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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 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,...
Development of Neutron Noise Simulator Based on the Boundary Element Method
, M.Sc. Thesis Sharif University of Technology ; Hosseini, Aboulfazl (Supervisor)
Abstract
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...
Development of 3D neutron noise simulator based on GFEM with unstructured tetrahedron elements
, Article Annals of Nuclear Energy ; Volume 97 , 2016 , Pages 132-141 ; 03064549 (ISSN) ; Vosoughi, N ; Sharif University of Technology
Elsevier Ltd
Abstract
In the present study, the neutron noise, i.e. the stationary fluctuation of the neutron flux around its mean value is calculated based on the 2G, 3D neutron diffusion theory. To this end, the static neutron calculation is performed at the first stage. The spatial discretization of the neutron diffusion equation is performed based on linear approximation of Galerkin Finite Element Method (GFEM) using unstructured tetrahedron elements. Using power iteration method, neutron flux and corresponding eigen-value are obtained. The results are then benchmarked against the valid results for VVER-1000 (3D) benchmark problem. In the second stage, the neutron noise equation is solved using GFEM and...
Pin Power Reconstruction Method by Nodal Core Calculation Results
, M.Sc. Thesis Sharif University of Technology ; Vosoughi, Naser (Supervisor) ; Ghaffari, Mohsen (Supervisor)
Abstract
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...
Reconstruction of neutron flux distribution by nodal synthesis method using online in-core neutron detector readings
, Article Progress in Nuclear Energy ; Volume 131 , 2021 ; 01491970 (ISSN) ; Ghofrani, M. B ; Sharif University of Technology
Elsevier Ltd
2021
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...
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...
Online Reconstruction of Neutron Flux Distribution using BNPP Operating Data
, M.Sc. Thesis Sharif University of Technology ; Ghofrani, Mohammad Bagher (Supervisor) ; Raji, Mohammad Hossein (Co-Advisor)
Abstract
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...
3D neutron diffusion computational code based on GFEM with unstructured tetrahedron elements: A comparative study for linear and quadratic approximations
, Article Progress in Nuclear Energy ; Volume 92 , 2016 , Pages 119-132 ; 01491970 (ISSN) ; Sharif University of Technology
Elsevier Ltd
2016
Abstract
In the present study, the comparison between the results obtained from the linear and quadratic approximations of the Galerkin Finite Element Method (GFEM) for neutronic reactor core calculation was reported. The sensitivity analysis of the calculated neutron multiplication factor, neutron flux and power distributions in the reactor core vs. the number of the unstructured tetrahedron elements and order of the considered shape function was performed. The cost of the performed calculation using linear and quadratic approximation was compared through the calculation of the FOM. The neutronic core calculation was performed for both rectangular and hexagonal geometries. Both the criticality and...
A new approach for solution of time dependent neutron transport equation based on nodal discretization using MCNPX code with feedback
, Article Annals of Nuclear Energy ; Volume 133 , 2019 , Pages 519-526 ; 03064549 (ISSN) ; Salehi, A. A ; Vosoughi, N ; Sharif University of Technology
Elsevier Ltd
2019
Abstract
This paper proposes a new method for solving the time-dependent neutron transport equation based on nodal discretization using the MCNPX code. Most valid nodal codes are based on the diffusion theory with differences in approximating the leakage term until now. However, the Monte Carlo (MC) method is able to estimate transport parameters without approximations usual in diffusion method. Therefore, improving the nodal approach via the MC techniques can substantially reduce the errors caused by diffusion approximations. In the proposed method, the reactor core is divided into nodes of arbitrary dimensions, and all terms of the transport equation e.g. interaction rates and leakage ratio are...
