MeV photons. With XCOM the mass interaction coefficients, and the total mass attenuation coefficients, were calculated. With the MCNP5 code a transmission experiment was modelled using a point-like source located 42 cm apart from a point-like detector. Between the source and the detector, cantera pieces with different thickness, ranging from 0 to 40 cm were included. The collided and uncollided photon fluence, the Kerma in air and the Ambient dose equivalent were estimated. With the uncollided fluence the linear attenuation coefficients were determined and compared with those calculated with XCOM. The linear attenuation coefficient for 0.662 MeV photons was compared with the coefficient measured with a NaI(Tl)-based �-ray spectrometer and a 137Cs source

Análisis Monte Carlo del Laboratorio de Patrones Neutrónicos del CIEMAT

Mediante métodos Monte Carlo se ha caracterizado el campo de neutrones producido por fuentes de calibración en el laboratorio de Patrones Neutrónicos del Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas. El laboratorio cuenta con dos fuentes de calibración de neutrones: 241AmBe y 252Cf que son almacenadas en una piscina de agua y se colocan sobre la bancada de calibración usando sistemas controlados a distancia. Para caracterizar el campo de neutrones se construyó un modelo tridimensional de la sala donde se incluyó la bancada de acero inoxidable, la mesa de irradiación y la piscina de almacenamiento. El modelo de las fuentes incluyó el doble encapsulado de acero, como encamisado. Con el fin de determinar el efecto que produce la presencia de los diferentes componentes de la sala, durante la caracterización se estimaron los espectros de neutrones, el flujo total y la rapidez de dosis equivalente ambiental a 100 cm de la fuente. La presencia de los muros, suelo y techo de la sala es la que causa la mayor modificación en los espectros y los valores integrales del flujo y la rapidez de dosis equivalente ambiental

Neutron field characteristics of Ciemat´s Neutron Standards Laboratoy

Monte Carlo calcultarions were carried out to characterized the neutron field produced by the calibration neutron sources of the Neutron Standards Laboratory at the Research Center for Energy, Environment and Technology (CIEMAT) in Spain. For 241 AmBe and 252 Cf neutron sources, the neutron spectra, the ambien dose equivalent rates and the total neutron fluence were estimated. In the calibration hall, there are everal itmes that modify the neutron field. To evaluate their effects different cases were used, front point-like source in vacuum upto the full model. Additionally, using the full model, the neutron spectra were estimated to different distances along the bench; with these spectra, the total neutron fluence and the ambient dose equivalent rates were calculated. The hall walls induce tha largest changes in the neutron spectra and the respective integral quantities. The free-field neutron spectrum is modified due the room return effect

Laboratorio de patrones neutrónicos del CIEMAT

Mediante una serie de cálculos con métodos Monte Carlo con el código MCNPX se ha caracterizado el campo de neutrones producido por las fuentes de calibración existentes en el laboratorio de Patrones Neutrónicos, LPN, del Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas, CIEMAT. El laboratorio cuenta con dos fuentes de calibración de neutrones una de 241AmBe y una 252Cf que son almacenadas en una piscina de agua. Se construyó un modelo tridimensional detallado de la sala con la bancada de acero inoxidable puntualizando en el seleccionador de las fuentes que las coloca en su posición de irradiar a 4 m del suelo, en la mesa de irradiación y la piscina de almacenamiento. Cada una de las fuentes se definió en el modelo en su doble encapsulado de acero. Se calcularon los espectros con distintos casos con el fin de calcular la aportación de cada elemento que impacta al transporte de los neutrones. Se calcularon los espectros de las fuentes de calibración a diferentes distancias colocadas respecto a la fuente desde 0, 15, 35, 50 a 300 cm sobre la bancada y de igual manera los valores de la dosis equivalente ambiental utilizando los criterios del ICRP 74. Los resultados muestran claramente que la mayor contribución en la modificación del espectro se le atribuye a las paredes, y suelo del recinto del LPN