A dosimetric study of prostate brachytherapy using Monte Carlo simulations with a voxel phantom, measurements and a comparison with a treatment planning procedure.

In prostate brachytherapy treatments, there is an initial swelling of the prostate of the patient due to an oedema related to the insertion of the seeds. The variation of the prostate volume can lead to variations in the final prescribed dose in treatment planning procedures. As such, it is important to understand their influence for dose optimisation purposes. This work reports on a dosimetric study of the swelling of the prostate in prostate brachytherapy using Monte Carlo simulations. Dosimetric measurements performed on a physical anthropomorphic tissue-equivalent prostate phantom and thermoluminescent dosimeters (TLDs) were used to validate the MC model. Finally the MC model was also used to simulate prostate swelling in a real treatment planning procedure. The obtained results indicate that the parameters mentioned above represent a source of uncertainty in dose assessment in prostate brachytherapy, and can be detrimental to a correct dose evaluation in treatment plannings, and that these parameters can be accurately determined by means of MC simulations with a voxel phantom.

Application of digital image processing for the generation of voxels phantoms for Monte Carlo simulation.

This paper presents the application of a computational methodology for optimizing the conversion of medical tomographic images in voxel anthropomorphic models for simulation of radiation transport using the MCNP code. A computational system was developed for digital image processing that compresses the information from the DICOM medical image before it is converted to the Scan2MCNP software input file for optimization of the image data. In order to validate the computational methodology, a radiosurgery treatment simulation was performed using the Alderson Rando phantom and the acquisition of DICOM images was performed. The simulation results were compared with data obtained with the BrainLab planning system. The comparison showed good agreement for three orthogonal treatment beams of (60)Co gamma radiation. The percentage differences were 3.07%, 0.77% and 6.15% for axial, coronal and sagital projections, respectively.

Percentage depth dose evaluation in heterogeneous media using thermoluminescent dosimetry.

The purpose of this study is to investigate the influence of lung heterogeneity inside a soft tissue phantom on percentage depth dose (PDD). PDD curves were obtained experimentally using LiF:Mg,Ti (TLD-100) thermoluminescent detectors and applying Eclipse treatment planning system algorithms Batho, modified Batho (M-Batho or BMod), equivalent TAR (E-TAR or EQTAR), and anisotropic analytical algorithm (AAA) for a 15 MV photon beam and field sizes of 1 x 1, 2 x 2, 5 x 5, and 10 x 10 cm 2 . Monte Carlo simulations were performed using the DOSRZnrc user code of EGSnrc. The experimental results agree with Monte Carlo simulations for all irradiation field sizes. Comparisons with Monte Carlo calculations show that the AAA algorithm provides the best simulations of PDD curves for all field sizes investigated. However, even this algorithm cannot accurately predict PDD values in the lung for field sizes of 1 x 1 and 2 x 2 cm 2 . An overdosage in the lung of about 40% and 20% is calculated by the AAA algorithm close to the interface soft tissue/lung for 1 x 1 and 2 x 2 cm 2 field sizes, respectively. It was demonstrated that differences of 100% between Monte Carlo results and the algorithms Batho, modified Batho, and equivalent TAR responses may exist inside the lung region for the 1 x 1 cm 2 field.

Diagnostic assessment to estimate and minimize neutron dose rates received by occupationally exposed individuals at cyclotron facilities.

Since 2003, radiopharmaceuticals for medical diagnostic purposes have been produced at the Instituto de Engenharia Nuclear, in Brazil, using two cyclotron accelerators - CV-28 and RDS111. As a result of the ever increasing production, a diagnostic assessment to reduce neutron dose rates received by occupationally exposed individuals during irradiation processes has been developed. The purpose of this work is to present this assessment, which is currently being applied to both the Fluorine and Iodine targets of CV-28 and RDS111 cyclotron accelerators.

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.

Optimal shielding design for bunkers of compact cyclotrons used in the production of medical radionuclides.

PURPOSE: There are several options to consider in the design of a vault that will house a cyclotron for radioisotopes production with regards to the door entrance. Alternatives are a direct-shielded door, a simple maze, or a double-legged maze. In this work, the impact of the neutron and photon doses at the vault entrance was evaluated for these options. METHODS: Monte Carlo simulations were carried out in order to assess photon and neutron ambient dose equivalents. Simulations results were compared to experimental measurements taken inside a vault with a direct-shielded door. RESULTS: The double-legged maze is the configuration that provides the higher degree of radiological protection at the vault entrance. In addition, the fact of the location of the cyclotron target plays an important role in the ambient dose equivalents. CONCLUSIONS: The comparison performed between measurements and results of MCNP simulations confirmed a favorable agreement; maze legs reduce the neutron energies at the entrances of the vault. However, a degree of dose reduction similar to the one obtained for a two-legged maze can be achieved with the addition of shielding against neutrons in the inner maze entrance to act as a second door. A choice of a vault design is more evident by comparing the results of this study.

