Sharif Digital Repository

If nuclear radiation is not weakened and absorbed, it can cause serious damage to human devices or the environment. Dual-purpose dry casks are one of the sources of nuclear radiation production as a place to keep the spent fuel assemblies. One of the ways to protect against them is to use shield. One of the most common of them is composite shield. The composite used in this study has a polymer matrix. The matrix of the used polymer is epoxy resin. In this study, two separate composites have been made to attenuate neutron and gamma radiation. In this thesis, first, by using the MCNPX code, the necessary simulations for the design and construction of composite shields have been carried out.... 

Since the beginning of discussion on neutron-gamma discrimination, different methods were suggested for this purpose. In early 2000s, Analogue methods were mainly used to discriminate the spectrum in mixed fields which mostly depends on electronic modules used in the lab setup; Nowadays, with improvements in digital pulse processing techniques, analyzing mixed neutron-gamma fields for pulse discrimination is also available.In this research, the recent MC developed toolkit called GEANT4 is used to simulate the 3-inch BC501A scintillation detector and 500 mCi Am-Be source (that emits neutron and gamma at the same time). For the next step, an experiment with same source and detector was done... 

The main purpose of this research is to select a suitable solution for regeneration of ion exchange resins of TR system of Bushehr nuclear power plant and also to compare and evaluate the possibility of using industrial cationic resin (Amberlite IR120) instead of nuclear grade cationic resin (Amberlite IRN97).First, different solutions for regeneration of cationic resins were investigated. For this purpose, three strong acids including hydrochloric acid (HCl), sulfuric acid (H2SO4) and nitric acid (HNO3) were selected and each of these acids was tested in two different concentrations and two different linear velocities; Also, to evaluate the appropriate solution for anionic resin... 

Neutron Radiography is one of the non-destructive testing methods of materials and due to its special application compared to gamma or X-ray imaging, it is highly regarded and developed in the world. A special application of neutron imaging is the imaging of materials with a low atomic number, even when coated with a material with a high atomic number. Neutron Radiography has different methods for capturing and recording images by neutrons. Among these methods, the following 7 methods can be mentioned: Neutron radiography (film), Track Etch, Digital neutron imaging, Neutron camera (DR System), Image plates (CR System), Flat panel silicon detectors (DR system) and Micro channel plates (DR... 

The widespread use of nuclear energy leads to obtain detailed information, such as neutron flux distribution (power) which is very effective in designing and evaluating the reactor safety. The neutron flux (power) reconstruction method uses the homogeneous flux distribution and the heterogeneous form function in a fuel assembly to calculate the heterogeneous power in the fuel rods. Therefore, this method has been widely developed in the last two decades. This study investigates to calculate two-dimensional and two-group neutron flux (power) in the fuel rod for both quadrilateral and hexagonal geometry related to core results by using nodal method. To achieve a more complete program and join... 

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.... 

Effective delayed neutrons fraction is one of important reactor dynamic parameters. Prompt neutrons decay constant is one of another reactor dynamic parameters that for its relation with effective delayed neutrons fraction is important. The Feynman- alpha method is one of famous methods in noise analysis. In this method, prompt neutrons decay constant can be obtained by obtaining variance to mean ratio of a detector counts in different time windows and fitting a specific formula to these ratios. In previous applied works, required data of Feynman- alpha method were obtaining mainly by experimental data or Monte Carlo simulations. In experimental way, detectors with high efficiency are needed... 

Neutron noise equations, which are obtained by assuming small perturbations of macroscopic cross sections around a steady-state neutron field and by subsequently taking the Fourier transform in the frequency domain, have been usually solved by analytical techniques or by resorting to diffusion theory, but in this thesis, in order to increase of accuracy of neutron noise calculation, has been used transport approximation for neutron noise calculation and the Monte Carlo method has been used to solve transport equation of the neutron noise in the frequency domain. Since the transport equation of the neutron noise is a complex equation, a new Monte Carlo technique for treating complex-valued... 

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... 

