Mehrnaz Zehtabvar; Dariush Sardari; Gholamreza Jahanfarnia
Abstract
The concentration of naturally occurring radioactive materials (NORM) in surface water and irrigation wells is measured using gamma ray spectrometry by HPGe detector. Measurement was carried out for samples that were collected over seventeen points in Tehran city and its suburbs. ...
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The concentration of naturally occurring radioactive materials (NORM) in surface water and irrigation wells is measured using gamma ray spectrometry by HPGe detector. Measurement was carried out for samples that were collected over seventeen points in Tehran city and its suburbs. The samples were prepared in compliance with the principles from irrigation wells of city. The specific radioactivity of Ra-226, Th-232 and K-40 were measured and the results from different locations covered a range with the minimum being below "minimum detectable activity" up to maximum of 4.04, and 6.85 and 4.7 Bq per liter of water, respectively. The accumulation of radioactive materials in the samples from southern areas of Tehran was more than that of central areas. Also, concentration of Ra-226 in all the samples was less than the Derived Release Limit of Canada and Environmental Protection Agency standard threshold.
Seyed Milad Miremad; Babak Shirani Bidabadi
Abstract
In this paper, the effect of anode's insert material on spatial distribution of X-ray emission zone of plasma focus device was studied. Anode's insert materials were fabricated out of aluminum, zinc, tin, tungsten and lead. For each insert ...
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In this paper, the effect of anode's insert material on spatial distribution of X-ray emission zone of plasma focus device was studied. Anode's insert materials were fabricated out of aluminum, zinc, tin, tungsten and lead. For each insert material at the constant operating voltage of 21 kV, the image of pinhole camera which monitors the surface and the top of anode was recorded at the various pressures of 0.3, 0.6, 0.9 and 1.2 mbar. The results indicated that the X-ray emission zone above the anode surface not only includes thermal radiation of plasma, but also depends on anode's insert materials. This zone could be due to the passage of high energy electrons from the vapor of anode's material above the anode's surface.
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.
Mahdi Ghaed Rahmati; Mostafa hasanzadeh; Seyed Amir Hossein Feghhi
Abstract
Kinetic and neutronic parameters play an important role in analysis of reactors dynamic behavior. Some of these parameters include: effective multiplication factor (keff), reactivity (ρ), neutron flux as well as power spatial distributions, effective delayed neutron fraction (βeff) and prompt ...
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Kinetic and neutronic parameters play an important role in analysis of reactors dynamic behavior. Some of these parameters include: effective multiplication factor (keff), reactivity (ρ), neutron flux as well as power spatial distributions, effective delayed neutron fraction (βeff) and prompt neutron lifetime (lp ). In this work, Monte Carlo modeling and analysis of Isfahan MNSR is performed for calculation of the kinetic and neutronic parameters of using MCNPX2.6 code, slope fit and perturbation methods. Relative differences between results of the MCNPX2.6 code in calculation of the ρ and βeff and the reference values are about 0.5% and 2.1%, respectively. The relative differences between the results of the slope fit and perturbation methods and MCNPX2.6 code in calculation of the parameter with the reference values are about 17.6%, 4.8% and 29.19%, respectively. Therefore, the results of these research show that the MCNPX2.6 code is suitable for calculation of the reactor kinetic parameters such as the βeff, while the perturbation method is a simple and convenient method for calculating the .
Zohreh Gholamzadeh; Rohollah Adeli; Mahdi Keivani
Abstract
Routine gamma dosimetry of spent fuels in nuclear power stations is mandatory to manage their storage in dry or wet spent fuel storages. Mostly the spent fuel gamma dose rate measurements out of the spent fuel pool is impossible because of the high exposures of the operators. Therefore, ...
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Routine gamma dosimetry of spent fuels in nuclear power stations is mandatory to manage their storage in dry or wet spent fuel storages. Mostly the spent fuel gamma dose rate measurements out of the spent fuel pool is impossible because of the high exposures of the operators. Therefore, determination of a conversion factor as precise as possible is important that could be applied to convert the measured gamma dose rate inside the water shield to the air values. Simulation methods are powerfully applied to investigate the conversion factor variation trends due to different burnup, cooling time and irradiation history of the spent fuels. The present work uses MCNPX Monte Carlo-based code to determine the trend. The obtained results of this computational study showed that the conversion factor would not have any dependency to the cooling times, burnup values and irradiation history if the detector is placed at special positions in air or water environments. Comparison of the simulation and experimental data showed an acceptable conformity, so that the experimental verified the simulation data trend
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.
