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#### A new neutron energy spectrum unfolding code using a two steps genetic algorithm

, Article Nuclear Instruments and Methods in Physics Research, Section A: Accelerators, Spectrometers, Detectors and Associated Equipment ; Volume 811 , 2016 , Pages 82-93 ; 01689002 (ISSN) ; Hosseini, S. A ; Sohrabpour, M ; Sharif University of Technology
Elsevier

Abstract

A new neutron spectrum unfolding code TGASU (Two-steps Genetic Algorithm Spectrum Unfolding) has been developed to unfold the neutron spectrum from a pulse height distribution which was calculated using the MCNPX-ESUT computational Monte Carlo code. To perform the unfolding process, the response matrices were generated using the MCNPX-ESUT computational code. Both one step (common GA) and two steps GAs have been implemented to unfold the neutron spectra. According to the obtained results, the new two steps GA code results has shown closer match in all energy regions and particularly in the high energy regions. The results of the TGASU code have been compared with those of the standard...

#### A novel neutron energy spectrum unfolding code using particle swarm optimization

, Article Radiation Physics and Chemistry ; Volume 136 , 2017 , Pages 9-16 ; 0969806X (ISSN) ; Sohrabpour, M ; Sharif University of Technology
Elsevier Ltd
2017

Abstract

A novel neutron Spectrum Deconvolution using Particle Swarm Optimization (SDPSO) code has been developed to unfold the neutron spectrum from a pulse height distribution and a response matrix. The Particle Swarm Optimization (PSO) imitates the bird flocks social behavior to solve complex optimization problems. The results of the SDPSO code have been compared with those of the standard spectra and recently published Two-steps Genetic Algorithm Spectrum Unfolding (TGASU) code. The TGASU code have been previously compared with the other codes such as MAXED, GRAVEL, FERDOR and GAMCD and shown to be more accurate than the previous codes. The results of the SDPSO code have been demonstrated to...

#### Sensitivity analysis of the transmission factor and resolution of a multiblade neutron velocity selector to the various parameters

, Article Radiation Physics and Chemistry ; Volume 177 , December , 2020 ; Hosseini, S. A ; Sharif University of Technology
Elsevier Ltd
2020

Abstract

The neutron velocity selector is a device used to produce a monochromatic neutron beam with continuous flux. The purpose of the present study is to investigate the sensitivity of the transmission factor and resolution of a multiblade neutron velocity selector to the various parameters using the McStas software. To this end, two instruments were created using the Arm, Progress_bar, Source_simple, DivMonitor, L_monitor, Guide_channeled and V_selector components of the McStas software. The used V_selector component to simulate the multiblade neutron velocity selector was created by considering three assumptions: 1. The absorption of colliding neutrons to selector blades, 2. No interaction of...

#### Simulation of the direct geometry spectrometer for neutron time of flight based on the Monte Carlo method to calculate the energy spectrum of the neutron source

, Article Nuclear Instruments and Methods in Physics Research, Section A: Accelerators, Spectrometers, Detectors and Associated Equipment ; Volume 949 , 2020 ; Mehrabi, M ; Sharif University of Technology
Elsevier B.V
2020

Abstract

In the present study, simulation of a direct geometry spectrometer for neutron Time-of-Flight (nTOF) is performed based on the Monte Carlo method to calculate the energy spectrum of the neutron source. To this end, the ability to simulate the simultaneous emission (such as the simultaneous emission of the neutron and gamma particles from the241Am-9Be neutron source) and the neutron Pulse Height Distribution (nPHD) were added to the MCNPX computer code. In addition, a post-processing software was developed to analyze the massive amounts of the data in the output of the PTRAC card. The application of the nTOF as a direct method to calculate the neutron spectrum eliminates the error induced by...

