Comparison of the effective dose rate to aircrew members using hybrid computational phantoms in standing and sitting postures.

Aircraft crew members are occupationally exposed to considerable levels of cosmic radiation at flight altitudes. Since aircrew (pilots and passengers) are in the sitting posture for most of the time during flight, and up to now there has been no data on the effective dose rate calculated for aircrew dosimetry in flight altitude using a sitting phantom, we therefore calculated the effective dose rate using a phantom in the sitting and standing postures in order to compare the influence of the posture on the radiation protection of aircrew members. We found that although the better description of the posture in which the aircrews are exposed, the results of the effective dose rate calculated with the phantom in the sitting posture were very similar to the results of the phantom in the standing posture. In fact we observed only a 1% difference. These findings indicate the adequacy of the use of dose conversion coefficients for the phantom in the standing posture in aircrew dosimetry. We also validated our results comparing the effective dose rate obtained using the standing phantom with values reported in the literature. It was observed that the results presented in this study are in good agreement with other authors (the differences are below 30%) who have measured and calculated effective dose rates using different phantoms.

Use of gamma radiation and artificial neural network techniques to monitor characteristics of polyduct transport of petroleum by-products.

This study presents a methodology based on the dual-mode gamma densitometry technique in combination with artificial neural networks to simultaneously determine type and quantity of four different fluids (Gasoline, Glycerol, Kerosene and Fuel Oil) to assist operators of a fluid transport system in pipelines commonly found in the petrochemical industry, as it is necessary to continuously monitor information about the fluids being transferred. The detection system is composed of a 661.657 keV (137Cs) gamma-ray emitting source and two NaI(Tl) scintillation detectors to record transmitted and scattered photons. The information recorded in both detectors was directly applied as input data for the artificial neural networks. The proposed intelligent system consists of three artificial neural networks capable of predicting the fluid volume percentages (purity level) with 94.6% of all data with errors less than 5% and MRE of 1.12%, as well as identifying the pair of fluids moving in the pipeline with 95.9% accuracy.

Host range expansion of an introduced insect pest through multiple colonizations of specialized clones.

Asexuality confers demographic advantages to invasive taxa, but generally limits adaptive potential for colonizing of new habitats. Therefore, pre-existing adaptations and habitat tolerance are essential in the success of asexual invaders. We investigated these key factors of invasiveness by assessing reproductive modes and host-plant adaptations in the pea aphid, Acyrthosiphon pisum, a pest recently introduced into Chile. The pea aphid encompasses lineages differing in their reproductive mode, ranging from obligatory cyclical parthenogenesis to fully asexual reproduction. This species also shows variation in host use, with distinct biotypes specialized on different species of legumes as well as more polyphagous populations. In central Chile, microsatellite genotyping of pea aphids sampled on five crops and wild legumes revealed three main clonal genotypes, which showed striking associations with particular host plants rather than sampling locations. Phenotypic analyses confirmed their strong host specialization and demonstrated parthenogenesis as their sole reproductive mode. The genetic relatedness of these clonal genotypes with corresponding host-specialized populations from the Old World indicated that each clone descended from a particular Eurasian biotype, which involved at least three successful introduction events followed by spread on different crops. This study illustrates that multiple introductions of highly specialized clones, rather than local evolution in resource use and/or selection of generalist genotypes, can explain the demographic success of a strictly asexual invader.

On the production of neutrons in laminated barriers for 10 MV medical accelerator rooms.

When space limitations are primary constraints, laminated barriers with metals can be an option to provide sufficient shielding for a radiotherapy treatment room. However, if a photon clinical beam with end point energy of 10 MeV or higher interacts with the metal inside the barriers neutrons are ejected and can result in an exposure problem inside and outside the vault. The empirical formulae existing in the literature to estimate neutron dose equivalents beyond laminated barriers do not take into account neutron production for spectra below 15 MV. In this work, the Monte Carlo code MCNP was used to simulate the production and transport of photoneutrons across primary barriers of 10 MV accelerator treatment rooms containing lead or steel, in order to obtain the ambient dose equivalents produced by these particles outside the room and in the patient plane. It was found that the neutron doses produced are insignificant when steel is present in the primary barriers of 10 MV medical accelerators. On the other hand, the results show that, in all cases where lead sheets are positioned in the primary barriers, the neutron ambient dose equivalents outside the room generally exceed the shielding design goal of 20 microSv/week for uncontrolled areas, even when the lead sheets are positioned inside the treatment room. Moreover, for laminated barriers, the photoneutrons produced in the metals are summed with the particles generated in the accelerator head shielding and can represent a significant component of additional dose to the patients. In this work, it was found that once lead sheets are positioned inside the room, the neutron ambient dose equivalents can reach the value of 75 microSv per Gray of photon absorbed dose at the isocenter. However, for all simulated cases, a tendency in the reduction of neutron doses with increasing lead thickness can be observed. This trend can imply in higher neutron ambient dose equivalents outside the room for thinner lead sheets. Therefore, when a medical accelerator treatment room is designed with laminated barriers to receive equipment with an end point energy equal to or higher than 10 MeV, not only the required shielding thickness for photon radiation attenuation should be considered, but also the dose due to photoneutrons produced in the metal, which may involve an increase of the lead thickness or even the use of neutron shielding.