Parviz Ashtari
Abstract
18F-FDG PET/CT is commonly used for evaluation and diagnostic of many types of cancer, such as; tumor diagnosis, treatment monitoring, and radiation therapy planning. Accurate diagnostic is needed in meticulous patient preparation, including restrictions ...
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18F-FDG PET/CT is commonly used for evaluation and diagnostic of many types of cancer, such as; tumor diagnosis, treatment monitoring, and radiation therapy planning. Accurate diagnostic is needed in meticulous patient preparation, including restrictions of diet and activity and management of blood glucose levels in diabetic patients, as well as an awareness of the effect of medications and environmental conditions. All of these conditions play important roles toward obtaining good-quality images, which are essential for accurate interpretation. This article introduces the new synthesis and quality control method for obtaining the best quality FDG which is used as radiopharmaceutical. All the reactions are carried out and completed in one reaction vessel without any replacement. The paper is including details of synthesis, quality control and transportation step. It is the first time that the alkaline FDG synthesis is introducing by details in Iran.
Zahra Shahbazi Rad; Fereydoun Abbasi Davani; Gholamreza Etaati; Samaneh Seifi
Abstract
The aim of this research was determination of the required time for coagulation of in vivo cut bleeding treated by non-thermal atmospheric pressure plasma. To meet this, an atmospheric pressure plasma jet device was designed and constructed. Helium was used as working ...
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The aim of this research was determination of the required time for coagulation of in vivo cut bleeding treated by non-thermal atmospheric pressure plasma. To meet this, an atmospheric pressure plasma jet device was designed and constructed. Helium was used as working gas. The electrical parameters and optical emission spectrum of helium plasma were measured. The averaged treatment time to coagulate the incision bleeding on the mouse liver was obtained 8.6 μs, and the average time of naturally incision bleeding coagulation was 10 min.
Mohammad Javad Safari
Abstract
It is well-known that response function of organic scintillation detectors does not appear with photopeaks. Instead, their dominant feature is a continuum, usually called the Compton edge that innately exposes the resolution characteristics of detection system. ...
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It is well-known that response function of organic scintillation detectors does not appear with photopeaks. Instead, their dominant feature is a continuum, usually called the Compton edge that innately exposes the resolution characteristics of detection system. While, accurate characterization of Compton edge is crucial for calibration purposes, it is also in charge of elaborating the energy resolution of detector. This paper presents a simple method for accurate characterization of the Compton edge in organic scintillation detectors. The method is based on the fact that differentiating the response function leads to accurate estimation of the constituting functions. The differentiation method, in addition to the location of the Compton edge, gives insights into the parameters of the folded Gaussian function which could lead to depict the energy resolution. Moreover, it is observed that the uncorrelated noise in the measurement of the response function does not impose significant uncertainties in the evaluations, so it could preserve its functionality even in lower-quality measurements. By simulation of the bounded electrons and considering the Doppler effects, we are able to demonstrate -the first ever- estimation for intrinsic Doppler resolution of an organic plastic scintillator. Even though, this possibility is an immediate result of benefiting the presented method for analysis of the Compton continua.
Ali Taaghibi Khotbesara; Faezeh Rahmani; Farshad Ghasemi
Abstract
This work presents an alternative method for Mo-99 production as a parent nuclide of Tc-99m which is the most used radioisotope in diagnostic imaging processes. Regarding to some benefits of accelerator-based methods over reactor-based methods for Mo-99 production, the electron ...
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This work presents an alternative method for Mo-99 production as a parent nuclide of Tc-99m which is the most used radioisotope in diagnostic imaging processes. Regarding to some benefits of accelerator-based methods over reactor-based methods for Mo-99 production, the electron Linac-based method has been selected. In this way of production, two approaches (one-stage and two-stage) are available using photoneutron reaction in Mo-100 target using bremsstrahlung photons. The superiority of one-stage approach and optimal dimension of target has been demonstrated by nuclear simulation using MCNPX2.6 code. Thermal analysis of the optimized target has been performed by COMSOL software, which has been led to select the indirect cooling system. The final suggested conceptual design of the target includes nine Mo-100 stripe plates with 0.2, 3, and 30 cm in thickness, width and length, respectively which being surrounded by two copper clamps as the cooling ducts. The velocity of 2.5 m/s of inlet coolant (water) is sufficient for the suggested cooling system to satisfy the conditions of the turbulent regime as the desired cooling regime.