#### Development of a Software for Rreconstruction of Neutron spectrum

, M.Sc. Thesis Sharif University of Technology ; Vossoughi, Naser (Supervisor) ; Etaati, Gholamreza (Co-Advisor)
Abstract

Dealing with natural and handmade radioactive materials and sources is of major aspects of nuclear science and technology. Useful applications of these materials and sources in different fields, such as energy production and health physics, caused the necessity of developing the detection and radiation protection methods. Each of these methods uses different equipment and approaches which are based on different kinds of radiations and radioactive sources. Despite the given ability of radiation detection, in some cases, recognition, distinguish, and estimation of a source radiation level is impossible due to bad effects of these equipment on measured spectrum. Detection of neutron spectrum in...

#### Development and experimental validation of a correlation monitor tool based on the endogenous pulsed neutron source technique

, Article Metrology and Measurement Systems ; Volume 24, Issue 3 , 2017 , Pages 441-461 ; 20809050 (ISSN) ; Khalafi, H ; Vosoughi, N ; Khakshournia, S ; Sharif University of Technology
Abstract

A correlation measuring tool for an endogenous pulsed neutron source experiment is developed in this work. Paroxysmal pulses generated by a bursts of neutron chains are detected by a 10-kbit embedded shift register with a time resolution of 100 ns. The system is implemented on a single reprogrammable device making it a compact, cost-effective instrument, easily adaptable for any case study. The system was verified experimentally in the Esfahan heavy-water zero power reactor (EHWZPR). The results obtained by the measuring tool are validated by the Feynman-α experiment, and a good agreement is seen within the boundaries of statistical uncertainties. The theory of the methods is briefly...

#### Simulated and measured neutron/gamma light output distribution for poly-energetic neutron/gamma sources

, Article Journal of Instrumentation ; Volume 13, Issue 3 , March , 2018 ; 17480221 (ISSN) ; Zangian, M ; Aghabozorgi, S ; Sharif University of Technology
Institute of Physics Publishing
2018

Abstract

In the present paper, the light output distribution due to poly-energetic neutron/gamma (neutron or gamma) source was calculated using the developed MCNPX-ESUT-PE (MCNPX-Energy engineering of Sharif University of Technology-Poly Energetic version) computational code. The simulation of light output distribution includes the modeling of the particle transport, the calculation of scintillation photons induced by charged particles, simulation of the scintillation photon transport and considering the light resolution obtained from the experiment. The developed computational code is able to simulate the light output distribution due to any neutron/gamma source. In the experimental step of the...

#### Neutron spectroscopy with soft computing: Unfolding of the neutron energy spectrum using the developed computer code based on Adaptive Group of Ink Drop Spread (AGIDS)

, Article Journal of Instrumentation ; Volume 14, Issue 3 , 2019 ; 17480221 (ISSN) ; Sharif University of Technology
Institute of Physics Publishing
2019

Abstract

This paper presents the developed computer codes based on the Adaptive Group of Ink Drop Spread (AGIDS) algorithm [1] for the reconstruction of the energy spectrum of the neutron source. The required data are generated via simulation of the neutron pulse height distributions due to randomly generated energy spectra with the MCNPX-ESUT computer code [2]. The simulated neutron pulse height distributions and corresponding randomly generated energy spectra are the input and output data of the developed computer code, respectively. As a case study, the 241Am-9Be neutron source is studied and the simulation of the neutron pulse height distribution of the NE-213 liquid organic scintillator is...

#### Neutronic analysis of HPLWR fuel assembly cluster

, Article Annals of Nuclear Energy ; Volume 50 , December , 2012 , Pages 38-43 ; 03064549 (ISSN) ; Salehi, A. A ; Jahanfarnia, G ; Abbaspour Tehrani Fard, A ; Sharif University of Technology
2012

Abstract

In the present study the neutronic analysis of fuel assembly cluster of the HPLWR is discussed. Neutronic calculations are performed using WIMS-D4 and CITATION codes. Thermal-hydraulic code containing the properties and specifications of the fuel assembly of HPLWR is utilized. The calculated axial power in each selected control volume is used in the thermal-hydraulic code to get the properties of the fluid and fuel needed for further neutronic analysis. The process of coupling continues until convergence is achieved. Finally, the obtained neutronic results including axial power distribution, neutron flux, and power peaking factors are discussed in the present article

