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vosoughi--naser
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Neutron Noise Calculation using Nodal Expansion Method
, M.Sc. Thesis Sharif University of Technology ; Vosoughi, Naser (Supervisor) ; Hosseini, Abolfazl (Co-Advisor)
Abstract
The present M.Sc. thesis consists of two sections including the static calculation and neutron noise calculation in rectangular and hexagonal geometries. The multi-group, two dimensional neutron diffusion equations and corresponding adjoint equations are solved in the static calculation. The spatial discretization of equations is based on Average Current Nodal Expansion Method (ACNEM). Size of nodes is the same size of the fuel assemblies in modeling both of rectangular and hexagonal geometries. The results are benchmarked against the valid results for BIBLIS-2D and IAEA-2D benchmark problems. In the second section, neutron noise calculations are performed for two types of noise sources,...
Localization of a Postulated Noise in VVER-1000 Reactor Core Using Neutron Noise Analysis Methods
, M.Sc. Thesis Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
In this thesis, localization of a postulated noise from limited neutron detectors sparsely distributed throughout the core of a typical VVER-1000 reactor is investigated. For this purpose, developing a 2-D neutron noise simulator for hexagonal geometries based on the 2-group diffusion approximation, the reactor dynamic transfer function is calculated. The box-scheme finite difference method is first developed for hexagonal geometries, to be used for spatial discretisation of both 2-D 2-group static and noise diffusion equations. Using the discretised static equations, a 2-D 2-group static simulator (HEXDIF-2) is developed which its results are benchmarked against the well-known CITATION...
Unfolding of the Gamma-Ray Spectrum of the Scintillator Detectors Using the Neural Network Method
, M.Sc. Thesis Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
The analysis of gamma-ray spectra of low resolution detectors is difficult and sometimes impossible, as a result of photo peak’s overlapping. The usual methods of radiation spectra analysis based on fitting of peaks to mathematical curves or detecting of peaks with numerical calculation, are valid for high resolution detectors. However these methods are less successful for lower resolution detectors such as the common scintillators. The wide peaks in the spectrum maybe overlap and make it difficult to analysis. To solve this problem, we test here a method, based on the use of the artificial neural network. At first spectra of elements are converted to the patterns which are suitable for the...
, Ph.D. Dissertation Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
The present ph.D. thesis consists of three sections including the static calculation, neutron noise calculation and neutron noise source unfolding in VVER-1000 reactor core. The multi-group, two dimensional neutron diffusion equations and corresponding adjoint equations are solved in the static calculation. The spatial discretization of equations is based on Galerkin Finite Element Method (GFEM) using unstructured triangle elements generated by Gambit software. The static calculation is performed for both linear and quadratic approximations of shape function; baesd on which results are compared. Using power iteration method for the static calculation, the neutron and adjoint fluxes with the...
Calculation of Pulse Height Distribution by Deterministic and Monte Carlo Methods
, M.Sc. Thesis Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
Radiation detection techniques have important applications including prompt gamma neutron activation analysis (PGNAA), radiotherapy dose measurement, medical imaging and homeland security. Pulse height distribution is also a useful parameter in Radionuclide identification. Gamma ray spectroscopy analysis is a common technique in measuring of radiation field to identify the radioactive source.
Probabilistic methods are commonly used for numerical simulation of Pulse height distribution. In Monte Carlo methods, each history is individually tracked and details of interactions are available. Using the Monte Carlo methods for calculation of pulse height distribution leads to more real...
Probabilistic methods are commonly used for numerical simulation of Pulse height distribution. In Monte Carlo methods, each history is individually tracked and details of interactions are available. Using the Monte Carlo methods for calculation of pulse height distribution leads to more real...
Noise Characterization in VVER-1000 Reactor Core
, M.Sc. Thesis Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
Neutron fluctuations in both low and high power reactors provide important information about the system. The origins of these fluctuations are different in both regimes. In low power the branching process is the reason of these fluctuations. But the random nature of technological processes such as boiling in the reactors cooling and mechanical structure vibrating such as control rod and fuel rod created the neutron noise. Mathematical descriptions and applications of neutron noise for each mode are different. In common mode, fluctuations in the neutron detection system, determine some parameters and determin the start abnormality in system. In this study the spatial dependence of the cross...
Investigating the Propagation Noise in PWRs via Coupled Neutronic and Thermal-Hydraulic Noise Calculations
, Ph.D. Dissertation Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
In operating nuclear reactor core, fluctuations (deviations from normal operating conditions) are usually produced and propagated. These fluctuations can be due to control rod vibrations, inlet coolant temperature fluctuations, inlet coolant velocity fluctuations and so on. The induced neutron noise can be detected by in-core neutron detectors. Noise source identifications (such as the type, location and propagating velocity) as well as the calculation of the dynamical parameters (such as moderator temperature coefficient in PWRs and Decay Ratio in BWRs) are of the main applications of the neutron noise analysis in power reactors.