Dose-image quality study in digital chest radiography using Monte Carlo simulation.

One of the main preoccupations of diagnostic radiology is to guarantee a good image-sparing dose to the patient. In the present study, Monte Carlo simulations, with MCNPX code, coupled with an adult voxel female model (FAX) were performed to investigate how image quality and dose in digital chest radiography vary with tube voltage (80-150 kV) using air-gap technique and a computed radiography system. Calculated quantities were normalized to a fixed value of entrance skin exposure (ESE) of 0.0136 R. The results of the present analysis show that the image quality for chest radiography with imaging plate is improved and the dose reduced at lower tube voltage.

Methodology for digital radiography simulation using the Monte Carlo code MCNPX for industrial applications.

This work presents a methodology for digital radiography simulation for industrial applications using the MCNPX radiography tally. In order to perform the simulation, the energy-dependent response of a BaFBr imaging plate detector was modeled and introduced in the MCNPX radiography tally input. In addition, a post-processing program was used to convert the MCNPX radiography tally output into 16-bit digital images. Simulated and experimental images of a steel pipe containing corrosion alveoli and stress corrosion cracking were compared, and the results showed good agreement between both images.

Multileaf shielding design against neutrons produced by medical linear accelerators.

This work aims at presenting a study using Monte Carlo simulation of a Multileaf Shielding (MLS) System designed to be used for the protection of patients who undergo radiotherapy treatment, against undesired exposure to neutrons produced in the components of the medical linear accelerator heads. The choice of radiotherapy equipment as the subject of study fell on the Varian Clinac 2,100/2,300 with MLC-120 operating at 18 MeV. The general purpose Monte Carlo N-Particle radiation transport code, MCNP5, was used in the computer simulation in order to determine the ambient dose equivalent, H (10), on several points on the patient's plane, with the equipment operation with and without the MLS. The results of the simulations showed a significant neutron dose reduction after the inclusion of the proposed shielding.

Neutron scattering in concrete and wood: Part II--Oblique incidence.

The knowledge of neutron reflection coefficients is of practical interest when projecting the shielding of radiotherapy rooms, since it is known that about 75% of the neutrons at the maze entrance of these rooms are scattered neutrons. In a previous paper, the energy spectra of photoneutrons were calculated, when reflected by ordinary, high-density concrete and wood barriers, using the MCNP5 code, considering normal incidence and neutron incident energies varying between 0.1 and 10 MeV. It was found that the mean energy of the reflected neutrons does not depend on the reflection angle and that these mean energies are lower in wood and barytes concrete, compared with ordinary concrete. In the present work, the simulation of neutron reflection coefficients were completed, considering the case when these particles do not collide frontally with the barriers, which constitute the radiotherapy room walls. Some simulations were also made to evaluate how neutron equivalent doses at the position of the room door is affected when the maze walls are lined with neutron absorbing materials, such as wood itself or borated polyethylene. Finally, capture gamma rays dose at the entrance of rooms with different maze lengths were also simulated. The results were discussed in the light of the albedo concepts presented in the literature and some of these results were confronted with others, finding good agreement between them.

COMPARISON OF THE NEUTRON ENERGY RESPONSE OF TWO DIFFERENT TLD ALBEDO DOSEMETERS.

Albedo dosemeters remain the most used dosemeters in neutron individual monitoring. In Brazil, most of the neutron occupational fields are from radionuclide sources, often without any moderation, where albedo dosemeters have poor energy response. The purpose of this work is to compare the HP(10) energy response of the IRD and ALNOR TLD albedo dosemeter systems, calculated by their modelling with Monte Carlo code MCNPX. Their energy responses are similar, as expected, but the IRD system is about five times more sensitive than the ALNOR one. IRD albedo system can measure the Brazilian monthly recording level of 0.2 mSv, even for bare 252Cf and 241Am-Be neutron fields. On the other hand, the ALNOR system can measure values higher than 0.2 mSv only after huge moderation of theses sources. These results show that IRD TLD albedo is more suitable than the ALNOR one to measure low doses at occupational fields from radionuclide sources.

