Radiation Physics and Engineering
Volume:4 Issue: 2, Spring 2023

The study is aimed at measuring the outdoor background ionizing radiation (BIR), the absorbed dose rate (ADR), the annual effective dose (AED) and excessive lifetime cancer risk (ELCR) at four sites in the Aniocha South local government area (LGA) of Delta State, denoted as A-D. The study was performed using a calibrated Geiger-Muller (GM) detector (Radiation Alert Inspector) as well as a geographic positioning system (GPS) to determine the longitude and latitude of each site. The average (range) outdoor BIR, ADR, and AED were 0.021±0.01 (0.01-0.04) mR/hr, 181.6±77.7 (60.9-322.8) nGy/hr, and 0.22±0.10 (0.07-0.40) mSv/yr, respectively. Among the processing sites, the average AED for granite, bitumen, and staff residential areas were 0.31, 0.12, and 0.17 mSv/yr, while surface measurements at the "burnt stone" had the highest AED (0.41 mSv/yr). ADR and AED were both considerably higher than the world average of 59 nGy/hr and 0.07 mSv/yr. The average effective lifetime cancer risk (ELCR) (0.77× 10-3) was higher compared to the world average of (0.25× 10-3), with the highest in the granites. The ELCR risk band indicated a concern for increased cancer risk. Educating the public about actions to reduce their exposure to environmental carcinogens is necessary.

Neutron data and cross-sections are highly regarded and are essential for developing nuclear equipment such as advanced fission and fusion reactors, accelerators, neutron shielding, physics studies, etc. The neutron cross-section should preferably be measured using a single-energy neutron beam, although the presence of a background in research reactors can affect its accurate determination. The Neutron Powder Diffraction (NPD) facility of Tehran Research Reactor (TRR) has been taken into consideration for measuring the neutron cross-section based on its properties, including neutron monochromator and multiple collimators. In this work, radiative capture cross-sections of Au, In, and Rh materials have been calculated using TRR monochromatic beam. MCNPX is a Monte Carlo particle transport code that has been applied to simulate the measurement system of the neutron cross-section and calculate the reaction rates. The effect of the presence and absence of different sections of the background on the cross-section values was investigated and the results were compared with EXFOR data library for validation. According to the findings, neutron backgrounds can have varying impacts depending on factors such as sample material, the isotope resonance regions, neutron source spatial distribution, and neutron monochromatic energy. However, the presence of fast neutron background contributes to the most uncertainty in the cross section values while its removal produces an average discrepancy from experimental libraries of 7.16%. Also, removing the cold neutron background also causes a relative difference equal to 7.65%.

Low energy I-125- seeds are considered as a common source in different brachytherapy techniques for treatment of different cancers. In this study, at first, we simulated and validated I-125 (model 6711) seed according to the TG-43U1 recommendation, by GEANT4 Monte Carlo toolkit. Moreover, we simulated new seeds containing cylindrical Ag+Al2O3 markers with different ratio of Ag and Al2O3 in the final composition of the marker and compared the radial dose functions and anisotropy functions of the sources. For validation and evaluation purposes, the radial dose function and anisotropy function were calculated at various distances from the center of the different simulated sources. The source validation results show that GEANT4 Monte Carlo toolkit produces accurate results for dosimetric parameters of the I-125 seed by choosing the appropriate physics list. On the other hand, results show a similarity between calculated dosimetric parameters of the I-125 seed (6711) and other sources, with a percentage difference of about 5%.

Simulation work provides valuable information on the behavior of different research reactor neutron analysis facilities. The present study considered neutron and secondary-gamma dose rate variations by applying a sapphire crystal inside the D channel in Tehran Research Reactor (TRR). The MCNPX computational code was used to model the channel and its designed shield. Neutron and gamma dose rates distributions were calculated with a sapphire crystal modeling to investigate the neutron diffraction facility hall dose rates. The data from the dose rate simulations were compared with the experimental data available at a power of 4.2 MW from the research reactor. The comparison showed that there is very good conformity between two data series. The simulated neutron dose rate in front of the main shield overestimated the measurement data by 57% in closed-shutter situation and underestimated the measured data by 32% in open-shutter measurement situation. The investigation has shown that adjusting the crystal size to the channel size is considerably effective, especially at high leakage positions.

