Particle Accelerators, Ion sources
Hamid Rahimpour; HamidReza Mirzaei; Masoomeh Yarmohammadi Satri; Zafar Riazi Mobaraki
Abstract
A continuous-wave solid-state-based power amplifier is designed and simulated in this paper to work as an RF injector into an ECR ion source chamber. Employing a solid-state radio frequency power amplifier, instead of microwave tubes, leads to having higher efficiency, lower price, compact size, and ...
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A continuous-wave solid-state-based power amplifier is designed and simulated in this paper to work as an RF injector into an ECR ion source chamber. Employing a solid-state radio frequency power amplifier, instead of microwave tubes, leads to having higher efficiency, lower price, compact size, and longer lifetime. Also, a modular design can be achieved for designing higher output power by repeating lower power sources and combining them. The proposed solid-state source can deliver more than 200-watt power to the ion chamber with a single high-power transistor. The selected Doherty high-power transistor is internally matched to 50 ohms and doesn’t need a bias sequence circuit. Two gain stages are applied to drive the high-power transistor. The designed RF source is simulated using the Advanced Design System (ADS) based on the measured scattering parameters of components. Simulations show an output power of more than 57 dBm with a tunable frequency bandwidth from 2.3 to 2.5 GHz.
Measurement of Radiation and Radioactivity
Milad Payandeh; Mahdi Sadeghi; Dylan Richeson; Somayeh Gholami
Abstract
It is important to have accurate information regarding the dose distribution for treatment planning and to accurately deposit that dose in the tissue surrounding the brachytherapy source. However, the practical measurement of dose distribution for various reasons is associated with several problems. ...
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It is important to have accurate information regarding the dose distribution for treatment planning and to accurately deposit that dose in the tissue surrounding the brachytherapy source. However, the practical measurement of dose distribution for various reasons is associated with several problems. In this study, 6711 I-125, Micro Selectron mHDR-v2r Ir-192, and Flexisource Co-60 sources were simulated using the MCNP5 Monte Carlo method. To simulate the sources, the exact geometric characteristics of each source, the material used in them, and the energy spectrum of each source were entered as input to the program, and finally, the dosimetric parameters including dose rate constant, radial dose function, and anisotropy function were calculated for considered seeds according to AAPM, TG-43 protocol recommendation. Results obtained for dosimetric parameters of dose rate constant, radial dose function, and anisotropy function for I-125, Ir-192, and Co-60 sources agreed with other studies. According to the good agreement obtained between the parameters of TG43 and other studies, now these datasets can be used as input in the treatment planning systems and to validate their calculations.
Radioanalytical Methods and Radiochemistry
Ali Biganeh; Banin Shakeri Jooybari
Abstract
Compton scattering events are the main source of error on the peak counting during the Neutron Activation Analysis (NAA). The Compton suppressor system in instrumental NAA reduces the detection limit of the technique and leads to a data with a higher degree of precision. In this paper, a two-dimensional ...
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Compton scattering events are the main source of error on the peak counting during the Neutron Activation Analysis (NAA). The Compton suppressor system in instrumental NAA reduces the detection limit of the technique and leads to a data with a higher degree of precision. In this paper, a two-dimensional pseudo coincidence Compton suppressor system is presented for the NAA technique. The system is established based on a CAEN digitizer which directly records the pre-amplifier output signals of the two HPGe detectors. The recorded events in the list mode file are analyzed offline by a Matlab code and the correlated photopeak events are realized. The performance of the system for Compton suppression is tested by measuring the gamma lines of Ba-133 and Cs-137 standard sources. The results show that the presented technique provides the peak to Compton ratio up to 104 and can be an alternative for conventional Compton suppressor systems.
Nuclear Reactor science and Technology
Abolfazl Shoghi; Seyed Ali Hosseini; Amir Saeed Shirani; Mehdi Zangian
Abstract
The pressurizer is a key equipment to ensure the safe operation of pressurized water reactor by maintaining the reactor coolant system pressure within allowed tolerances. Various pressure control systems (Pressurizer) are adopted in industrial applications to satisfy their characteristics. In accordance ...
