Radiation Physics and Engineering
Volume:4 Issue: 3, Summer 2022

Digital spectroscopy with Silicon Photomultiplier (SiPM) array coupled on CsI(Tl) crystal, needs some consideration to achieve desirable energy resolution. an array of SiPMs must be used with large scintillation crystal, therewithal signal from whole SiPMs is not ideal. Also silicon pixels are arranged in arrays at regular intervals. The distances between the pixels and the interference of the cross-talk of adjacent pixels are undesirable factors for energy resolution when the light is received by the crystal and transmitted to the SiPM array. On the other hand, due to the advantages of the SiPM array, there is a need to improve the energy resolution in order to be able to be applied. In this paper, an attempt is made to obtain the desired energy resolution by using the digital spectroscopy method and digital filters and properly shaping the output pulse of the preamplifier by the trapezoidal filter method. In this case, it will be possible to use it in different applications such as spectroscopy by the detector.

Primary standardization of radioactivity is related to the direct measurement of activity in radioactive decay. A large variety of primary standardization techniques have been developed in the past years. The photon-photon coincidence counting is one of the methods for activity determination. This method is particularly applied for the standardization of I-125 using the detection of X-ray and gamma-ray coincident counting. In this paper, a 2D photon-photon coincidence digital system with two similar ‎‎2'' × 2''‎‎ NaI(Tl) detectors for absolute activity measurement is developed. The system is established based on a 100 MHz CAEN waveform digitizer (DT5724) which directly records the pre-amplifier output signals of the two NaI(Tl) detectors. The sampled signals was transformed to trapezoidal signals using pulse height analyzer firmware and coincidence events were recorded in a list file. The list file was analyzed offline using a Matlab code to realize correlated gama lines of Co-60 source.  The Volkovitsky formulas were used for the activity calculation and the details of the experimental setup were also discussed. Standardization of  the two Co-60 standard sources was performed using this system. Results are in good agreement with the reference activity of Co-60 sources. The presented formula can be modified for absolute calibration of the other medical radioisotopes. The technique can be generalized for absolute activity measurement of I-125 which uses for ophthalmic plaque radiation therapy.

Neutron detection techniques are widely studied in many articles. Most of this research requires a lot of electronic equipment. In this study, using the Thick Gas electron multiplier (THGEM) detector, a new method for neutron detection is proposed to reduce electronic equipment. In the neutron detection system, the converter material is used for converting neutrons to protons that are directed to the THGEM detector. By filling the detector space with noble gas and applying special voltage, THGEM enters to Self-Quenched Streamer (SQS) mode for protons detection. All these steps are examined by simulation, then the detection system is made and is examined in the laboratory. Finally, the simulation results and laboratory results are compared.    The results show that the 1 mm Plexiglas layer is suitable for converting neutrons to protons. The suitable distance between the converter layer and the THGEM detector is 3 cm. Also, the SQS mode happens in the most number of THGEM holes when the THGEM voltage is 980 volt. Investigating an approach to neutron detection by placing THGEM in SQS mode can be useful because, firstly, placing the THGEM detector in SQS mode simplifies electrical circuits and secondly, with this proposed detection system; it is possible to design detectors with different dimensions for neutrons.

In this work, the concentration of tritium in D2O of various degrees of purity was measured. Samples were taken from the Arak heavy water plant and tritium concentrations were determined using a liquid scintillation detector (LSC) based on tritium decay. In this work, instead of simple distillation, is used the azeotropic distillation method. Absorption and fluorescence spectra were recorded using a Shimadzu UV-2100 spectrometer and an LS50B fluorescence spectrometer. The tritium concentration in the samples varied from 1.75 ± 0.80 to 6.16 ± 1.01 Bq.L-1 in D2O enrichment from 0.35% to 77.50%. The correlation coefficient between tritium concentration and D2O purity in heavy water was obtained as R2 = 0.853. Deviation for 99.8% D2O enriched in heavy water. This was observed from a straight line, leading to a drop in R2. The results of this measurement showed that the tritium concentration did not exceed the value set by the Nuclear Regulatory Commission (NRC).