Design, construction, and experimental investigation of the plasma water activation device have been presented in this article. In this design, one of the electrodes, which is plate ss316, is placed in water. The other electrode which is made from tungsten is placed inside a glass tube and immersed in water. Air is also blown into the water through a constant rate air pump of 5 L.min-1. An AC power supply with voltage and current of 15 kV and 30 mA has been used to create plasma in water. The results of the analysis of nitrite, nitrate, and pH in three water samples that have been irradiated with plasma for 10, 20, and 30 minutes showed a very significant change compared to the control sample. The pH of PAW is drastically decreased with an increase in treatment time due to the formation of strong acids. Nitrite and nitrate concentrations of PAW are increased with an increase in treatment time.

A numerical model was developed and a system of the nonlinear equations of deuterium-tritium burn-up in inertial confinement fusion have been solved to find the minimum conditions which are required for the formation of hot spot and starting the thermonuclear reactions in a self-sustaining mode. The effect of all the dominant phenomena in the nonequilibrium plasma, including the alpha particle energy deposition in the hot spot and transferring to ions and electrons, ions-electron coupling energy, and the main photons-matter interactions, which includes the bremsstrahlung radiation and the Compton scattering, were investigated. By using the Klein-Nishina equation for scattering cross-section of high energy photons, the effects of the photon-matter interactions from a relativistic point of view have also been studied. It was shown that the change of photon distribution shape can have a significant effect on the photon temperature, the photon-electron coupling energy and as a result on the electrons and the ions’ temperature in a diluted plasma.

During the operation of Graphite -fuel HTGR (High-Temperature Gas-cooled Reactor) nuclear reactors, Graphite  used as a neutron moderator, is irradiated and has a variety of contaminants (such as Cs-137, Co-60, and Sr-90) and due to industrial and environmental considerations, decontamination of irradiated Graphite  is very important. In this study, the decontamination of Cs-137 trapped in Graphite  pores of Graphite -fuel (HTGR) nuclear reactors has been analyzed. The proposed method for decontamination of irradiated Graphite  surfaces is the thermal plasma-sputtering method with noble feed gases, which are used to reduce the risk of radioactive Graphite  waste and in this regard, a mathematical model was developed to describe the process of decontamination of irradiated Graphite, which is prone to release Wigner energy due to defects and torsion caused by radiation. The results show that the decrease in radiation pollution of irradiated Graphite  waste and various parameters of its decontamination process depend on the release of Wigner energy. The results obtained are in good agreement with the other researchers results.

The specific activity of radionuclides in the soil of the Borujerd region using high purity Germanium detector (HPGe) was measured and the associated radiological hazards were calculated. The mean specific activity of radionuclides of Ra-226, Th-232, K-40, and Cs-137 in soil was obtained at 10.99±5.11, 35.36±4.44, 324.20±10.24, and 2.93±0.60 Bq.kg-1. These values were below the global average. Also, the value of basic radiological risk parameters, such as Raeq, AEDout, AEDint, Hex, Hin, and Iγ, ranged from 52.02 to 139.54 in Bq.kg-1, from 24.98 to 68.27 and from 42.90 to 117.22 in mSv.y-1, 122.57 to 334.93, 0.14 to 0.37, 0.16 to 0.40, and 0.27 to 1.04, respectively. The range of excess lifetime cancer risk (ELCR) value for the surrounding soil samples varied from 0.15×10-3 to 0.41×10-3, in which samples S4, S14, S24, S27, S28, S29, and S30 exceeded the global average of 0.29×10-3. A radiological map of the city of Borujerd was prepared using the GIS program. The study showed that the level of radioactivity in the Borujerd area did not exceed the critical value and is in line with the global results.