Journal of Radiation Safety and Measurement
Year:2019 | Volume:7 | Issue:3|Papers Count:7

Brachytherapy is one type of internal radiation therapy where radiation sources, which are usually encapsulated are placed as close as possible to the tumor site inside the patient's body. In this technique, it is important to determine dose distribution around the brachytherapy capsule. Hereby, in this paper, dosimetric parameters of I-125 brachytherapy source model 6711 are estimated according to TG-43U1 protocol using GATE 8. 1 Monte Carlo code. In this work, we used GATE_v8. 1 to calculate dosimetric parameters of the I-125 brachytherapy source model 6711. At first, validation of the GATE platform were performed by some criteria such as radial dose function, 2D anisotropy function inside liquid water according to the AAPM TG-43U1. On the other hand, since the attenuation coefficient of the sources in the water phantom is different from that of various tissues, the effects of the various tissues on the radial dose function parameter of the I-125 brachytherapy source were investigated using GATE 8. 1 code. Dosimetric parameters of simulated I-125 brachytherapy capsule show good consistency compared with the other study. The maximum average deviation was about 3. 61% and 7. 29% at radial dose function and anisotropy function, respectively. The relative deviation of radial dose function in the fat, muscle, breast and lung tissue compared with water phantom in radial distance of 5cm were about 68. 73%, 10. 98%, 25. 83% and 12. 23%, respectively. There was a good agreement between the results of this work and other study in calculation of dosimetric parameters of brachytherapy I-125 source base on the recommendations of TG-43U1 protocol. The results show that the dosimetric parameters of I-125 brachytherapy can be calculated using the GATE code and appropriate physic list in spite of low energy of radiation and high variation in dose rate with increasing distance from the center of the source. The results of the dose calculation in different phantom could be used in clinical treatment planning systems.

In order to recovery the thermoluminescence (TL) sensitivity of CaF2: Mn (TLD-400) thermoluminescent dosimeter it was exposed to high gamma radiation dose of 10kGy, then by applying standard annealing followed by irradiation to 1Gy gamma dose, the TL signal was recorded. The TL response reduced due to the damage effect of high dose irradiation in comparison with control dosimeter with received dose of 1Gy. By applying different thermal treatment and variation of annealing time, the appropriate annealing time for recovery of sensitivity

In the recent years some studies has been done to consider the capability of Tehran Research Reactor for Boron neutron capture therapy (BNCT). The purpose of this study is to evaluate the sensitive organs dose during the treatment of patient with deep brain tumor by TRR. The calculation has been carried out using the Monte Carlo code MCNPX for ZUBAL head and neck phantom. The method was tested for 3 different boron concentration injected to patient located in TRR thermal column from human head. The total dose (Dw) was defined as the sum of physical dose components (Di) multiplied by weighting factors (wi) of each dose component in a tissue. The MCNP simulations were carried out with the MCNPX version of 2. 6. In order to calculate the dose absorption, the tally F4 was used. For the dose conversion, pointwise KERMA factors from ICRU-46 were directly input with DE and DF cards. Treatment time based on absorbing 20 Gy-Eq by tumor approximately from 15 to 30 minutes changes for all trials. The results indicate that increasing boron concentration causes decreasing lens and thyroid dose received. In all trials parietal lobes receive the most dese rather than other parts. It was found that fast neutron dose component has most contribution in skin and lenses doses. But for the thyroid gamma dose component has most contribution. It is considered that side-irradiation would not be safe treatment for vital organs and take long time.