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%.
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.
Payam Rafiepour; Shahab Sheibani; Daryiush Rezaey Uchbelagh; Hossein Poorbaygi
Abstract
Utilizing radioactive stents is a usual method for treatment of advanced esophageal cancer. It is necessary to investigate the dose distribution of radioactive esophageal stents before the clinical use. This study presents a dosimetric comparison between three radioactive esophageal ...
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Utilizing radioactive stents is a usual method for treatment of advanced esophageal cancer. It is necessary to investigate the dose distribution of radioactive esophageal stents before the clinical use. This study presents a dosimetric comparison between three radioactive esophageal stents: I-125 seed-loaded stent, iodine-eluting stent and double-layered iodine-eluting stent. Depth-dose and angular dose distributions were carried out using Geant4 toolkit. Moreover, the effect of interval distance between two adjacent seeds on the dose distribution was investigated. Esophageal stents loaded with I-125 seeds seems to be better than iodine-eluting stents, with the distance less than 15 mm between two adjacent seeds.
Hossein Sharifian; Mahdi Aghaie; Ahmad Zolfaghari
Abstract
In this work, dynamic responses of a WWER-1000 reactor in reactivity insertions are studied using a coupling method. The ANSYS-CFX is implemented for thermal hydraulic study of the core and the point kinetic equation (PKE) is coupled as a FORTRAN subroutine. For transient ...
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In this work, dynamic responses of a WWER-1000 reactor in reactivity insertions are studied using a coupling method. The ANSYS-CFX is implemented for thermal hydraulic study of the core and the point kinetic equation (PKE) is coupled as a FORTRAN subroutine. For transient analysis of the core, the thermal feedback of the fuel is added to coolant, and numerical solver of cylindrical heat transfer for obtaining the irradiated fuel rod temperature profile is also included. In order to investigate the irradiation effect, the fuel and gap properties in burnup with appropriate correlations could be calculated. Using memory management system (MMS) and data transfer arrays, coupling between numerical subroutines is carried out. It is shown that the dynamic response of the core depends on burnup, and the response could be varied in time. In addition, the coupling method is reliable for other dynamic calculations.
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.
Mohsen Salehi; Mohsen Shayesteh
Abstract
The analysis deals with the assessment of best estimate code RELAP5/SCDAP mod3.4 in the simulation of double-ended loss coolant accident as a LBOCA, 4 in break as a SBLOCA an SBO accident with considering except accumulator water where no core cooling water systems are available. ...
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The analysis deals with the assessment of best estimate code RELAP5/SCDAP mod3.4 in the simulation of double-ended loss coolant accident as a LBOCA, 4 in break as a SBLOCA an SBO accident with considering except accumulator water where no core cooling water systems are available. The reference plant is SURRY nuclear power plant as a Westinghouse three-loop nuclear power plant. In order to mitigation accident, the in-vessel retention strategy was investigated for the prevention of lower plenum failure. It has been concluded that during the SBLOCA, LBLOCA conditions bottom of active fuel is uncovered at 6340 s and 2160 s, respectively. It occurred for two times at 11650 s and 15608 s in SBO. At 6792 s and 57002 s in the LBLOCA and SBO due to reaching melting point and in the SBLOCA at 15215 s due to lower plenum creep rupture, failure of the reactor pressure vessel occurred. The results show that hydrogen production in the SBO is more than the other two cases. For the prevention of the lower plenum failure, the in-vessel molten material retention strategy is investigated as a passive system. The results show that lower plenum heat flux can be kept below the critical heat flux and its integrity is preserved in two cases of this analysis.
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%.
Farahnaz Saadatian Derakhshandeh
Abstract
Numerical solution of the multi-group static forward and adjoint neutron diffusion equation (NDE) using the Finite Elements Method (FEM) is investigated in detail. A finite element approach based on the generalized least squares method is applied for the spatial discretization of the NDE ...
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Numerical solution of the multi-group static forward and adjoint neutron diffusion equation (NDE) using the Finite Elements Method (FEM) is investigated in detail. A finite element approach based on the generalized least squares method is applied for the spatial discretization of the NDE in 3D-XYZ geometry. A computer code called GELES was also developed based on the described methodology covering linear or quadratic tetrahedral elements generated via the mesh generator for an arbitrary shaped system. A number of test cases are also studied to validate the proposed approach. Moreover, to assess the output dependency to the number of elements, a sensitivity analysis is carried out at the end.