A new Monte Carlo approach for solution of the time dependent neutron transport equation based on nodal discretization to simulate the xenon oscillation with feedback
, Article Annals of Nuclear Energy ; Volume 141 , 2020 ; Salehi, A. A ; Vosoughi, N ; Sharif University of Technology
Elsevier Ltd
2020
Abstract
In this paper a probabilistic methodology based on core nodalization is proposed to estimate the core power in the presence of xenon oscillation. A time-dependent Monte Carlo neutron transport code named MCSP-NOD is developed for dynamic analysis in arbitrary 3D geometries to simulate xenon oscillations as well as sub-critical condition with feedbacks. The new code is based on the approach adopted in MCNP-NOD which was previously introduced as a tool for core transient analysis using the MCNPX platform. As before, the core is divided into nodes of arbitrary dimensions, and all terms of the transport equation e.g. interaction rates, leakage ratio are estimated using the MC techniques....
Neutron noise simulator based on the boundary element method (BEM)
, Article Annals of Nuclear Energy ; Volume 159 , 2021 ; 03064549 (ISSN) ; Mohamadbeygi, S ; Sharif University of Technology
Elsevier Ltd
2021
Abstract
The purpose of the present study is to develop the neutron diffusion solver and neutron noise simulator based on the Boundary Element Method (BEM). The 2-D, 2-G neutron/adjoint diffusion equation and corresponding neutron/adjoint noise equation were solved using the mentioned method. The developed neutron static and noise simulator based on the finite element method gives accurate results when the more number of the elements is used. The motivation of the present research is to use the boundary element method to reduce the computational cost. The boundary element method attempts to use the given boundary conditions to fit boundary values into the integral equation, rather than values...
Development of SD-HACNEM neutron noise simulator based on high order nodal expansion method for rectangular geometry
, Article Annals of Nuclear Energy ; Volume 162 , 2021 ; 03064549 (ISSN) ; Vosoughi, J ; Vosoughi, N ; Sharif University of Technology
Elsevier Ltd
2021
Abstract
In this study, the SD-HACNEM (Sharif Dynamic - High order Average Current Nodal Expansion Method) neutron noise simulator in two energy groups using a second-order flux expansion method for two-dimensional rectangular X Y-geometry has been developed. In the first step, the calculations were performed for the steady state and results of ACNEM (Average Current Nodal Expansion Method) and HACNEM (High order Average Current Nodal Expansion Method) were examined and compared. To solve the problem, the power iteration algorithm has been used to calculate the distribution of neutron flux and neutron multiplication factor by considering the coarse-mesh (each fuel assembly one node). To validate the...
Simulation and Optimization of Neutron Activation Analysis Using the K0- IAEA Software and Comparison with the Experimental Results
, M.Sc. Thesis Sharif University of Technology ; Vossoughi, Naser (Supervisor) ; Hosseini, Abolfazl (Supervisor)
Abstract
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...
Galerkin and Generalized Least Squares finite element: A comparative study for multi-group diffusion solvers
, Article Progress in Nuclear Energy ; Volume 85 , 2015 , Pages 473-490 ; 01491970 (ISSN) ; Saadatian Derakhshandeh, F ; Sharif University of Technology
Elsevier Ltd
2015
Abstract
Abstract In this paper, the solution of multi-group neutron/adjoint equation using Finite Element Method (FEM) for hexagonal and rectangular reactor cores is reported. The spatial discretization of the neutron diffusion equation is performed based on two different Finite Element Methods (FEMs) using unstructured triangular elements generated by Gambit software. Calculations are performed using Galerkin and Generalized Least Squares FEMs; based on which results are compared. Using the power iteration method for the neutron and adjoint calculations, the neutron and adjoint flux distributions with the corresponding eigenvalues are obtained. The results are then validated against the valid...