Analyzes of environmental samples regarding their radioactivity is of important concern for health purposes. We need standard sources to determine radioactive components and their activities. These sources are usually produced regarding type of the sample. One of the fundamental and precise tools to recognize radioactive materials and their activities is HPGe detector. To reach this goal, the detector needs to be scaled by standard sources with the same shape and the same components with environmental samples. It is also needed to determine the efficiency of the detector in a wide range of energies. The most precise way to determine detector efficiency is by doing experiment using standard sources. Since experimental methods are time consuming and difficult to apply in some cases, it is worth using simulating method which takes a short time and is precise. In the first part of this research the detector efficiency is determined in two different ways: experimental and simulating for energies from 121 keV to 1408 keV for a volume source. In simulating method, the extracted results from the Monte Carlo code MCNPX was in agreement with experimental data. In the second part, the activity of Eu152 and Cs137 of volume source which is standardized by these components was calculated using efficiencies and simulating outputs and experimental data by Eu152 point source and it is shown that it is possible to use a point source to determine the activity of radionuclide with unrecognized activity in volume environmental samples.

The 252Cf brachytherapy source is a spontaneous fission decay source, which is used as a neutron-emitting source. In addition to neutrons emitted from this source, gamma rays are also emitted with the average energy of 1 MeV. In this study, using the Monte Carlo N-Particle code (MCNPX), the absorbed dose rates of the neutrons, the primary gamma and the secondary gamma that generated by thermal neutron capture in the hydrogen of the water were calculated at different distances from the source in the water phantom. Also, the equivalent dose rates of the total, the neutrons, and the gammas were obtained at different intervals from the source. The results indicate that gamma rays from this source can provide significant energy at distances close to the source, so the contribution of these rays to total absorption doses should be calculated. The neutron dose rate and total gamma decrease with increasing distance from the source and at the distances close to the source are deposited the most energy. The equivalent dose of neutrons at distances close to the source (lower than 2cm) with differences to gamma rays has the highest effect at the equivalent total dose. So, the values at the distances of 0. 5 to 2. 0 cm from the source reach the value of 46. 30 cSv/h. µ g to 2. 95 cSv/h. µ g, while the equivalent dose rate of gamma at distances of 0. 5 to 2. 0 cm from the source is 4. 30 cSv/h. µ g to 0. 272 cSv/h. µ g.

Comparison of radiation shields of different thicknesses and structures (especially different particle sizes of their constituents) is helpful to select an appropriate shield for patient protection in medical imaging tests particularly CT-scan. In this study micro-and nano-powders of bismuth were mixed with silicone matrix to make composite radiation shields for breast. The shields were provided in dimensions of 20×20 cm, thickness of 1 mm, and bismuth weight ratio of 10%. The chest CT test (TOSHIBA 16 multi-slice device) was performed on a female phantom with normal breast size in 100 kVp and 50 mA with slice thickness of 0. 5 mm and pitch of 1 mm. To avoid image artifacts, a foam layer of 1 cm was wrapped around the phantom under the shields. Measurement of dose in the surface (equivalent to skin) and forth layer (equivalent to glandular tissue) of phantom breast was done using thermoluminescence dosimeter. Results showed that dose on the surface of phantom breast declined up to 12% and 18. 4% in the presence of silicon-bismuth micro-and nanocomposite shields, respectively, and both composites reduced the breast dose significantly (p<0. 5). Also, the nanocomposite shields reduced radiation dose more than the microcomposite shields. Additionally, less noise variation was observed in CT images acquired with the nanocomposite shields.

One of the most widely used diagnostic devices in medical imaging is computed tomography scanning with the use of X-ray ionization. After years of using this tool in the diagnosis and treatment, external doses of these beams can pose a risk of secondary cancer, which is significant in terms of radiation safety and protection. In this paper, using the Monte Carlo code MCNPX2. 6 performed over three phases, a 40-years-old male MIRD and CT room was designed and simulated at different locations of the CT room and the absorbed dose rate was calculated. In the first phase, by calculating and simulating, the mean absorption dose rate was obtained and then the safe and unsafe locations were identified. In the second step, by analyzing the results, the highest mean dose rate of the phantom absorption dose was obtained in the simulation of abdominal area scans and its internal organs. In the third stage of this article, by designing a lead shield, a solution was proposed to reduce the effectiveness of absorptive doses that could improve the safety level.