نشریه سنجش و ایمنی پرتو
سال هشتم شماره 6 (پیاپی 35، پاییز 1399)

When an ionizing beam enters a living cell, it interacts with the cellular material, thereby transferring some of its energy to it. In this study, the propagation of Auger electrons and their effects as DNA damage by some radionuclides, were analyzed using the Geant4-DNA Toolkit with the 1ZBB model. The 1ZBB model is selected from the protein data bank library which simulates the location of atoms in the sequence of cellular chromosomes. The average number of DSBs is shown as a function of energy. In this study, it is shown that the most damage is caused by Auger electrons with energies of less than 1 keV which corresponds to Auger electrons in layers M, N.

It is time-consuming and costly to build geometric shapes of neutron shields using machining techniques. In addition, in the case of large shields or the need for quick construction, common polymer shields such as boron-containing polyethylene are not efficient. In this study, Poly Methyl Methacrylate (PMMA) as a polymer base, boric acid powder as a neutron absorber and high-density polyethylene powder for increasing the amount of hydrogen have been using to build composite shielding by in situ polymerization method. The percentage of optimal composition of composite materials was calculated using simulation in the MCNPX code and the made shields in the layout including the 3He detector and the Am-Be neutron source were evaluated. By the results, the thermal neutron flux is decreased about 80% after composite shielding containing 67% PMMA, 28% Boric Acid and 5% High Density Polyethylene (HDPE) comparing with pure polyethylene shielding. The results showed a good performance of this composite shield.

Determination of X-ray spectrum at various nuclear radiology centers is important in terms of device quality control. Due to the pile-up effect and dead time, this wonchr('39')t be possible by simple methods, and by placing the detector in front of the X-ray beams. Therefore, to obtain the X-ray spectrum, one can use indirect methods like spectroscopy Compton photons scattered at 90° angle. The primary spectrum is related to the Compton spectrum matrix by the response matrix in an equation system. As the coefficients matrix of this equation is ill-conditioned, MTSVD regularization mathematical method was used in this study for spectrum unfolding the X-ray spectra. Finally, the accuracy of the spectrometry was obtained by comparing the unfolded spectra with the IPEM78 theory spectrum. To compare the closeness of the unfolded spectra with the theory, the root-mean-square deviation was obtained, and the percentage error was used for comparison of the half value layer. According to the results obtained in this study, the indirect Compton method can unfold the X-ray spectrum by a good accuracy, i.e., less than 5%, and the relative error for obtaining the HVL, was found to be 2%.

In this work, the damage produced in Carbon and Silicon by fast neutrons in the energy range of 0.15 to 10 MeV was calculated using TRIM code. In this range of energy, the dominant interaction is the elastic scattering. The information about recoil, nuclei energy, scattering angle and the depth in which the interaction was occurred is necessary to run TRIM code. For calculating the energy and scattering angle of recoil nuclei, the spectrum of scattered nuclei should be known. The spectrum of scattered Silicon and Carbon was calculated by using an analytical method to relate the statistical distributions to each other. Comparison of the calculated scattered spectrum with experimental one confirmed the validity of the calculation method. Calculations revealed that the damage produced in silicon is higher than the one produced in the Carbon. This is because the displacement energy threshold in Carbon is higher than the Silicon. So, using of Carbon in neutron exposure is recommended.

In this paper, boron nitride (BN) nanofibers have been produced and tested to absorb thermal neutron radiation in order to use in hospital gown raw materials. Samples were prepared for boron with 7%, 10%, 12% and 15% weight percent using electrospinning method. The samples were tested by exposure to a thermal neutron source with an indium detector. The total mass cross-section of the samples as well as the cadmium sheet was measured under the same conditions for the thermal neutrons, and the results were compared. The outcomes show that increasing boron content in the samples directly reduces the passing neutrons, which indicates the effect of boron nitride content and uniform distribution of boron nitride particles in the produced nanofibers. Results also revealed that the total mass cross-section of the samples is much higher compared to cadmium. The produced samples are very light with a fabric-like structure which makes it perfectly suitable to produce neutron radiation shields.

In this work, the effect of the gamma radiation on the threshold voltage of metal oxide field effect transistor (MOSFET) was investigated. First, the purchased P typeMOS transistor was considered. Due to a nonlinear response to the administrated dose, the different regions were analyzed. Then we constructed an Al/n-Si/SiO2/Al MOS diode. These diodes were irradiated to mili to kilo Gray dose and their electrical behavior were investigated. In forward bias, the current variation was low but in reverse bias and lower than -20 volt, the current variation was large and measurable. A non-constant responsivity was obtained for diodes which is due to the non linearity of dose - current curve. Three different regions were chosen and the responsivity in these regions was obtained.

Holmium-166 is one of the most effective radionuclides used for treatment of bone marrow cancer and rheumatoid arthritis. Using 166Dy/166Ho in vivo generator to deliver 166Ho to the target tissue causes the minimum exposure to the non-target tissues and increased absorbed dose rate in the target tissue compared to other similar radiopharmaceuticals. In this work, theoretical calculations of 166Dy/166Ho in vivo generator production in Tehran Research Reactor were performed through solving the decay chain differential equations simultaneously for natural and enriched dysprosium targets and the obtained results were compared with experimental values. The results showed that there is a good compatibility between the measured values and the results of the theoretical calculations. The outcomes of this study are used to produce 166Dy / 166Ho radionuclide generator.

The gamma emission tomography system is one of the most common medical imaging systems that have been used to monitor the integrity and defect of nuclear fuel in recent years. In this research, the capability of gamma emission tomography system for partial defect verification of irradiated fuel assemblies is investigated. For this reason, a typical gamma emission tomography system is simulated. This system is capable for obtaining cross-sectional images of the irradiated fuel assembly in the spent fuel pool and investigates the fuel partial defect by processing them. It is noted that the gamma attenuation correction is considered in this study. Also, the capability of gamma emission tomography system for calculating the relative activity distribution of fuel rods in a nuclear fuel assembly is investigated by a developed algorithm.