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The pressurizer is a key equipment to ensure the safe operation of pressurized water reactor by maintaining the reactor coolant system pressure within allowed tolerances. Various pressure control systems (Pressurizer) are adopted in industrial applications to satisfy their characteristics. In accordance with the purpose of using nuclear facilities, Steam, Gas-Steam, and Gas Pressurizer (PRZ) have been used. In nuclear industry, the dynamic behavior of each PRZ is different. Peak pressure is one of the important parameters in choosing the type of PRZ. This study has been evaluated for the University of Wisconsin High-Pressure Critical Heat Flux (WHPCHF) facility as the base loop. Three PRZs are connected to the WHPCHF loop to evaluate their performance during the in-surge scenario. The Peak pressure of the three PRZs is evaluated during transients. The results showed that the use of the Non-condensable Gas (NCG) increases the peak pressure due to the lack of NCG condensation during transient conditions. The use of gas PRZ makes it possible to change the pressure quickly. Also, the pure gas PRZ has the highest peak pressure but has straightforward control logic. The gas PRZ is the best choice for small reactors and high-pressure test facilities.
Particle Accelerators, Ion sources
Amir Charkhi; Parisa Zaheri; Amjad Sazgar; Iman Dehghan
Abstract
Since the production of tellurium hexafluoride gas requires the design of a suitable reactor system, so the study of tellurium oxide fluorination kinetics is of great importance. For this purpose, a novel laboratory system was designed and constructed to study the fluorination reactions by the volumetric ...
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Since the production of tellurium hexafluoride gas requires the design of a suitable reactor system, so the study of tellurium oxide fluorination kinetics is of great importance. For this purpose, a novel laboratory system was designed and constructed to study the fluorination reactions by the volumetric method. Fluorine gas was injected into the reactor containing a tellurium oxide pellet, and the reaction was studied by following the changes in pressure of the gas phase using a pressure transmitter instrument. In this volumetric system, the kinetic parameters of the reaction between tellurium oxide pellet and fluorine gas have been derived for a pressure range of 137.9 and 181.2 kPa by monitoring the gas phase pressure. The reaction temperature was adjusted to 204±1 ◦C using a heater. The results showed that the fluorination reaction of tellurium oxide is a first-order reaction. The reaction rate constant is calculated to be 6.86 × 10-4 s-1.
Measurement of Radiation and Radioactivity
Akbar Abdi Saray; Hossein Zaki Dizaji; Mortaza Taheri Balanoji
Abstract
To monitor personal safety in the fields of biomedical and health physics, it is necessary to be aware of radiation doses to protect the health and safety of persons. Radiation protection quantities such as air kerma, ambient dose equivalent, and exposure dose rate are obtained by the measured spectrum ...
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To monitor personal safety in the fields of biomedical and health physics, it is necessary to be aware of radiation doses to protect the health and safety of persons. Radiation protection quantities such as air kerma, ambient dose equivalent, and exposure dose rate are obtained by the measured spectrum to determine energy-dependent conversion coefficients/factors. This study aims to obtain and compare an ambient dose equivalent to H∗(10) from the measured gamma-ray spectra by the NaI(Tl) scintillation detector using two various methods. The first method, which is based on the detector response function to find the conversion function, is called the G(E) method. The second method is subdividing the measured gamma-ray spectra into the multiple energy bins, and then obtaining the ambient dose equivalent by using conversion coefficient functions (ω(E)), which were determined by the conversion coefficients (ωi) of each energy bin for three energy intervals of ≤185 keV, 185 to 850 keV, and ≥850 keV. To calculate the detector response matrix and the conversion coefficients of each region of energy, the Monte Carlo simulation code was used for the quasi-mono energetic gamma radiation sources and the synthetic spectra. The results indicate that using the technique based on subdividing the measured spectrum into multiple energy bins helps to avoid the inverse detector response matrix dimension limitations that occur in the G(E) method and also have a lower error percentage in the dose quantity calculation. Consequently, NaI(Tl) scintillation detector has an excellent potential to replace the classical dose rate instruments, i.e. Geiger-Muller, for the early warning of environmental radiation monitoring.
Fusion Technology and Equipment
Jam yazdanpanah
Abstract
The domestic plasma modeling framework AZERAP is introduced and its capabilities in simulating the plasma based accelerators and intense beam-plasma interaction are discussed. The current first beta-release of AZERAP exploits the fully kinetic, electromagnetic relativistic PIC algorithm as its numerical ...
