RESUMO

In this work, were determined some radiation shielding and dosimetric parameters of three types of bricks for photons energy from 1 keV to 100 GeV photons using the Phy-X/PSD software, and for comparison also has been calculated the same parameters for NBS concrete. The parameters calculated are the linear attenuation coefficients (LAC), effective atomic numbers (Zeff), half value layers (HVL), the energy absorption buildup factors (EABF) and the exposure buildup factors (EBF), as well as the coefficients to use the geometric progression (G-P) fitting method. Obtained results show that the three types of bricks can be used safely for the design of medical facilities housing mammography units (less than 30 keV).

Assuntos

RESUMO

The shielding features of amethyst, chalcedony, crystal rock, milky, pink, flint, and jasper quartz were calculated. The seven varieties of quartz were collected within the territory of the state of Zacatecas in Mexico. Shielding characteristics include the linear attenuation coefficients, the effective atomic numbers and the mass interaction coefficients for 1 keV to 100 GeV photons. The interaction coefficients were calculated for the coherent scattering, the photoelectric absorption, the Compton scattering, and the pair production occurring in the electric field of the atomic nucleus and the electrons. The linear attenuation coefficients were compared with those of the Portland concrete widely used as radiation shielding. In the low energy region where the photoelectric absorption is dominant the mass interaction coefficients and the total linear attenuation coefficients have three resonances due to the Si, Ca and Fe concentration, these resonances are also noticed in the effective atomic number. For photons below 60 keV the shielding performance of quartz can be used to shield X-ray devices working with less than 60 kV.

RESUMO

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.

RESUMO

A thermal neutron system intended to be used in neutron activation analysis has been designed by Monte Carlo methods. The device is based on a241Am/9Be neutron source of 111 GBq, placed inside a cylindrical cavity open inside a parallelepiped of moderator material. Three different moderator materials, water, graphite and high-density polyethylene (HDPE), were simulated to check what is the most suitable for the detection system, concluding that HDPE reach the better performance. The device achieves an increased thermal neutron flux by taking advantage of neutron moderation in the polyethylene and the neutron scattering in the irradiation chamber walls. The thermal fluence rates obtained were 904 cm-2  s-1, i.e. 8.144 cm-2 s-1 GBq-1, with a fraction of thermal neutrons at the best point of 83% of pristine fast neutrons emitted by the source. The device has been designed by Monte Carlo techniques using the MCNP6 code, and the main tasks developed were to select the moderator material and to maximize the thermal neutrons flux in the irradiation chamber.

RESUMO

Optical, chemical, and radiation shielding characteristics of 6 types of Mexican onyx from the semi-desert of Zacatecas have been determined. Onyx samples are Yellow, Blue, Brown, Orange, Red and Green. Characterization was carried out using Energy dispersed X-Ray fluorescence, Visible Ultraviolet Spectrometry and Infrared Spectroscopy. With these techniques it was possible to study the chemical, physical and optical characteristics of the Mexican Onyx; from where the elemental composition was obtained, the absorption peaks, the vibrational modes and the optical band gap. As well as the interactions of these materials with radiation was studied. From this last study we obtained the Mass interaction coefficient for the coherent and incoherent dispersion, the photoelectric effect, the Compton scattering and the pair production, in the atomic nucleus and the electron field. For all types of onyx the total mass attenuation coefficients are alike showing a resonance due to K-shell of Ca. Mexican onyx has a good shielding features to photons of energy less than 100 keV because the most probable event is the photoelectric effect.

RESUMO

MgB4O7 doped with rare earths and alkaline elements has been reported as a good TLD because of its high sensitivity, effective atomic number close to that of biological tissue and low fading. In this work, thermoluminescent matrices were synthesized of MgB4O7:Dy and MgB4O7:Dy, Na to evaluate their thermoluminescent response (TL) when exposed to γ-rays and neutrons. The amount of Dy was studied in a concentration range of 0.01-1.5 mol% of total doping, while for Na the concentration of 0.5 mol% was established to determine the TL response as a function of doping. The synthesis of the powders was carried out by the method of wet reaction assisted by heat treatment and the samples were characterized by techniques of scanning electron microscopy and X-ray diffraction to determine the size of grain and crystallographic phase. For the dosimetric study, thermoluminescent phosphors were irradiated with a source of 137Cs at an estimated dose 6.8 ±â€¯0.4 mGy to evaluate their response to γ-rays exposure, while for neutrons a source of 241AmBe was used (estimated dose of 3.1 ±â€¯0.1 mGy). The thermoluminescent responses are similar for all materials exposed to γ-rays as for neutrons, the differences are shown to 280 °C, where a peak of high temperature is observed in materials exposed to neutrons.

RESUMO

X-ray images used for radio-diagnosis are very useful to evaluate the progress of a treatment or to have a better diagnosis. However, during the interaction between the incoming X-ray beam and the body surface, part of the radiation is scattered out reaching other parts of the body delivering an undesirable dose. In this work the dose in eye lenses, thyroid, and gonads of a solid water phantom was measured using thermoluminescent dosimeters, while a Computer Tomography of the torso was obtained. With the measured absorbed dose the effective dose was calculated. Thus, the effective dose in the eye lens, thyroid, and gonads is approximately 57, 214 and 9 µSv respectively. The largest effective dose was on that area located nearest to the region where the radiation is scattered.

Assuntos

RESUMO

In the aim to support treatments and diagnostics in Dentistry the most common radiographies are the periapical radiography and the dental panoramic. In the dental panoramic the X-ray beam is addressed into a large zone of the face obtaining the image of the upper and lower jaw. In this procedure part of the X-rays are scattered reaching some other parts of the patient body. In this work the absorbed dose in the eye lens, thyroid, salivary glands, mammary glands, and gonads, was measured in patients undergoing a dental orthopantomography. Measurements were carried out with thermolmuminiscent dosimeters. With the absorbed doses the Effective dose were calculated. The highest dose was obtained in the salivary glands (30.4 µSv) and the lowest dose was obtained in thyroid (3.8 µSv).

Assuntos

Assuntos

RESUMO

Mediante métodos Monte Carlo se calcularon las dosis en los órganos internos de una mujer, con tres meses de embarazo, debidas al radioiodo captado por su tiroides, así como a 1 metro de la glándula. Se utilizó un modelo tridimensional del cuerpo de la mujer y mediante Monte Carlo, los fotones de radioiodo se transportaron isotrópicamente desde la tiroides hacia todo el cuerpo y se calculó la dosis absorbida por sus órganos internos. También se determinó el Kerma en aire (K) y la dosis equivalente ambiental (H*(10)) a 1 m de la glándula. Se determinaron dos factores de actividad a dosis, también llamados Factores Gamma, que permiten estimar la dosis que la paciente produce a personas a su alrededor. De la radiación gamma que emite el 131I en la tiroides se encontró que el timo recibe la mayor dosis mientras que el útero es el órgano que menor dosis recibe. Los factores gamma determinados fueron: ΓKAire = 52.06 μGy-m2-h-1-GBq-1, y ΓH*(10) = 67.72 μSv-m2-h-1-GBq-1. La distribución de la dosis absorbida por los órganos internos se atribuye a la distancia relativa entre la tiroides y los otros órganos, al blindaje inter-órganos, su tamaño y a su composición elemental. Los factores ΓKAire y ΓH*(10) permiten estimar la exposición que la paciente produce sobre el personal a su alrededor. Con esto, el médico nuclear, el físico médico o el responsable de la seguridad radiológica en el hospital pueden dar indicaciones mas precisas sobre el comportamiento de las personas en torno al paciente.

Assuntos