COMMITTED EFFECTIVE DOSE DUE TO THE INTAKE OF 40K, 226Ra, 228Ra AND 228Th CONTAINED IN FOODS INCLUDED IN THE DIET OF THE RIO DE JANEIRO CITY POPULATION, BRAZIL.

Annual effective dose due to the consumption of 40K, 226Ra, 228Ra and 228Th was estimated from high-resolution gamma spectrometry, food consumption data for the inhabitants of Rio de Janeiro City and dose coefficients published by the ICRP 119. A total of 31 samples of cereals, grains, vegetables, flours, liquid and perishables were analyzed. 40K was measured in all samples, and bean sample presented highest specific concentration 489.36 ± 23.70 Bq kg-1. The highest specific concentration for 226Ra, 228Ra and 228Th was measured in pumpkin (7.82 ± 1.09 Bq kg-1), carrot (30.18 ± 1.99 Bq kg-1) and beet (2.43 ± 0.48 Bq kg-1), respectively. The highest contribution to annual effective dose came from beans (556.3 µSv), potato (12.5 µSv), carrot (10.3 µSv), banana (4.7 µSv) and beet (3.3 µSv). It was observed that updates of daily food consumption values was the main reason for an up to 10-fold difference between the annual effective dose found in the present study and literature data.

Comparison of simulated and experimental values of self-absorption correction factors for a fast and credible adjust in efficiency curve of gamma spectroscopy.

Self-absorption correction factors are fundamental in spectroscopy to correct the efficiency of the samples detection whose density is different from the radioactive standard. Mathematical simulations have been widespread as a tool to facilitate the procedure of correction factors calculation. In this paper, LabSOCS was used to calculate the self-absorption correction factor for some geometries and the values found were compared to those obtained in MCNP and experimental values. The percentage deviations found for the self-absorption correction factor calculated by LabSOCS were below 1.6% when compared to experimental values. Deviations were below 1.9% in the curve extrapolation of the experimental procedure found in literature. Results obtained show that the deviations increase proportionally to the difference between the density values of the radioactive standard and the sample. High percentage deviations were also noticed in simulations whose samples had high densities, complex geometries and low energy gamma-rays.

The use of high-density concretes in radiotherapy treatment room design.

With the modernization of radiotherapic centers, medical linear accelerators are largely replacing (60)Co teletherapy units. In many cases, the same vault housing the (60)Co teletherapy unit is reused for the linear accelerator and, when space is at a premium, high-density concrete (3.0-5.0 g/cm(3)) is employed to provide shielding against the primary, scatter and leakage radiation. This work presents a study based on Monte Carlo simulations of transmission of some clinical photon spectra (of 4-10 MV accelerators) through some types of high-density concretes, normally used in the construction of radiotherapy bunkers. From the simulations, the initial and subsequent tenth-value layers (TVL) for these materials, taking into account realistic clinical photon spectra, are presented, for primary radiation.

Neutron dose calculation at the maze entrance of medical linear accelerator rooms.

Currently, teletherapy machines of cobalt and caesium are being replaced by linear accelerators. The maximum photon energy in these machines can vary from 4 to 25 MeV, and one of the great advantages of these equipments is that they do not have a radioactive source incorporated. High-energy (E > 10 MV) medical linear accelerators offer several physical advantages over lower energy ones: the skin dose is lower, the beam is more penetrating, and the scattered dose to tissues outside the target volume is smaller. Nevertheless, the contamination of undesirable neutrons in the therapeutic beam, generated by the high-energy photons, has become an additional problem as long as patient protection and occupational doses are concerned. The treatment room walls are shielded to attenuate the primary and secondary X-ray fluence, and this shielding is generally adequate to attenuate the neutrons. However, these neutrons are scattered through the treatment room maze and may result in a radiological problem at the door entrance, a high occupancy area in a radiotherapy facility. In this article, we used MCNP Monte Carlo simulation to calculate neutron doses in the maze of radiotherapy rooms and we suggest an alternative method to the Kersey semi-empirical model of neutron dose calculation at the entrance of mazes. It was found that this new method fits better measured values found in literature, as well as our Monte Carlo simulated ones.

Monte Carlo simulation of scattered and thermal photoneutron fluences inside a radiotherapy room.

In this work, the behaviour of scattered and thermal photoneutron fluences in a radiotherapy treatment room was investigated by means of Monte Carlo simulation. The MCNP code was used to study the dependence of these neutron fluences on room design, room area and the effect of neutron moderator materials placed on the room walls. The results of the investigation showed a poor agreement between the simulations and empirical approximations, suggesting that the formulae found in the literature can underestimate the neutron flux inside a radiotherapy room.