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The domestic plasma modeling framework AZERAP is introduced and its capabilities in simulating the plasma based accelerators and intense beam-plasma interaction are discussed. The current first beta-release of AZERAP exploits the fully kinetic, electromagnetic relativistic PIC algorithm as its numerical engine. It is implemented in the object oriented language C++ and utilizes the Message Passing Interface (MPI) for parallelization. The main idea behind the development of AZERAP has been establishing a software platform for virtual plasma laboratory for plasma based particle beam sources and high power electromagnetic generators. Achieving this goal has implied attaining high functionality in introducing the input problem, supporting abstraction of the field and plasma structures/modules, and supporting high flexibility for future developments. The present first beta-release of AZERAP paws the way toward these objectives. Moreover, it offers a very comfortable user experience with code compile, debugging, execution, data accusation and data animation, simulating plasma based accelerators.
Radiation Sources
Ali Adeli Ahmadabadi; Zahra Shahbazi rad; Fereidon Abbasi Davani; Behjat Ghasemi
Abstract
In this research, the effect of ions produced in deuterium plasma on Tungsten (W) and Aluminum (Al) plates has been investigated using a plasma focus device with the specifications of (C=10.4 μF, V=23 kV, E=2.75 kJ). The W samples used because it is one of the key elements in the Tokamak device. Because ...
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In this research, the effect of ions produced in deuterium plasma on Tungsten (W) and Aluminum (Al) plates has been investigated using a plasma focus device with the specifications of (C=10.4 μF, V=23 kV, E=2.75 kJ). The W samples used because it is one of the key elements in the Tokamak device. Because we wanted to put the W samples at the distance from the anode top with maximum plasma produced ions, we should find the optimum place. Due to the high cost of W samples, we used Al samples to find the optimal conditions. The samples were irradiated at 8 cm distance from the anode top with deuterium ions produced by a plasma focus device. The sample analyses were done by the SEM and EDX methods. The sample irradiation by deuterium plasma ions caused a lot of damages and bubble formation on the sample surfaces. The analyses showed the extent of surface damage and the number of ions deposited on the surface. The number of damages on the Al surface was much higher than W. Bubbles were formed on the surface were due to the impact of deuterium ions on the W and Al samples. Also, the deuterium ion energy was measured with a Faraday cup as about 50 keV.
Measurement of Radiation and Radioactivity
Vahidreza Babaei; Peiman Rezaeian; Sedigheh Kashian; Reza Pourimani; Azam Akhavan
Abstract
In this paper, the spectrophotometric properties of a colored Nickel-based solution complex (Nickel nitrate hexahydrate and Methyl Orange (MO)) were investigated as a stable chemical dosimeter for using in radiation processing of agricultural products. Its simple synthesis method as well as low cost ...
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In this paper, the spectrophotometric properties of a colored Nickel-based solution complex (Nickel nitrate hexahydrate and Methyl Orange (MO)) were investigated as a stable chemical dosimeter for using in radiation processing of agricultural products. Its simple synthesis method as well as low cost made it a suitable dosimeter for use in radiation processing. The variation of absorbed dose was applied to measure the absorbed dose. The maximum absorbance for the solution was observed at 460 nm. This solution was irradiated at three different concentrations of Ni(No3)2.6H2O and MO by Co-60 gamma-ray. Also the variation of the absorbance as a function of PH of the solution was investigated. The results showed the solution absorbance decreases with an increase in doses, and this solution can be used as a routine dosimeter and has a linear response in the 50 to 1500 Gy range with acceptable stability in environmental conditions up to 40 days before and after irradiation.
Experimental and Theoretical Nuclear Physics
Sepideh Shafiei; Mohammad Lamehi-Rachti
Abstract
In the present paper, the dechanneling and the energy loss of protons at the energy interval of 1400 to 2200 keV along the {100} and the {110} planar directions of Si were studied by the simulation of the measured channeling Rutherford back-scattering spectra based on the exponential dechanneling function ...
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In the present paper, the dechanneling and the energy loss of protons at the energy interval of 1400 to 2200 keV along the {100} and the {110} planar directions of Si were studied by the simulation of the measured channeling Rutherford back-scattering spectra based on the exponential dechanneling function with a parameter λ. This parameter is proportional to the dechanneling rate and represents the mean distance that ions travel along the channel before escaping from the channel. The Levenberg-Marquardt algorithm was used to set the best values of the channeling to random energy loss ratio, and the mean channeling distance. The experimental results are well reproduced by this simulation. The data analyzed in this energy range did not show any particular trend with regard to energy dependence of the parameters. The differences between both the planar channels in the Si crystal and their influence on the energy loss ratio and dechanneling of proton ions are described.
Nuclear Reactor science and Technology
Farrokh Khoshahval; Mohammad Rajaee; Nafiseh Tehrani
Abstract
Actinide concentration and activity analysis of the nuclides resulted from the burnup (depletion) process during nuclear reactor operation lifetime is an essential problem in reactor design. Inventory and the corresponding activities of the Tehran Research Reactor (TRR) are evaluated using different ...