Reza Ziyaee Sisakht; Fereydoun Abbasi Davani; Rouhollah Ghaderi
Abstract
Distinguishing naturally occurring radioactive (e.g. ceramics, fertilizers, etc.) from unauthorized materials (e.g. high enriched uranium, Pu-239, etc.) to reduce false alarms is a prominent characteristic of radiation monitoring port. By employing ...
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Distinguishing naturally occurring radioactive (e.g. ceramics, fertilizers, etc.) from unauthorized materials (e.g. high enriched uranium, Pu-239, etc.) to reduce false alarms is a prominent characteristic of radiation monitoring port. By employing the energy windowing method for the spectrum correspond to the simulation of a plastic scintillator detector using the MCNPX Monte Carlo code together with an artificial neural network, the present work proposes a method for distinguishing naturally occurring materials and K-40 from four unauthorized sources including high enriched uranium and Pu-239 (as special nuclear materials), Cs-137 (as an example of dirty bombs), and depleted uranium.
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.
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.
Afshin Hedayat
Abstract
Both of small and medium sized reactors and small modular reactors are called SMRs. They are reviewed and discussed in this paper, particularly integral Pressurized Water Reactors (iPWRs). Studies show that PWRs are the most interested, designed and constructed ...
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Both of small and medium sized reactors and small modular reactors are called SMRs. They are reviewed and discussed in this paper, particularly integral Pressurized Water Reactors (iPWRs). Studies show that PWRs are the most interested, designed and constructed nuclear reactor type worldwide. Some innovative small modular PWRs like the MASLWR, NuScale, CAREM-25, SMART and ACP-100 have several outstanding characteristics to be promisingly recognized as near term options of the next generation of small modular PWRs. They have several inherently safety features and improved passive safety system. They require smaller infrastructure and capital costs. They can be also developed rapidly in different and independent modular unites even for remote area or outlands without required infrastructure or electrical grids. It should be noted that new modern economy strategies like the Return of Investment (ROI) issues may advice medium or large reactors rather than small units for developed and industrial countries while small modular plans can be much more interesting and accessible for new comers or even developing countries. Finally, multi-applicability is an appropriate solution to develop expensive nuclear power plants economically as well as multi-purpose research reactors (especially by means of small modular iPWRs).
Reza Gharaei
Abstract
The role of saturation property of cold nuclear matter is examined in order to describe the steep falloff phenomenon of the measured fusion cross sections at energies far below the Coulomb barrier for 58Ni+54Fe colliding system. For this aim, the ...
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The role of saturation property of cold nuclear matter is examined in order to describe the steep falloff phenomenon of the measured fusion cross sections at energies far below the Coulomb barrier for 58Ni+54Fe colliding system. For this aim, the double-folding microscopic approach which is modified by modeling the repulsive core effects in the nucleon-nucleon interactions is used to calculate the nuclear interaction potential. Moreover, the theoretical values of the fusion cross section, S factor, and the logarithmic derivative are computed using the coupled-channel technique, including couplings to the low-lying 2+ and 3- states in target and projectile. The results obtained reveal that the corrective effects of cold nuclear matter can be responsible for the description of the fusion hindrance phenomenon in our chosen system.
Peiman Rezaeian; Vahideh Ataenia; Sepideh Shafiei
Abstract
In this paper, dose uniformity ratio in irradiation cell of GC-220 is specifiedutilizing an analytical method based on the multipole moment expansion. In this method, the values of monople, dipole and quadrupole moments for source arrangements of GC-220 are ...
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In this paper, dose uniformity ratio in irradiation cell of GC-220 is specifiedutilizing an analytical method based on the multipole moment expansion. In this method, the values of monople, dipole and quadrupole moments for source arrangements of GC-220 are calculated by numerical integrating. Appling these values, the dose uniformity ratio in the irradiation cell of GC-220 is calculated equal to 1.92. Monte Carlo simulation is applied to validate calculations. There is a relative difference about 12% between the results obtained from the analytical calculation and Monte Carlo simulation, which confirm the used method. In comparison with Monte Carlo methods, this method is not time consuming, so, this method can be used for the conceptual designing and the source load planning of irradiators.
Malihe Omrani; Hossein Sadeghi; Samaneh Fazelpour
Abstract
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 ...
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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.
