New filter for iodine applied in nuclear medicine services.

In Nuclear Medicine, radioiodine, in various chemical forms, is a key tracer used in diagnostic practices and/or therapy. Medical professionals may incorporate radioactive iodine during the preparation of the dose to be administered to the patient. In radioactive iodine therapy doses ranging from 3.7 to 7.4 GBq per patient are employed. Thus, aiming at reducing the risk of occupational contamination, we developed a low cost filter to be installed at the exit of the exhaust system (where doses of radioiodine are handled within fume hoods, and new filters will be installed at their exit), using domestic technology. The effectiveness of radioactive iodine retention by silver impregnated silica [10%] crystals and natural activated carbon was verified using radiotracer techniques. The results showed that natural activated carbon and silver impregnated silica are effective for I2 capture with large or small amounts of substrate but the use of activated carbon is restricted due to its low flash point (423 K). Besides, when poisoned by organic solvents, this flash point may become lower, causing explosions if absorbing large amounts of nitrates. To hold the CH3I gas, it was necessary to use natural activated carbon since it was not absorbed by SiO2+Ag crystals. We concluded that, for an exhaust flow range of (145 ± 2)m(3)/h, a double stage filter using SiO2+Ag in the first stage and natural activated carbon in the second stage is sufficient to meet radiological safety requirements.

Detection of drugs and explosives using neutron computerized tomography and artificial intelligence techniques.

In this study the development of a methodology to detect illicit drugs and plastic explosives is described with the objective of being applied in the realm of public security. For this end, non-destructive assay with neutrons was used and the technique applied was the real time neutron radiography together with computerized tomography. The system is endowed with automatic responses based upon the application of an artificial intelligence technique. In previous tests using real samples, the system proved capable of identifying 97% of the inspected materials.

TLD environmental monitoring at the Institute of Nuclear Engineering in Brazil.

Since 2003 the Institute of Nuclear Engineering in Rio de Janeiro city, Brazil, operates a new cyclotron, RDS-111, to produce (18)F-Fluorodeoxyglucose to be used in nuclear medicine. Additionally, the IEN radioactive waste repository has been enlarged during the past last years, receiving a considerable amount of radioactive materials. Therefore, it became necessary to evaluate a possible increase of the environmental gamma exposure rates at the institute site due to the operation of the new accelerator and the enlargement of the institute waste repository as well. LiF:Mg,Cu,P, TLD-100H, and TL detectors were employed for environmental kerma rate evaluation and the results were compared with previous results obtained before the RDS-111 operation initialisation and the enlargement of IEN waste repository. No significant contribution for the enhancement of environmental gamma kerma rates was detected.

Gadolinium sheet converter for neutron radiography.

This work describes a methodology developed for the confection of gadolinium sheet converter for neutron radiography using the gadolinium chloride (GdCl3) as material converter. Though manufactured at a relatively low cost, they are as good as the sheet converter on the market. Here, we present neutron radiography of the penetrameter, the edge spread function, the modulation transfer function and characteristic curves for each set sheet-AA400 Kodak film.

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.

Dose reconstruction of a Brazilian industrial gamma radiography partial-body overexposure case.

In May 2000, an operator of a (60)Co industrial gamma radiography apparatus, during a routine service, was involved in a partial-body radiological accident, which caused serious injuries to his left hand. Dose reconstruction was started aiming to assess the radiation doses, in order to assist the medical staff in the evaluation and prescription of suitable medical procedures for the patient's treatment and follow-up. This work presents the dose reconstruction used for assessment of the distribution of doses on the patient's left hand, which was made using two methods: physical and computational techniques. For the first technique a physical hand simulator was built. The computational method was performed using microcomputer software for external dose calculations, named 'Visual Monte-Carlo-VMC', together with a hand voxel simulator. The values obtained through both methods for the distribution of absorbed doses on the operator's left hand were compared. About half of them were similar within a range of uncertainty of 20%.

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.

The study of the effects of low-level exposure to ionizing radiation using a bio-indicator system.

The object of the present study was to apply the high sensitivity of a botanical mutagenicity test (Tradescantia stamen-hair mutation bioassay), in "in situ" bioassays to determine the responses induced by the exposure to low levels of ionizing radiation. Mutagenesis was evaluated during 28 days, in an environment that presented gamma radiation exposure rates of 1.6 (control), 25.0 and 75.0 microR min(-1). From the results we observed that there was no linear response throughout the exposure time when compared mutagenic events and gamma radiation exposure rates were compared. The results obtained in the second week of exposure of Tradescantia showed that after exposure to 25.0 and 75.0 microR min(-1) there was a significant increase (p<0.05) in the mutation levels in Tradescantia stamen-hair. In the third and fourth week after exposure 25.0microR min(-1), the plants demonstrated a mutation rate that was not significantly different from the control (p>0.05).

Calculation of conversion coefficients for clinical photon spectra using the MCNP code.

In this work, the MCNP4B code has been employed to calculate conversion coefficients from air kerma to the ambient dose equivalent, H*(10)/Ka, for monoenergetic photon energies from 10 keV to 50 MeV, assuming the kerma approximation. Also estimated are the H*(10)/Ka for photon beams produced by linear accelerators, such as Clinac-4 and Clinac-2500, after transmission through primary barriers of radiotherapy treatment rooms. The results for the conversion coefficients for monoenergetic photon energies, with statistical uncertainty <2%, are compared with those in ICRP publication 74 and good agreements were obtained. The conversion coefficients calculated for real clinic spectra transmitted through walls of concrete of 1, 1.5 and 2 m thick, are in the range of 1.06-1.12 Sv Gy(-1).

Neutron dose rate evaluation for medical linear accelerators.

During X-ray therapeutic irradiation with energies above the threshold of (X,n) reactions in the structural materials of medical accelerators, a photoneutron fluence is generated. In Brazil, no measurements of neutron doses in radiotherapy rooms are being done yet, when licensing these equipment. Consequently, it is very important to obtain accurate analytical formulae and/or simulation of these dose rates, in order to estimate the increase in dose received by the patient and staff, as well as to correctly project the additional shielding for the treatment room. In this work, we present MCNP simulation of dosimetric quantities at the isocentre of some models of high-energy linear accelerators, and compare it with the values given by the manufacturers, finding good agreement between both.

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.

Shielding design studies for a neutron irradiator system based on a 252Cf source.

This study aims to investigate a shielding design against neutrons and gamma rays from a source of 252Cf, using Monte Carlo simulation. The shielding materials studied were borated polyethylene, borated-lead polyethylene and stainless steel. The Monte Carlo code MCNP4B was used to design shielding for 252Cf based neutron irradiator systems. By normalising the dose equivalent rate values presented to the neutron production rate of the source, the resulting calculations are independent of the intensity of the actual 252Cf source. The results show that the total dose equivalent rates were reduced significantly by the shielding system optimisation.

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.