#### Monte Carlo simulation and benchmarking of pulsed neutron experiments in variable buckling Beo systems

, Article Annals of Nuclear Energy ; Volume 36, Issue 5 , 2009 , Pages 547-549 ; 03064549 (ISSN) ; Ezzati, A. O ; Sharif University of Technology
2009

Abstract

Pulsed neutron decay simulation of BeO moderator as a result of injecting 14 MeV neutron pulses into finite size systems are carried out with the MCNP Monte Carlo code. The simulated decay constants as compared against a previous experimental work showed variations of about 2%. The resulting decay constants and the fitted diffusion coefficients based on the simulation results are found to agree with the experimental results within a margin of about 4%. © 2009 Elsevier Ltd. All rights reserved

#### Direct Discrete Method (DDM) and its application to neutron transport problems

, Article Scientia Iranica ; Volume 14, Issue 1 , 2007 , Pages 78-85 ; 10263098 (ISSN) ; Salehi, A. A ; Shahriari, M ; Heshmatzadeh, M ; Sharif University of Technology
Sharif University of Technology
2007

Abstract

The objective of this paper is to introduce a new direct method for neutronic calculations. This method, called Direct Discrete Method (DDM), is simpler than the Neutron Transport Equation and more compatible with the physical meanings of the problem. The method, based on the physics of the problem, initially-runs through meshing of the desired geometry. Next, the balance equation for each mesh interval is written. Considering the connection between the mesh intervals, the final discrete equation series are directly obtained without the need to first pass through the set-up of the neutron transport differential equation. In this paper, a single and multigroup neutron transport discrete...

#### Adaptive group of ink drop spread: A computer code to unfold neutron noise sources in reactor cores

, Article Nuclear Engineering and Technology ; 2017 ; 17385733 (ISSN) ; Esmaili PaeenAfrakoti, I ; Sharif University of Technology
Korean Nuclear Society
2017

Abstract

The present paper reports the development of a computational code based on the Adaptive Group of Ink Drop Spread (AGIDS) for reconstruction of the neutron noise sources in reactor cores. AGIDS algorithm was developed as a fuzzy inference system based on the active learning method. The main idea of the active learning method is to break a multiple input-single output system into a single input-single output system. This leads to the ability to simulate a large system with high accuracy. In the present study, vibrating absorber-type neutron noise source in an International Atomic Energy Agency-two dimensional reactor core is considered in neutron noise calculation. The neutron noise...

#### Evaluation of a new neutron energy spectrum unfolding code based on an adaptive neuro-fuzzy inference system (ANFIS)

, Article Journal of Radiation Research ; Volume 59, Issue 4 , 2018 , Pages 436-441 ; 04493060 (ISSN) ; Esmaili Paeen Afrakoti, I ; Sharif University of Technology
Oxford University Press
2018

Abstract

The purpose of the present study was to reconstruct the energy spectrum of a poly-energetic neutron source using an algorithm developed based on an Adaptive Neuro-Fuzzy Inference System (ANFIS). ANFIS is a kind of artificial neural network based on the Takagi-Sugeno fuzzy inference system. The ANFIS algorithm uses the advantages of both fuzzy inference systems and artificial neural networks to improve the effectiveness of algorithms in various applications such as modeling, control and classification. The neutron pulse height distributions used as input data in the training procedure for the ANFIS algorithm were obtained from the simulations performed by MCNPX-ESUT computational code...