In this work, neutron and gamma shielding were simulated using MCNPX code for an inertial electrostatic confinement Fusion (IECF) device. In this regard, various properties of shields were investigated. Portland reinforced concrete was considered as the first layer. In addition to being effective in reducing the dosage of fast neutrons, concrete layer was also considerably effective in reducing the dose of gamma rays. As for the second and third layers, we opted for paraffin and boric acid based. These layers were chosen based on parameters such as lethargy, macroscopic slowing down power (MSDP), etc. in order to reduce the speed of epithermal neutrons and then absorb the thermal neutrons, thus reducing the transmitted neutron dosage as much as possible. A layer lead was used after these three layers of shielding to attenuate the gamma ray reaching this layer. In this study, a fusion source based on D-T fuel with homogeneous and isotropic radiation of neutrons was used and then dosimetry was performed for different parts. Afterwards, the thickness of the shielding layers was optimized in such a way that the neutron and gamma doses were reduced according to the standards. We found that it is possible to achieve safe neutron and gamma fluxes and doses by applying about 5 layers of 50 cm thickness. We compared the results of our study with the those of another study done on shielding for the IECF device, which were in good agreement.

Hot springs are known as one of the hydrotherapy centers in the world and have been welcomed due to their healing properties. Due to the presence of radon and radioactive elements in hot spring sediments, water and soil, these components are radioactive. So far, radiation hazards and the annual effective dose of hot spring components in the body organs have not been investigated in Iran. The purpose of this study was to calculate the amount of  U-238, Cs-137, Th-232, and K-40 elements in soil, water, and sediments of Jooshan hot springs in the Kerman region. The presence of these elements causes radiation hazards and an effective annual dose in people who use these hot springs. In addition to the healing properties of hot springs, the high amount of radiation hazard and effective annual dose may cause cancer risk. Experimental results with CsI(Tl) detector showed that the total activities of these elements in soil, water, and sediments of Jooshan hot spring were 95.26±9.76, 52.86±7.27, and 51.61±7.18 Bq.kg-1 respectively. The Jooshan hot spring's radiation hazards were calculated using activity measurement of the radioactive elements in soil, water, and sediments which was less than the permission level. The result of the Monte Carlo simulation with the MCNPX code showed that the effective annual dose of sediment, water, and radon in Jooshan hot spring are 5.43E-06, 3.00×10-3 and 1.16×10-1 mSv.year-1 respectively, which is less than effective annual dose (5 mSv.year-1). The maximum time for treatment by hot spring water is considered equal to one year.

In recent years, various designs for controlled thermonuclear fusion based on the p11B reaction have been reviewed and optimized. In this article, to innovate in achieving a better power and energy gain of neutron-free p11B fusion reaction, the improvement of the cross-section and also the kinetic effects. Then, the effects of bremsstrahlung radiation and ion and electron energy exchange rate have been evaluated by introducing relativistic effects and its role on improving fusion energy gain. As a result, the temperature of the electron is kept lower than that of the ion, which improves fuel performance. Finally, it leads to an increase in the number of protons at higher energies compared to the pure Maxwellian distribution and it causes a significant increase in reactivity compared to previous research. Also, the number of alpha particles obtained through calculations coincides with the latest research and leads to an enhancement of approximately 13%. This means that by improving the fusion cross-section of p11B, our calculations show that considering the avalanche effects, the range of achievable energy gain in the temperature range of 300 to 500 keV and the stable characteristic time of 0.64 ps reaches 89 to 111. While in the same temperature range and with the stable characteristic time of 0.74 ps, regardless of the improved cross-section, the energy gain range is 75 to 98.

Proton therapy of liver tumors can be challenging due to the absorbed dose of produced secondary particles in non-target organs. This study aims to evaluate the absorbed dose of secondary particles during the proton therapy of liver cancer through the MCNPX Monte Carlo (MC) code by a simplified MIRD-UF standard phantom. At first, a simplified MC model of MIRD-UF standard phantom was simulated using MCNPX. After the proper proton energies calculation ranging from 90 to 120 MeV for 4×4×4 cm3 tumor irradiation, mesh tally type 3 and F6 tally were used to calculate the depth dose profiles as well as the absorbed dose of protons and secondary particles in non-involved organs. The obtained results illustrated that the fluence of internal secondary particles doses was considerably small in comparison with primary protons. Furthermore, most of neutrons and photons doses were absorbed around the liver tissue for all performed proton energies (i.e., 90 and 120 MeV) which non-target organs did not receive a significant high dose. Furthermore, the absorbed dose of secondary photons and neutrons had slight variations in considered normal tissues near the liver. The calculated results in this study indicated that during the proton therapy of liver cancer, the most contribution of the secondary particle doses was absorbed inside the liver tissue. Hence, it can be expected the probable side effects (secondary cancers) associated with the liver cancer proton therapy may be decreased however, the presence of secondary particles should not be ignored.