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Actinide concentration and activity analysis of the nuclides resulted from the burnup (depletion) process during nuclear reactor operation lifetime is an essential problem in reactor design. Inventory and the corresponding activities of the Tehran Research Reactor (TRR) are evaluated using different methods and compared with each other. WIMS-CITATION, ORIGEN, and MCNP codes are used for plate type inventory calculations. The important actinides, fission products, and fissile inventory ratio of TRR have been calculated at different burnup steps. It is worth noting to mention that knowing the value of inventory helps us for safe handling of the spent fuels and to have a perfect design for transport cask of spent fuels. In this paper, the fuel isotope inventories were calculated for the first and 83rd core configuration of the Tehran Research Reactor, which is named “Core01” and “Core83” respectively. The calculations were first performed using WIMS-D5 and CITATION neutronic codes and then the results are compared with that of ORIGEN and MCNPX code. The total radioactivity of the TRR core at the end of the reactor core life (Core83) is estimated to be 6.47 × 105 Ci.
Nuclear Reactor science and Technology
Hamed Khodadadi; Kamyar Sabetghadam
Abstract
Probabilistic uncertainty and sensitivity analysis is frequently recommended for safety and reliability assessment of computer simulations. For this purpose, SUAP has been developed, and its latest version is capable of working on analysis results obtained using five well-known nuclear codes (i.e. FRAPCON, ...
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Probabilistic uncertainty and sensitivity analysis is frequently recommended for safety and reliability assessment of computer simulations. For this purpose, SUAP has been developed, and its latest version is capable of working on analysis results obtained using five well-known nuclear codes (i.e. FRAPCON, FRAPTRAN, FEMAXI, MCNP, and COBRA). SUAP provides support to properly quantify input uncertainties as to probability distributions and appropriate dependency functions. Using the Monte-Carlo sampling method, random combinations of different uncertain input parameters are generated and used to make input files for the corresponding code applied for the modeling. To quantify uncertainties, SUAP determines the variation range for each specific output parameter at any chosen time and/or location. Moreover, sensitivity analysis is accomplished based on the Spearman correlation. In this study, in order to evaluate SUAP applicability, UQ&SA for fuel performance modeling of VVER-1000 fuel rods using FRAPCON code has been accomplished. Acquired results exhibit the possible range of uncertainties in fuel centerline temperature, as well as the importance of different uncertain input parameters on that.
Measurement of Radiation and Radioactivity
Reza Pourimani; Monire Mohebian; Mobina Abdi
Abstract
Nuclear radiations are harmful to the human body. The main sources of nuclear radiation are the decay chains of U-238, U-235, and Th-232 and also some radionuclides as K-40, which are present in small amounts in the materials of the earth's crust, including plants, rocks, soil and water. Radioactive ...
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Nuclear radiations are harmful to the human body. The main sources of nuclear radiation are the decay chains of U-238, U-235, and Th-232 and also some radionuclides as K-40, which are present in small amounts in the materials of the earth's crust, including plants, rocks, soil and water. Radioactive substances are transferred to the human body in a variety of ways, including plant and animal products. Therefore, it is very important to determine the amount of radioactive substances in food products. In this research, seven samples of pistachios with different types were collected from Tehran markets in Iran. In this project, ultra-pure germanium spectroscopy system model GCD30195 was used. The specific activities of Ra-226, Th-232 and K-40 varied from <1.96 to 9.86, from 1.21 to 1.95, and from 317.22 to 382.80 Bq.kg-1. The artificial radionuclide of Cs-137 in all samples was lower than minimum detectable value (MDA). Calculations of the radiological impact showed that consumption of pistachios would endanger human health. The results of this study also showed that the amount of natural radionuclides in pistachio cores is higher than pistachio shells.
Radiation Sources
Blessing Okeoghene Ijabor; Akintayo Daniel Omojola; Augustine Onyema Nwabuoku; Funmilayo Ruth Omojola
Abstract
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 ...
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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.
Mahya Pazoki; Hamid Jafari; Zohreh Gholamzadeh
Abstract
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 ...
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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%.
Measurement of Radiation and Radioactivity
Payvand Taherparvar; Ali AziziGanjgah
Abstract
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 ...
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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%.
Medical Applications
Zohreh Gholamzadeh
Abstract
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 ...
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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.
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