Mohammadreza Abbasi
Abstract
The Jacobian-Free Newton-Krylov (JFNK) method has been widely used in solving nonlinear equations arising in many applications. In this paper, the JFNK solver is examined as an alternative to the traditional power iteration method for calculation of the fundamental eigenmode ...
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The Jacobian-Free Newton-Krylov (JFNK) method has been widely used in solving nonlinear equations arising in many applications. In this paper, the JFNK solver is examined as an alternative to the traditional power iteration method for calculation of the fundamental eigenmode in reactor analysis based on even-parity neutron transport theory. Since the Jacobian is not formed the only extra storage required is associated with the workspace of the Krylov solver used at every Newton step. A new nonlinear function is developed for the even-parity neutron transport equation utilized to solve the eigenvalue problem using the JFNK. This Newton-based method is compared with the standard iterative power method for a number of multi-groups, one and two dimensional neutron transport benchmarks. The results show that the proposed algorithm generally ends with fewer iterations and shorter run times than those of the traditional power method.
Amir Veiskarami; Mahdi Sadeghi; Dariush Sardari; Shahryar Malekie
Abstract
Collision of protons with background gas and beamline wall in proton therapy causes the creation of secondary particles, e.g. neutrons, which results in more difficulties in curing the tumors. In the present simulation-based study, the optimum diameter of ...
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Collision of protons with background gas and beamline wall in proton therapy causes the creation of secondary particles, e.g. neutrons, which results in more difficulties in curing the tumors. In the present simulation-based study, the optimum diameter of proton beamline was determined to minimize the production of secondary particles in the presence of electric field with the magnitude of 50 kV/m, perpendicular equal magnetic fields of 0.7 T, and background gas of argon under Bounce boundary conditions via finite element method. The results showed that the optimum diameter of the beamline for minimization of the secondary particles in the spot scanning proton therapy in the aforementioned conditions was 7 mm. Also, the values of drift velocities of protons were plotted in different time steps of 10 ns to 50 ns for the optimized size of the beamline. Due to few interactions of forwarding particles with background gas, the results showed that the forwarding particles in the propagation direction have greater velocities than those of rear particles. The results can be used in spot scanning proton therapy for curing the localized cancers.
Gholam Hossein Roshani; Alimohammad Karami; Ehsan Nazemi; Cesar Marques Salgado
Abstract
The used metering technique in this study is based on the dual energy (Am-241 and Cs-137) gamma ray attenuation. Two transmitted NaI detectors in the best orientation were used and four features were extracted and applied to the model. This paper highlights the application of ...
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The used metering technique in this study is based on the dual energy (Am-241 and Cs-137) gamma ray attenuation. Two transmitted NaI detectors in the best orientation were used and four features were extracted and applied to the model. This paper highlights the application of Adaptive Neuro-fuzzy Inference System (ANFIS) for identifying flow regimes and predicting volume fractions in gas-oil-water multiphase systems. In fact, the aim of the current study is to recognize the flow regimes based on dual energy broad-beam gamma-ray attenuation technique using ANFIS. In this study, ANFIS is used to classify the flow regimes (annular, stratified, and homogenous) and predict the value of volume fractions. To start modeling, sufficient data are gathered. Here, data are generated numerically using MCNPX code. In the next step, ANFIS must be trained. According to the modeling results, the proposed ANFIS can correctly recognize all the three different flow regimes, and other ANFIS networks can determine volume fractions with MRE of less than 2% according to the recognized regime, which shows that ANFIS can predict the results precisely.
Mohammad Amin Amirkhani; Mostafa Hassanzadeh; Safar Ali Safari
Abstract
Spallation process is the most significant process for neutron generation in industry and medicine. This process has been used in the subcritical reactor core. In this research, we study the neutronic behavior of non-fissionable and fissionable spallation targets consists ...
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Spallation process is the most significant process for neutron generation in industry and medicine. This process has been used in the subcritical reactor core. In this research, we study the neutronic behavior of non-fissionable and fissionable spallation targets consists of U-238, Th-232, Lead Bismuth Eutectic (LBE) and W-184 materials in cylindrical and conic shapes using MCNPX code. Neutronic parameters consist of spallation neutron yield, deposition energy, and angular spectrum of the neutron output. The gas production rate and residual mass spectrum were investigated. The results of this research indicate that the shape of the target must be selected based on target material and operational purposes. The number of neutrons per energy unit is stable at energies higher than 1 GeV, and the rate of change in neutron generation has been reduced after that. Furthermore, hydrogen is the principal factor in swelling of spallation target and consists of about 88% of gas production. It was found that a target of LBE provides the most favorite parameters for both neutronic and physical properties.
Fereshteh Gholami; Ehsan Alibeigi; Mojtaba Shamsaei-Zafarghandi; Ehsan Nazemi
Abstract
By the rapid development of imaging systems such as PET/CT for diagnosis of cancer, the protection of staff and public has become a main health concern. Due to serious and irreversible harms of ionization radiations, protection of all those who are exposed is the main ...
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By the rapid development of imaging systems such as PET/CT for diagnosis of cancer, the protection of staff and public has become a main health concern. 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. In this study, the current shielding status of the PET/CT installations in Tehran's Shariati hospital was investigated using the MCNPX Monte Carlo code to ensure that the dose limits for both the controlled and uncontrolled area are not violated. The proposed simulation method was benchmarked with a validated analytical method. Shariati hospital provides services to four patients every day, leading to a dose rate in the range of 2.6 × 10-6 to 9.35 × 10-3 mSv/week. The minimum dose rate in this range represents the value behind the door of the waiting room (public uncontrolled area), while the maximum in this range corresponds to the value behind the glass of the scanner room (operator controlled area). The simulation results for 8 patients/day in this center showed that the dose rate behind the wall of the injection room will increase from 4.88 ×10-6 mSv/week to 2.81 × 10-2 mSv/week, which is well below the recommended levels. This indicates that the present shielding is adequate for up to four more patients per day.
Zohreh Gholamzadeh; Mohadeseh Gholshanian; Seyed Mohammad Mirvakili
Abstract
Today thorium based fuels are being investigated as an alternative fuel technology. However, the majority of thorium fuel research studies are limited to reactor physics investigations, which leaves a gap for dose evaluation and shielding concerns of such spent fuels. The present work investigates thorium ...
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Today thorium based fuels are being investigated as an alternative fuel technology. However, the majority of thorium fuel research studies are limited to reactor physics investigations, which leaves a gap for dose evaluation and shielding concerns of such spent fuels. The present work investigates thorium oxide fuel assemblies in Tehran research reactor. The fuel gamma dose rates are calculated at different burnups and cooling times. A comparison between the reactor routine fuel and the thorium oxide fuel is conducted to reveal the thorium-based fuel application shielding challenges. The obtained results showed that inverse to U3O8-Al routine fuel the spent ThO2 gamma dose rates are completely dependent to the burnup values. In addition, for transporting the spent ThO2 fuel with the routine transport casks there is needed to be waited for the higher cooling times than U3O8-Al transportation time or construction of thicker transport casks is needed for transportation of the thorium-based spent fuels at shorter times.