#### High accurate three-dimensional neutron noise simulator based on GFEM with unstructured hexahedral elements

, Article Nuclear Engineering and Technology ; Volume 51, Issue 6 , 2019 , Pages 1479-1486 ; 17385733 (ISSN) ; Sharif University of Technology
Korean Nuclear Society
2019

Abstract

The purpose of the present study is to develop the 3D static and noise simulator based on Galerkin Finite Element Method (GFEM) using the unstructured hexahedral elements. The 3D, 2G neutron diffusion and noise equations are discretized using the unstructured hexahedral by considering the linear approximation of the shape function in each element. The validation of the static calculation is performed via comparison between calculated results and reported data for the VVER-1000 benchmark problem. A sensitivity analysis of the calculation to the element type (unstructured hexahedral or tetrahedron elements) is done. Finally, the neutron noise calculation is performed for the neutron noise...

#### Reconstruction of neutron flux distribution by nodal synthesis method using online in-core neutron detector readings

, Article Progress in Nuclear Energy ; 2020 ; Ghofrani, M. B ; Sharif University of Technology
Elsevier Ltd
2020

Abstract

The safety and optimal performance of nuclear reactors require online monitoring in the core. The present paper describes a method that avoids the solution of the time-dependent neutron diffusion equation, and it uses online readings of the fixed in-core neutron detectors to reconstruct the three-dimensional (3D) neutron flux distribution. The essential idea of the nodal synthesis method is the separation of time and space-dependence of the neutron flux distribution. The time-dependent section of the flux distribution is determined by in-core neutron detector readings, and the space-dependent section is obtained from pre-computed harmonics of the neutron diffusion equation. In online...

#### Neutron spectrum unfolding using artificial neural network and modified least square method

, Article Radiation Physics and Chemistry ; Volume 126 , 2016 , Pages 75-84 ; 0969806X (ISSN) ; Sharif University of Technology
Elsevier Ltd
2016

Abstract

In the present paper, neutron spectrum is reconstructed using the Artificial Neural Network (ANN) and Modified Least Square (MLSQR) methods. The detector's response (pulse height distribution) as a required data for unfolding of energy spectrum is calculated using the developed MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). Unlike the usual methods that apply inversion procedures to unfold the energy spectrum from the Fredholm integral equation, the MLSQR method uses the direct procedure. Since liquid organic scintillators like NE-213 are well suited and routinely used for spectrometry of neutron sources, the neutron pulse height distribution is...

#### Energy spectra unfolding of fast neutron sources using the group method of data handling and decision tree algorithms

, Article Nuclear Instruments and Methods in Physics Research, Section A: Accelerators, Spectrometers, Detectors and Associated Equipment ; Volume 851 , 2017 , Pages 5-9 ; 01689002 (ISSN) ; Esmaili Paeen Afrakoti, I ; Sharif University of Technology
Abstract

Accurate unfolding of the energy spectrum of a neutron source gives important information about unknown neutron sources. The obtained information is useful in many areas like nuclear safeguards, nuclear nonproliferation, and homeland security. In the present study, the energy spectrum of a poly-energetic fast neutron source is reconstructed using the developed computational codes based on the Group Method of Data Handling (GMDH) and Decision Tree (DT) algorithms. The neutron pulse height distribution (neutron response function) in the considered NE-213 liquid organic scintillator has been simulated using the developed MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif...

#### Optimization of Parameters of Neutron Activation Analysis Using K0-IAEA Code

, M.Sc. Thesis Sharif University of Technology ; Vossoughi, Naser (Supervisor)
Abstract

There are several methods for identifying unknown elements of an unknown compound. One of the common methods for this purpose, using neutron activation analysis. In conjunction with this method, different computer codes and softwares have been developed and used. One of these methods is K0 method that is a single comparator method. The code that based on this method is K0-IAEA code that is written by the International Atomic Energy Agency and access to it is free. Foremost objective of this study is to determine how to work with this code. In order to evaluate the accuracy of the results of the code, examples have taken into consideration the amount of time previously unknown elements by...

#### , 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...

#### Determination of Reactor Dynamic Parameters Using Correlation Equation

, M.Sc. Thesis Sharif University of Technology ; Vosoughi, Naser (Supervisor)
Abstract

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...