Investigating the propagation noise in PWRs (specifically...
Investigating the propagation noise in PWRs (specifically...
Source Localization by Analysis the Response of Detectors Using Inverse Methods
,
M.Sc. Thesis
Sharif University of Technology
;
Vosoughi, Naser
(Supervisor)
Abstract
Localization of a neutron point source using a designed computer program namely “MCMC-MATURE” is performed. The computer program analyses several detector responses in some certain media by Markov Chain Monte Carlo (MCMC) method and a new iteration algorithm. Identification of the possible regions of source position would be found by analyzing the initial fluxes generated by mesh tally of MCNPX computer code. The designed computer program is capable to generate the flux between detectors. “Regular-Sequential”, “Irregular-Sequential” and “Non-Sequential” are three methods used for sampling the generated random number in two dimensions. Each sample multiplied by a sampling function and lead to...
Spatiotemporal Modeling & Simulation of the Second Order Moments of the Transport Equation
, Ph.D. Dissertation Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
Precise knowledge of the laws which govern a system is of interest for two reasons. First, it paves the way to understand the subtle behaviors of the system, which are not understandable from simpler models of the system. Second, it helps in the design of experimental requirements needed to observe these behaviors. The behavior of neutrons in a system, which could be a reactor or a detector, is stochastic from two perspectives. First of all, since the place of the atoms of a medium are randomly sited, at least from a neutrons point of view, the collision sites are random places, which contributes to the stochasticity of the transport phenomena. This type of randomness is somewhat similar to...
Investigation of Direct Discrete Method (DDM) for Transport Equation Solution in two Dimensional & Two-Dimensional Generalized Geometery
, M.Sc. Thesis Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
The emergence of complex equations in engineering sciences leads to development of numerical methods for solving these equations. Obviously, each of the proposed Algorithms and various methods in this field were evaluated according to the limitations of time, software and hardware and based on these limitations the advantages and disadvantages of these methods were determined. Direct Discrete method is one of the modern methods used in solving the conservation equations. In this method, in addition to the advantages of numerical methods, equations will be produced in discrete space to avoid involving with differential equations. This method was first studied in the field equations. The...
Solving the Neutron Transport Equation Using Unstructured Spatio-temporal Elements by the Direct Discrete Method (DDM)
, M.Sc. Thesis Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
Nowadays, the use of numerical methods is very common in solving complex equations, and various methods have been developed in this field. One of the new methods in this field, is the direct discrete method(DDM), Which was initially used to solve the electromagnetic field equations. In the past years, this method has been used to discretization some of the neutronics equations and acceptable results are recorded. So that the convergence order for the neutron diffusion equation in this method is higher than other numerical methods. In this method the geometry of the problem is divided into primary and secondary cell that primary cell platform is fixed but there is the possibility of making...
Neutron Noise Calculation Using High order Nodal Expansion Method
, M.Sc. Thesis Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
This study consists of two parts: steady state calculations and neutron noise calculations in the frequency domain for two rectangular and hexagonal geometries. In the steady state calculation, the neutron diffusion and its adjoint equations are approximated by two-dimensional coordinates in two-group energy and are solved using the average current nodal expansion method. Then, by considering the node size in the dimensions of a fuel assembly, different orders of flux expansion are investigated. For verification purposes, the calculations have been performed by power iteration method for two test problems of BIBLIS-2D and IAEA-2D. For rectangular geometry with increasing flux expansion order...
Range Verification and Dose Evaluation in Proton Therapy by Using Monte Carlo Method
, M.Sc. Thesis Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
In this study, the emission and detection of secondary particles such as gamma and neutron along the beam path was investigated to evaluate proton range during treatment. Simulations and detections were performed by Geant4 toolkit that was developed base on Monte Carlo method. First, a mono-energetic proton beams irradiated a homogeneous water phantom. Then the factors influencing the accuracy of beam range estimation have been investigated and finally proton beam range was evaluated by simulation the human eye phantom. According to the results, the accuracy of range verification by prompt gamma is very high, about 1 mm. In phantoms with more oxygen in their composition, the percentage of...
