A. X. da Silva

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 μSv/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 μSv 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.

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.

Photon doses at the entrance of 60Co and low-energy medical accelerator rooms under unusual irradiation conditions

In this paper, the general-purpose Monte Carlo code MCNP5 was used to study the dose variance due to the position of medical linear accelerators, under unusual conditions, for shielding design of radiotherapy facilities. It was found that the computational methods generally used to estimate the scattered photon doses at the entrance of radiotherapy unit vaults provide conservative results when compared with the MCNP results, considering the standard condition. On the other hand, for the situations where the axis of gantry rotation is redirected at, for example, 45 degrees with respect to the walls of the room, the photon doses at the entrance can reach values up to seven times higher than those obtained under the standard condition, depending on the energy of the primary beam.

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.

Neutron tomography using projection data obtained by Monte Carlo simulation for nondestructive evaluation

This work present the application of a computer package for generating of projection data for neutron computerized tomography, and in second part, discusses an application of neutron tomography, using the projection data obtained by Monte Carlo technique, for the detection and localization of light materials such as those containing hydrogen, concealed by heavy materials such as iron and lead. For tomographic reconstructions of the samples simulated use was made of only six equal projection angles distributed between 0o and 180o, with reconstruction making use of an algorithm (ARIEM), based on the principle of maximum entropy. With the neutron tomography it was possible to detect and locate polyethylene and water hidden by lead and iron (with 1cm-thick). Thus, it is demonstrated that thermal neutrons tomography is a viable test method which can provide important interior information about test components, so, extremely useful in routine industrial applications.

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.

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.

SU‐GG‐I‐68: Dosimetryc Study to Voxel Model Applied to Carciac Procedures in Nuclear Medicine

Purpose: The objective of this study is to realize a computational dosimetry on a simplified model of miocardic perfusion, considering the usual different ways to describe the spectra emitted for Tc‐99m. Method and Materials: The GEANT4 code was used to simulate two geometries: the radial dose distribution from an isotropic point source described by the three spectra, modeled with photon emission from the center of a 1.3‐m‐diam sphere of muscle tissue; and a second considering an uniform distribution of Tc‐99m in the heartmuscle of the adult male voxel model MAX, to take the organs dose distribution. The different ways to describe the spectra emitted for Tc‐99m was: monoenergetic spectrum of 140 keV; three photonsemissions spectrum (2.1, 141 and 143 keV) and total spectrum (including characteristics x‐rays and Auger electrons).Results: For distance low than 1 cm from the point source the radial dose distribution is higher for total spectrum. The radiation dose in a sphere with 0.01 mm defined at the center of the sphere of muscle was 0.369 mGy Bq1 h1 (total spectrum), 0.276 mGy Bq1 h1 (three photonspectrum), and 0.005 mGy Bq1 h1 (monoenergetic spectrum). This data shown that include Auger electrons,characteristic x‐rays, and low energy gamma give a significant contribution to total energy deposition. This results corroborates the simulations realized using the voxel model. The data variation shown that monoenergetic and three gamma spectra, comparing to total spectra simulated, produce a decrease on absorbed dose on cardiac tissue of 19.2% and 7.1%, respectively. Conclusion: The results shown that the combined transport of electrons Auger and characteristics x‐rays of the Tc‐99m increase the radiation dose, especially on organs/tissues closer to those have had absorbed the radiopharmaceuticals. This study indicates the importance on describe the complete radiopharmaceuticalsspectrum on dosimetric simulations in Nuclear Medicine.