نشریه سنجش و ایمنی پرتو
سال دوم شماره 2 (پیاپی 6، بهار 1393)

During the last few years¡ real time gamma and neutron dosimeters have been developed and semiconductor diodes are frequently used in these dosimeters. Semiconductor diodes are used to charged particle detection. For sensitizing the detector to neutrons¡ a converter layer is contacted on the front surfaces of them. Incident neutrons interact with the converter and produce charged particles that can deposit their energy in the detectors and produce a signal. To cover the energy range of thermal and fast neutrons¡ a silicon diode detector with thermal neutron converters (6LiF) and fast neutron converter (polyethylene) have been used. Response of the diode detector to neutron radiation is depending on the neutron converter and varying of the neutron converter characteristics is performed for achieving dosimetry in the different energies. To achieve proper dosimetry response in a wide range of energy response 0.01 eV-15 MeV¡ the optimum thickness of the thermal and fast neutron converter is determined 4 micron and one millimeter respectively. The results show that the dosimetry response in the different neutron energies¡ to some extent was appropriate¡ and the deviation is acceptable in the wide range of energy.

The effect of the different parameters on the thermoluminescence behavior of LiF:Mg,Ti powder produced by melting method is reported. Standard powder of LiF:Mg,Ti manufactured by Harshaw company, was converted to smaller grains size, to obtain different meshes after melting and cooling steps. Powder in this study are of mesh size: 80 to 100, 120 to 140 and 170 to 200 µm. The results show that the melting process affect the glow curve shape, fading factor and repeatability factor of the produced phosphor.

Nowadays medical imaging systems like PET/CT are widely used in cancer diagnosis. Due to serious and irreversible harms of ionization radiations, protection of all those who are exposed is the main concern of health issues. The main basis of the calculation of the shielding design in the medical imaging systems is that the absorbed dose should not exceed the allowed limit. Using the Monte Carlo N-Particle Transport Code (MCNP), a potential suite for a positron emission tomography/Computed Tomography (PET/CT) facility was designed. The design considers three different areas as required in a PET/CT facility: scanning room, waiting area, and injection room. MCNP-X calculation code is used to ensure that the controlled and uncontrolled areas (with allowed dose of 0.1 mSv/week and 0.02mSv/week, respectively) does not exceed the limits. Since this center accepts four patients every day with mean total activity of 77 mCi, the dose rates are calculated from 2.6E-06 mSv/Week to 9.35E-3 mSv/week. The simulation results revealed that even if this center increase its service to eight patients per day, the dose rates will be from 4.88E-06 mSv/week to 2.81E-02 mSv/week, which is still below the allowed dose rate.

In this paper, graphite calorimeter device which is using to calibrate the electron beam of therapy accelerators in the range of the low dose was simulated using the EGS4 calculation code. Then, the electron beam dose distributions were calculated in it and the effects of parameters such as thickness of the calorimeter core, the distance to the source, thickness of air gap around the core, thickness of graphite plates around the core, and the dimensions of the electron beam window, upon the amount of absorbed dose in the colorimeter core were investigated.

All building materials contain various amounts of natural radioactive nuclides. Materials derived from rock and soil contain mainly natural radionuclides of the uranium (238U) and thorium (232Th) series, and the radioactive isotope of potassium (40K). Radiation exposure due to building materials can be divided into external and internal exposure. The external exposure is caused by direct gamma radiation. The internal exposure is caused by the inhalation of radon (222Rn), thoron (220Rn) and their short-lived decay products. In this study, the external exposure caused by direct gamma radiation from granite samples as the building materials is considered. Total absorbed dose rate and annual effective dose rate in air due to under test granite samples were calculated. The obtained values for absorbed dose rate in all samples, except for sample No. 2 is more than the corresponding population-weighted world-averaged value of 60 nGy h−1. Applying dose criteria recommended by the European Commission Report on Radiological Protection Principles Concerning the Natural Radioactivity of Building Materials, 1999 for superficial materials, all investigated samples meet the exemption dose limit of 0.3 mSv y-1.

Lowness of the registered doses by the personal dosimeters (film badge in Iran) for most of dental radiology practitioners has made some of them doubtful about the rightness of the employed dosimetry method. On the other hand, some relatively considerable occupational exposures in dental radiology have been reported. These two opposite facts propose a comprehensive statistical and analytical research on the exposure levels in this practice. A thorough investigation on the available personal dosimetry data for dental radiology was carried out to define the occupational dose distribution of the practitioners. Furthermore, employing a supplementary assessment through questioners on the cases exceeding the investigation dose level, validity of such cases was verified. The results showed that more than 90% of the dental radiology practitioners receive doses near to the background level and 99% of them receive doses less that the annual dose limit for the public (1mSv/y). The investigations showed that only few cases have been registered doses over the investigation level, out of which, only one case has been the real exposure of the person to the ionizing radiation. In sum, occupational exposure in dental radiology is negligible and needlessness of the practice to personal dosimeters can be concluded.

Radon concentration was measured in some dwellings in Kashan city. List of buildings was chosen randomly including old buildings made up of mud brick & clay and new buildings constructed with bricks and cement. Population weighted Radon concentration survey was followed and measurements were carried out using polycarbonate (lexan) track detector at the levels of basement to second floor of the buildings. Obtained results show the maximum average radon concentration level in winter to be 127 Bqm-3, while in autumn is 108 Bqm-3, in spring is 96 Bqm-3and the minimum average radon concentration was measured in summer to be 85 Bqm-3. Also the analysis of the measured data show that in %95 of the measured dwellings the radon concentration level is bellow 300 Bqm-3, the reference level proposed by ICRP in 2009.