صدیقه کاشیان

Detection and dosimetry of ionizing radiation are important and fundamental issues in the nuclear industry. In this study, calculations related to the optimal thickness of a new beta detector for 90Sr source based on Polycarbonate/Bismuth oxide composite material (PC-Bi2O3) for different weight percentages of bismuth oxide fillers in polymer matrix, namely 0 wt%, 10 wt%, 20 wt%, 30 wt%, 40 wt% and 50 wt% were carried out using the ESTAR program. Results showed that the weight percentage of bismuth oxide particles was effective in determining the range and stopping power of the electrons in the composite detector. Finally, for 90Sr, using the evaluation of beta particles range in the aforementioned composite material, the optimal thickness for this detector was calculated.

In this study, calculations related to the mass attenuation coefficient (µ/ρ) of Polyethylene/Bismuth Oxide composite (HDPE-Bi2O3) for different weight percentages of Bi2O3 in the polymer matrix, namely 0 wt%, 1 wt%, 2 wt%, 3 wt%, 5 wt% and 8 wt% for photons with 59 keV energy that can be used in dental radiography centers were simulated using the MCNP code and results were compared with XCOM data. The simulation results of the mass attenuation coefficient of the composite in different weight fractions via the methods of MCNP and XCOM at 59 keV showed a good agreement with each other. The results of this study showed that the Polyethylene/Bismuth Oxide composite material in optimal weight fraction and thickness can be used as a radiation shield in dental radiography centers.

Considering the importance of quality control of radiotherapy systems and 3D dose mapping possibility done with a Fricke agarose gel dosimeter, this dosimeter based on ferrous solution and agarose gel was prepared. The prepared Fricke gel dosimeters was investigated under gamma irradiation with a dose range up to 20 Gy. Concerning the effect of the gel dosimeter constituent’s concentration on the response of the dosimeter, different compositions were surveyed for better sensitivity and repeatability. So, the responses were evaluated for different concentrations of ferrous ion, sulphuric acid, and the indicator. We found that the Fricke agarose gel dosimeter has a linear behavior up to 20 Gy. In addition, the best composition of Fricke agarose gel dosimeter with high sensitivity and stability was determined to be 0.2 mM ferrous, 25 mM sulphuric acid, and 0.15 mM Xylenol orange.

External radiotherapy is used for treatment of various types of cancers. Due to the impossibility of measuring the absorbed dose delivered to different organs during irradiation, treatment planning systems (TPSs) have been utilized for calculation of absorbed dose before a radiotherapy procedure. Thus, the accuracy and precession of the TPS is essential.The aim of this study is investigation of accuracy the TPS based on Collapse Cone Convolution (CCC) algorithm in a lung tissue equivalent material. The charge generated in the sensitive volume of PTW-30013 ionization chamber in water and lung tissue equivalent phantoms placed in the radiation fields of Primus 6MV linac was calculated using MCNP.4C code and in the ratio of generated charge in this phantom was determined. To validate the simulations, the ratio of generated charge in sensitive volume of ionization chamber in mentioned phantoms was determined experimentally. The agreement between the calculations and measurement confirm the simulation method. The calculated absorbed dose delivered in the lung tissue equivalent material for 200 MU radiation was 154.99 cGy using simulations. The CCC algorithms predicted this value as 163.98 cG. As well as, the absorbed dose in different depths was measured using GR-200 Dosimeters. The relative differences between the values obtained by simulation and CCC algorithm and between the results of TLD measurements and CCC algorithms are more than 5.5% and 14%, respectively. So, by considering the acceptable uncertainties suggested by ICRU for TPS algorithms and the results of this work, it can be concluded that, the CCC algorithm is not sufficiently accurate to determine of absorbed dose delivered to tissues with density lower than water.

Precise calibration of energy and efficiency of gamma spectrometry systems for different geometries is very important. In most cases, preparation of these sources for educational and research centers  is very difficult due to high costs and lack of access to the reactor. The purpose of this paper is to explain the procedure for production of calibration sources 65Zn and 60Co using instrumental neutron activation analysis (INAA) method. These sources are used for calibration of gamma spectrometry systems. To do so, 65Zn (3.83 and 9.38 µCi) and 60Co (3.65 and 3.85 µCi) were produced. The quality control of the sources indicates the acceptable quality of the produced sources.

In this paper, the ratio of adsorbed dose to the tissues of muscle, blood and bone to water for medical accelerators of Electa 25MV and Siemens 6MV is calculated using an analytical method based on cavity theory and Monte Carlo simulation. To confirm the validation of these methods, the ratio of absorbed dose to air to water was determined experimentally for radiation field of 60Co in Secondary Standard Dosimetry Laboratory (SSDL). The relative difference between the measured value and calculated values was less than 10% which confirms the validation of analytical and simulation methods.
The ratio of absorbed dose to muscle and blood to water is close to unity due to their atomic composition and density similar to water. The ratio of absorbed dose to bone to absorbed dose to water is calculated 0.96 and 0.91 for radiation fields of Electa 25MV and Siemens 6MV, respectively due to the discrepancy between atomic composition and density of bone and water. Further, the obtained results reveals that the ratio of absorbed dose to different tissues to absorbed dose to water is depended on radiation field energy. The results of this paper can be useful to determine the absorbed dose delivered to an organ in a treatment planning systems.

Fast neutron flux (14.8 MeV) of a neutron generator has been measured by activation technique. The measurements performed using Cu and Ni threshold detectors. 62Cu and 57Ni were produced through 63Cu (n, 2n) 62Cu and 58Ni (n, 2n)57Ni reactions. They decay by emitting 511 keV and 1377 keV gamma rays, respectively. The half life of 62Cu is 9.74min and that of 57 Ni is 36 hours. The flux of neutron has been calculated by measuring the activity after the irradiation time. Gamma spectroscopy of the activated foils was performed using a HPGe detector. By employing this technique the neutron flux of 2.64×107±3%n/s was obtained for 60 μΑ deuteron of 110 keV energy, bombarding a solid target of 3H.