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.

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.

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.

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.

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 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.

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.

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.

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.

Goiânia: 12 years after the 137Cs radiological accident.

In 1987, in the city of Goiânia, Brazil, occurred one of the worst radiological accidents ever reported. The remains of 137Cs contamination in a terrain where part of a radiotherapy unit had been manipulated in 1987 were measured in 1999-2000, and some of the results are presented here. Using the technique of gamma ray spectrometry in situ and in the laboratory, the ambient dose equivalent rate at 1 m above the ground and 137Cs concentration in soil were determined. Values higher than the ones established by the Brazilian National Nuclear Energy Commission (CNEN) as action levels in 1987, namely 0.8 microGy x h and 22.5 kBq x kg(-1), were obtained in that terrain. The 137Cs distribution profile in the soil shows high values of the specific activity in a layer located at a depth of 10-40 cm from the surface, where the soil is mixed with rubble, reaching values as high as 175 kBq x kg(-1).

Measurements performed in Goiânia after a new intervention action in 2001.

Brazil's worst radiological accident took place in 1987, in the city of Goiânia. In 1999 and 2000, detailed measurements of 137Cs contamination were performed in junkyard II, one of the places involved in the accident. High values of 137Cs activity per unit mass were found in soil layers at depths between 10 and 40 cm from the surface, reaching values as high as 175 kBq x kg(-1). High values of 137Cs concentration in fruits and plants were also observed. Moreover, values of ambient dose equivalent rate at 1 m above the ground were found to be higher than the limit of 1.0 microSv x h(-1) set by the Brazilian National Nuclear Energy Commission (CNEN) in 1987. In February 2000, the CNEN was informed about the results of our measurements. As consequence, in August 2001, the CNEN performed a new intervention action in the area, covering all its extension with a concrete layer and removing some plants and trees. The new remedial action reduced the dose rate to approximately 13% of the value prior to covering the site in concrete, reaching values below the CNEN limit, as demonstrated by the measurements presented here.

Remains of 137Cs contamination in the city of Goiânia, Brazil.

The results of measurements, performed in 1999, of the remaining 137Cs contamination in some of the sites where fragments of a radioactive source of a teletherapy unit had been manipulated in 1987 are presented. This episode occurred in the city of Goiânia, during Brazil's worst radiological accident ever reported. Using the technique of gamma ray spectrometry, analyses of both surface and profile soil and vegetable samples were made. High values of 137Cs activity per unit mass were found in soil layers at depths between 10 and 40cm from the surface. Some values exceeded by up to eight times the action level of 22.5 kBq x kg(-1) proposed by the Brazilian National Nuclear Energy Commission (CNEN) during the decontamination process at the time of the accident, for the first year after the accident. Absorbed dose rates at 1 m above the ground were calculated from the data of 137Cs concentration in the soil and compared with those obtained from in situ gamma ray spectrometry and from thermoluminescence dosimetry.

Ambient dose equivalent rate in Goiânia 12 years after the 137Cs radiological accident.

This paper describes the situation of ambient dose equivalent rates in four of the main foci of 137Cs contamination in the city of Goiânia, Brazil, in 1999, 12 y after one of the worst radiological accidents in the world. During the decontamination, all the buildings of the three highly contaminated sites were demolished and the top soil removed. Afterwards, the soil of two of these lots was covered with concrete, and they remain vacant today. The soil of the third of these lots, identified here as E, previously known as junkyard II, was covered only with clean soil. Three to four years after the accident, new houses were constructed on this lot, and some very poor people live and work there collecting recyclable material. Gamma ray spectrometry, with a portable survey meter, was performed in the quoted places along with outdoor measurements in many other locations of Goiânia. The average ambient dose equivalent rate due to natural background radiation from radionuclides in the soil and cosmic radiation in non-contaminated areas of the city of Goiânia is 62 nSv h(-1). In most of the highly contaminated sites during the accident, the average ambient dose equivalent rate ranged from around 100 to 1,000 nSv h(-1). The only exception was site E, where values of ambient dose equivalent rate as high as 2.6 microSv h(-1) were found.