Simulation of Power Reactor Noise Based on Transport Equation
, Ph.D. Dissertation Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
In every nuclear reactor core, there are neutron flux fluctuations around the mean value. The neutron noise is the deviation between the time-dependent neutron flux and its expected value, assuming that all process is stationary and ergodic in time. These fluctuations can be caused by the stochastic nature of neutron interactions or mechanical oscillations in the reactor cores. In a reactor working at a high-power level, mechanical fluctuations are the major cause of the fluctuations measured by the detectors. These mechanical vibrations include control rod or fuel rod vibrations, changes in the incoming coolant speed, or changes in the coolant temperature. These perturbations are seen as...
Accurate Generation of Scintillator Detector Energy Spectrum by Monte Carlo Method for Gamma-Ray Spectrum Analysis Using whole Spectrum Data
, M.Sc. Thesis Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
Generating the experimental spectrum for all gamma-ray elements is difficult and practically impossible in some industrial applications. Furthermore, in many applications such as neutron activation, it is not possible to obtain an experimental spectrum for a single isotope on account of the fact that the gamma-ray spectrum obtained from neutron activation is derived from the gamma-rays of various elements in a sample. Therefore, using computational techniques like Monte Carlo method for calculating the detector response functions seems an appropriate solution. The detector response functions for 3''×3'' NaI detectors have been simulated using MCNPX 2.7 code and Ptrac card in this project....
Feasibility Study of Rapid Estimation of SF Parameter of Cells Using the Monte Carlo Simulation of Carbon Therapy in Combination with Cascade Feed-forward Neural Network (CFNN) by Bayesian Regularization Learning Algorithm
, M.Sc. Thesis Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
Our main goal in this project is to calculate the cell survival fraction (SF) in unhealthy tissue after carbon ion beams are emitted and hit the target tissue. At present, in the stages of carbon therapy, it usually takes a long time to make the necessary calculations for that particular person, and this makes the treatment process progress slowly, but in this project, we intend to address this problem with Eliminating the use of neural network. The main reason is the use of carbon ions to increase efficiency in the treatment of tumors and deep cancerous tissues. To do this project, we first convert the carbon ion energy to the dose absorbed at the desired point. For this phase of the...
Rapid Estimation of SF Parameter of Cells Using the Monte Carlo Simulation of Proton Therapy in Combination with Cascade Feed-forward Neural Network (CFNN)with LM Learning Algorithm
, M.Sc. Thesis Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
In radiation therapy, different radiation were used to cancer treatment. The use of hadrons is more common than gamma rays for cancer treatment. Hadron therapy is appropriate to treat cancers that are close to organs at risk and can be severely damaged by X-rays during treatment. The survival fraction of cells was irradiated is important during treatment. In this study, using Geant code, simulation of choroidal melanoma tumor in the phantom of the eye was performed. The linear energy transfer and dose data of proton beam radiation were calculated, in addition, the cell survival fraction were calculated using Survival codeand the Geant simulation data such as number of inputs and dose with...
Optimizing the Efficiency of Thermal to Electrical Power Conversion in Static Heat Converter for Radioisotope Heat Sources
, M.Sc. Thesis Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
Nowadays, due to various advances in the space industry, nuclear generators are widely used to provide the electrical power needed by satellites, spacecrafts and other instruments used in space; So that there is practically no other way than the nuclear industry to produce the required power in some missions; Therefore, it is vital to design nuclear mechanism-based electric power generators. Radioisotope Thermoelectric Generator (RTG) is one of the best and most widely used of these generators. These generators consist of 3 main components: heat source, thermoelectric converter and body. The source of thermal power used is General-purpose heat source (GPHS), which produces thermal power at...
Coupling of ANISN & MCNP Computer Codes for Neutron Importance Calculation
, M.Sc. Thesis Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
For many reasons the Montecarlo method is chosen for investigation of problems like shielding. Nowadays, superiority of utilizing the codes based upon Montecarlo methods such as using cross sections with continuous energy or using exact models of complex geometries, is obvious for everyone. But obtaining results with least possible variance is extremely important. variance reduction methods usedfor achieving exact results with less calculating time. Up to now in the MCNP codes, many methods have been utilized for reducing the variance. One of the methods which have recently attracted attentions is using the importance function. For obtaining importance function we need to solve the...
Designing a System for Radiation Detection in the Marine Environment
, M.Sc. Thesis Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract
The most common method for radioactivity measurements in the environmental sciences is gamma ray spectrometry. The traditional laboratory analysis is a time consuming method and demands special acilities and know-how for the chemical treatment of the samples and half-life limitations . NaI(Tl) are the most common crystals for long-term measurements due to low consumption, good efficiency and low cost ,but HPGe is not suitable due to the high cost of it and cooling system.The aim of this study was designing and making enclosure and basehold for a NaI(Tl) detector to use the water at a depth of 100meters and calibrate.Applications of this detector is measuring the pollution caused by the...