Steels as structural materials of pressure vessels of nuclear reactors, in addition to high temperatures and pressures, are exposed to ionizing radiation such as neutrons. The primary effects of damage caused by exposing these solids to radiation are the displacement of atoms from their equilibrium positions and the formation of point defects and damage clusters caused by damage accumulation due to displacement cascades produced by transmitting the energy of the incident particle to an atom by interactions such as elastic and inelastic scatterings neutrons with the nucleus. These microstructural changes cause large structural defects such as swelling, cracking, cracking, creep, reducing... 

Scintillation detectors are categorized as low energy resolution detectors in the classification of energy-sensitive detectors. The presence of broad peaks in the gamma-ray spectra of these detectors increases the overlapping probability of photopeaks of different isotopes and makes it extremely difficult to identify the spectra. The common method for identifying and quantifying of elements in a sample using nuclear spectroscopy is to use photopeak counts obtained with high-energy resolution detectors such as semiconductor detectors. To deal with the low energy resolution of scintillation detectors in quantitative analysis, the whole information of gamma-ray spectra is used instead of using... 

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... 

The present M.Sc. thesis consists of two sections including static calculation and neutron noise calculations 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 equation is based on Boundary Element Method (BEM). The result are benchmarked against the valid results for BIBLIS-2D and IAEA -2D benchmark problem. In the second section, neutron noise calculation are performed for two types of noise sources, i.e. absorber of variable strength and Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position (ILOFAIP). The... 

To study the structure of materials, among materials containing hydrogen, an instrument called the "small-angle neutron scattering instrument" is used. Having a monochromatic neutron beam is the most basic component of this instrument. This beam is often produced using a device called the "neutron velocity selector". The neutron velocity selector is a rotating mechanical piece that allows for passage of neutrons at a certain speed, according to its rotational speed. In fact, this device produces a monochromatic neutron beam with continuous flux. Until now, it has been designed and manufactured in a variety of models, which are generally divided into two groups of multi-disc and multi-blade... 

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... 

The core behaviour of a nuclear reactor might be fluctuated (deviations from the normal operating conditions) in operating conditions due to various reasons. Control rod vibrations and the alterations of coolant temperature and velocity could be the main reasons for the fluctuations. These fluctuations lead to neutronic flux noise and subsequently power noise. The main objective of the current thesis was to study the thermal-hydraulic noise of the PWRs in two-phase with the emphasis on VVER-1000, which is similar to the Bushehr-1 nuclear reactor. By considering the possibility of existing two-phase flow (maximum allowable quality is about 14%) in hot channel of the PWRs, thermal-hydraulic... 

Since direct measurement of neutron energy, unlike measuring the energy of ionizing radiation, is difficult, and the use of indirect methods for detecting and measuring neutron energy with high resolution and acceptable efficiency are not possible, the neutron time of flight method is the only direct neutron energy spectroscopy method. This issue is considered and evaluated in two methods: direct and scattering neutron time of flight. For this purpose, we have considered the 241Am-9Be source. In this method, neutron velocity can be determined by measuring time. is the time for the neutron that travels at distance of . The next method is the dispersed neutron time of flight method, in... 

The safety and optimal performance of nuclear reactors require online monitoring in the core. One of the most important requirements of core monitoring is the knowledge at all time of the neutron flux distribution in the core. The present M.Sc thesis describes a method which avoids the solution of time dependent neutron diffusion equation and uses online readings of the fixed in-core neutron detectors to reconstruct the three-dimensional (3D) neutron flux distribution. The essential idea of nodal synthesis method is separation of time and space dependence of the neutron flux distribution. The time dependent section of the flux distribution is determined by neutron detector readings and space... 

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,... 

Neutron activation is a nuclear method for determination of elements in each sample. In this method, the samples are exposed to neutron irradiation by neutrons which are usually produced by research reactor, and then the core becomes unstable due to capturing the neutron and emits gamma rays. The emitted gamma rays can be detected by a detector like a HPGe semiconductor which has a high resolution. All nucleuses have a probability of neutron absorption, expressed by the absorption cross section area. The radioactive nuclei value depends on the nuclear half-life and the number of unstable nuclei. The only Windows environment commercial software for neutron activation analysis was K0 for...