Neutron scattering in concrete and wood.

In this work, the energy spectra of photoneutrons, scattered by ordinary, high-density concrete and wood barriers, have been evaluated using the MCNP4B code. These spectra were calculated for different scattering angles, and for incident neutron energies varying between 0.1 and 10 MeV. The results presented are required to simulate typical photoneutron fluence, produced by medical accelerators, which is scattered by the room walls and reaches the door. It was found that the mean energy of the scattered neutrons does not depend on the scattering angle. Furthermore, it was found that the scattered neutron energies are lower in wood and baryte concrete, which indicates that these materials can be used for lining the maze walls in order to reduce neutron dose at the room door. These data will help to estimate the personal dose received by the patient and staff in radiotherapy facilities.

MCNP SIMULATION OF THE HP(10) ENERGY RESPONSE OF A BRAZILIAN TLD ALBEDO NEUTRON INDIVIDUAL DOSEMETER, FROM THERMAL TO 20 MeV.

The Brazilian Instituto de Radioproteção e Dosimetria (IRD) runs a neutron individual monitoring system with a home-made TLD albedo dosemeter. It has already been characterised and calibrated in some reference fields. However, the complete energy response of this dosemeter is not known, and the calibration factors for all monitored workplace neutron fields are difficult to be obtained experimentally. Therefore, to overcome such difficulties, Monte Carlo simulations have been used. This paper describes the simulation of the HP(10) neutron response of the IRD TLD albedo dosemeter using the MCNPX transport code, for energies from thermal to 20 MeV. The validation of the MCNPX modelling is done comparing the simulated results with the experimental measurements for ISO standard neutron fields of (241)Am-Be, (252)Cf, (241)Am-B and (252)Cf(D2O) and also for (241)Am-Be source moderated with paraffin and silicone. Bare (252)Cf are used for normalisation.

Monte Carlo simulation of response function for a NaI(Tl) detector for gamma rays from 241Am/Be source.

The general purpose computer code MCNP4B was used to simulate the response function of a bare NaI(Tl) detector crystal for gamma rays from an 241Am/Be source capsule. The simulated spectral shape generated by the MCNP4B code was compared with the measured spectral shape obtained using a gamma-ray spectrometer with a cylindrical shape, 7.62 cm x 7.62 cm, NaI(Tl). In general, the agreement between the simulation and the experimental response function was good.

A study of neutron spectra from medical linear accelerators.

Medical accelerators with photon energies over 10 MeV generate an undesired fast neutron contamination in the therapeutic beam. In this work, the Monte Carlo code MCNP was used to simulate the transport of these photoneutrons across the head of various medical accelerators of high energy. The average and most probable neutron energies were obtained from these spectra, before and after crossing the accelerator shielding. The degradation of these spectra, when they cross concrete barriers with thickness which vary between 25 and 100 cm, was also studied.

Moderator-collimator-shielding design for neutron radiography systems using 252Cf.

This paper is concerned with the presentation of a study of the general design of an optimized neutron radiography system that utilizes 252Cf. Moderation, collimation and shielding aspects are considered. A Monte Carlo code, MCNP, was used to obtain a maximum and more homogeneous neutron flux in the collimator outlet next to the image plane, taking into account geometric characteristics and an adequate radiation shielding strategy that complies with the radiological protection rules. Among the various moderator materials investigated, the high density polyethylene proved to be the most efficient, with a thermalization factor of 56 cm2. Using a collimator design assembly it was possible to obtain a normalized thermal neutron flux, at the image plane, equals 6 x 10(-6) n cm(-2) s(-1) at an effective collimator ratio of 7.5, or 3.2 x 10(-7) n cm(-2) s(-1) at an effective collimator ratio of 50. The total dose equivalent rates were significantly reduced by the shielding optimization process.

Explosives detection using prompt-gamma neutron activation and neural networks.

This work describes a study of the application of a neural network to determine the presence of explosives using the neutron capture prompt gamma-ray spectra of the substances as patterns which were simulated via Monte Carlo N-particle transport code, version 4B. After the training of the neural networks, it was possible to determine the presence of the C-4 explosive, even when they were occluded by several materials. The neural network was a powerful tool, able to recognize prompt gamma-ray explosive patterns in spite of the presence of occluding materials. Besides that, the network was able to generalize, identify the presence of explosive in cases in which it had not been trained. In that way, it was revealed as a potential tool for in situ inspection systems.