Enhanced finite difference scheme for the neutron diffusion equation using the importance function
, Article Annals of Nuclear Energy ; Volume 96 , 2016 , Pages 412-421 ; 03064549 (ISSN) ; Vosoughi, N ; Gharib, M ; Sharif University of Technology
Elsevier Ltd
2016
Abstract
Mesh point positions in Finite Difference Method (FDM) of discretization for the neutron diffusion equation can remarkably affect the averaged neutron fluxes as well as the effective multiplication factor. In this study, by aid of improving the mesh point positions, an enhanced finite difference scheme for the neutron diffusion equation is proposed based on the neutron importance function. In order to determine the neutron importance function, the adjoint (backward) neutron diffusion calculations are performed in the same procedure as for the forward calculations. Considering the neutron importance function, the mesh points can be improved through the entire reactor core. Accordingly, in...
Higher order power reactor noise analysis: the multigroup diffusion model
, Article Annals of Nuclear Energy ; Volume 111 , 2018 , Pages 354-370 ; 03064549 (ISSN) ; Hosseini, A ; Vosoughi, N ; Sharif University of Technology
Elsevier Ltd
2018
Abstract
Power reactor noise analysis is one of the most powerful tools in online monitoring and diagnostics of nuclear power reactors. Unfortunately, since such an analysis belongs to the non-linear “parametric excitation” realm, its theoretical aspects and relations have been mostly carried out after linearization. In this paper a general framework, i.e. the Ladder Expansion Method, is developed to convert such equations to a series of coupled linear equations, up to any desired accuracy. This method is then applied to the single mode random fluctuations of the absorption cross sections in a power reactor which is modelled by the multigroup diffusion equation with multiple delayed neutron groups. A...
Developing 3D neutron transport kernel for heterogeneous structures in an improved method of characteristic (MOC) framework
, Article Progress in Nuclear Energy ; Volume 127 , 2020 ; Hadad, K ; Salehi, A. A ; Moghadam, A ; Sharif University of Technology
Elsevier Ltd
2020
Abstract
Given the importance and complexity of the three-dimensional (3D) neutron transport equation solution, in the current research, a new Modular Ray Tracing (MRT) Algorithm and 3D characteristic kernel for heterogeneous structures are presented. Improvement of memory management and cache coherency are achieved to some acceptable level. Also, parallel implementation of transport algorithm utilizing OpenMP, cause significant reduction in runtime. To validate our Algorithm, first, a self-constituted pin cell and a lattice arrangement are modeled and results are compared with Monte-Carlo simulation. Second, the well-known 3D benchmark, Takeda model one and two, are investigated and results compared...
Validation of a new MCNP-ORIGEN linkage program for burnup analysis
, Article Progress in Nuclear Energy ; Volume 63 , 2013 , Pages 27-33 ; 01491970 (ISSN) ; Abbaspour, A ; Pazirandeh, A ; Sharif University of Technology
2013
Abstract
The analysis of core composition changes is complicated by the fact that the time and spatial variation in isotopic composition depend on the neutron flux distribution and vice versa. Fortunately, changes in core composition occur relatively slowly and hence the burnup analysis can be performed by dividing the burnup period into some burnup spans and assuming that the averaged flux and cross sections are constant during each step. The burnup span sensitivity analysis attempts to find that how much the burnup spans could be increased without any significant deviation in results. This goal has been achieved by developing a new MCNP-ORIGEN linkage program named as MOBC (MCNP-ORIGEN Burnup...
Startup of "cANDLE" burnup in a Gas-cooled Fast Reactor using Monte Carlo method
, Article Annals of Nuclear Energy ; Volume 50 , December , 2012 , Pages 44-49 ; 03064549 (ISSN) ; Abbaspour, A ; Pazirandeh, A ; Sharif University of Technology
2012
Abstract
During the past decade, the CANDLE burnup strategy has been proposed as an innovative fuel cycle and reactor design for complete utilization of uranium resources. In this strategy the shapes of neutron flux, nuclide densities and power density distribution remain constant but the burning region moves in axial direction. The feasibility of this strategy has been demonstrated widely by using the diffusion technique in conjunction with nuclide transmutation equations. On the other hand since the Monte Carlo method provides the exact solution to the neutron transport, the Monte Carlo technique is becoming more widely used in routine burnup calculations. The main objective of this work is startup...