V. M. Hernández-Dávila

Niveles de plomo en la población de alto riesgo y su entorno en San Ignacio, Fresnillo, Zacatecas, México

Objective. To determine the lead concentration in the blood of children and nursing or pregnant women from San Ignacio, Fresnillo, in Zacatecas, Mexico as well as in soil, plants, ash and lead-glazed pottery, in order to determine exposure due to a metal-recycling facility. Material and Methods. The study was carried out from December 2004 to April 2005. Lead in blood was measured with anodic stripping voltammetry, while dispersive energy X-ray fluorescence was used in the other matrices. Results. Based upon the criteria outlined in the Official Mexican Standards, 90% of the children was identified as category I, 5% as category II and another 5% as category III. The soil in the land near the facility contained from 73 to 84 238 ∝g/g, with an average of 4940 ∝g/g. Larger lead concentrations were found on sites located closer to the facility. San Ignacio’s soil contained, on average, 109 ∝g/g. High lead levels were found in glazed pottery and the concentration in agricultural crops was greater than 300 ∝g/g. Conclusions. Although the majority of children in San Ignacio have blood lead concentrations considered to be acceptable according to the Official Mexican Standards, several studies indicate that deleterious effects on children’s health exist even at low concentrations. The land around the metal recycling facility is contaminated with lead, and to that extent, the crops that are produced there, once ingested, are a source of contamination, which is compounded by the use of glazed pottery. Manzanares-Acuna E, Vega-Carrillo HR, Salas-Luevano MA, Hernandez-Davila VM, Letechipia-de Leon C, Banuelos-Valenzuela R. Niveles de plomo en la poblacion de alto riesgo y su entorno en San Ignacio, Fresnillo, Zacatecas, Mexico. Salud Publica Mex 2006;48:212-219.

Neutron spectra at two beam ports of a TRIGA Mark III reactor loaded with HEU fuel.

The neutron spectra have been measured in two beam ports, one radial and another tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research in Mexico. Measurements were carried out with the reactor core loaded with high enriched uranium fuel. Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a (6)LiI(Eu) scintillator and 2, 3, 5, 8, 10 and 12 in.-diameter high-density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code. For each spectrum total flux, mean energy and ambient dose equivalent were determined. Measured spectra show fission, epithermal and thermal neutrons, being harder in the radial beam port.

Determination of neutron fluence-to-dose conversion coefficients by means of artificial neural networks

An Artificial Neural Network has been designed to determinate the effective dose, the ambient dose equivalent and the personal dose equivalent fluence-to-dose conversion factors using seven count rates obtained with a Bonner Sphere Spectrometer. The data of 211 neutron spectra and their respective fluence-to-dose conversion coefficients were used to train and to test ANN. The ANN was trained using the trainsec algorithm, the definitive ANN was 7:8:9:10. From the 30 set of data used to test the ANN performance the largest difference was 11% that is close to the difference obtained in neutron dosimetry.

Artificial Neural Networks in Spectrometry and Neutron Dosimetry

The ANN technology has been applied to unfold the neutron spectra of three neutron sources and to estimate their dosimetric features. To compare these results, neutron spectra were also unfolded with the BUNKIUT code. Both unfolding procedures were carried out using the count rates of a Bonner sphere spectrometer. The spectra unfolded with ANN result similar to those unfolded with the BUNKIUT code. The H*(10) values obtained with ANN agrees well with H*(10) values calculated with the BUNKIUT code.

Monte Carlo-based method to determine the strength of a neutron source

The utilization of a gamma-ray spectrometer with a 7.62 Ø×7.62 cm NaI(Tl) detector, with a spherical moderator, has been studied with the aim to measure the neutron fluence rate and to determine the neutron source strength. Moderators with a large amount of hydrogen are able to slowdown and thermalize neutrons; once thermalized, there is a probability for thermal neutrons to be captured by hydrogen, producing 2.22 MeV gamma rays. The pulse-height spectrum collected in a multichannel analyzer shows a photopeak around 2.22 MeV whose net area is proportional to total neutron fluence rate and to the neutron source strength. The characteristics of this system were determined by a Monte Carlo study using the MCNP 4C code, where a detailed model of the NaI(Tl) was utilized. Spheres of diameters 3, 5, and 10 inch were used as moderators, and the response was calculated for monoenergetic and isotopic neutrons sources.

FPGA embedded multichannel analyzer

An multichannel analyzer has been designed, and its performance has been evaluated. The multichannel analyzer is embedded into a Field programmable gate array. The design incudes the virtual instrument in order to hand and to visualize the pulse height spectrum. Two commercially available multichannel analyzers using a NaI(Tl) and HPGe detectors were used to obtain the pulse height spectra of 137Cs, 60Co and 152Eu sources and were compared with the pulse height spectra obtained with the embedded multichannel analyzer, being alike the spectra obtained with the commercial multichannel analyzer. Our design is smaller, low cost and it has options to add other features.

Absorbed Dose in the Uterus of a Three Months Pregnant Woman Due to 131I

The use of 131I is widely used in diagnostic and treatment of patients. If the patient is pregnant the 131I presence in the thyroid it becomes a source of constant exposition to other organs and the fetus. In this study the absorbed dose in the uterus of a 3 months pregnant woman with 131I in her thyroid gland has been calculated. The dose was determined using Monte Carlo methods in which a detailed model of the woman has been developed. The dose was also calculated using a simple procedure that was refined including the photons’ attenuation in the woman organs and body. To verify these results an experiment was carried out using a neck phantom with 131I. Comparing the results it was found that the simple calculation tend to overestimate the absorbed dose, by doing the corrections due to body and organs photon attenuation the dose is 0.14 times the Monte Carlo estimation.