International Journal of Modern Physics E (2183013)(8)
A plasma boundary reconstruction code has been designed by using current filament method to calculate the magnetic flux and consequently plasma boundary in Damavand tokamak. Hence, a computer-based code "The Plasma Boundary Reconstruction Code in Tokamak (PBRCT)" was developed to make a graphical user interface and to speed up the plasma boundary estimation algorithm. All required tools as the plasma boundary and magnetic surface display (MSD), error display, primary conditions and modeling panel as well as a search motor to determine a good position and number of the current filaments to find a precise model have been considered. The core is a 3000 lines Matlab code and the graphical user interface is 10,000 lines in C# language. © 2014 World Scientific Publishing Company.
International Journal of Nuclear Energy Science and Technology (17416361)(1)pp. 37-48
In this paper, the effects of internal reflector size on the neutron flux of PBMR reactors by means of MCNP code has been investigated as a part of establishing Monte Carlo computation system for PBMR core analysis. At first, high temperature library in this reactor is built by NJOY code then, cross sections in the library are used by MCNP code. Because of complex fuel structure, several simplifications are assumed in order to limit the need for any further approximations when defining code models. In neutronic simulation, at first, all the important neutronic parameters such as thermal and fast flux distribution and power density of the reactor core are compared with the results of other codes and the accuracy of the simulation has been validated. Then the best size of internal reflector has been chosen by comparing the neutron flux in some different sizes and obtaining the best form factor. Copyright © 2014 Inderscience Enterprises Ltd.
International Journal of Modern Physics E (2183013)(5)
The shape and position of the plasma and consequently the plasma boundary are determined by using the Current Filament (CF) method from the experimental data of the magnetic measurements in Damavand tokamak. The method can calculate the magnetic flux without solving the equilibrium equation directly by coupling with the Current Moment (CM) method. The plasma and current-carrying coils in the tokamak will be modeled by using this method as some virtual filaments that will enable us to calculate the flux and consequently the plasma boundary. To calculate the flux of these virtual filaments, one needs to determine the Green Function and the inverse by means of the Singular Value Decomposition (SVD) method. Finally, the model was evaluated by employing 12 independent pickup coils with mean error of less than 2%. The aim of this paper is to give a brief exposition of CF method applied in Damavand tokamak. © 2014 World Scientific Publishing Company.
Pilehvar a.f., ,
Aghaie m., ,
Esteki, M.H.,
Zolfaghari A.,
Minuchehr A.,
Daryabak a., ,
Safavi a., Annals of Nuclear Energy (3064549)pp. 185-194
In this study, using porous media approach, the compressible flow within the core of a Pebble Bed Modular Reactor (PBMR) is simulated. This reactor has been composed of coated fuel particles with compressible gas as a coolant and graphite as a moderator and reflector. Containing about 450,000 fuel complexes, the reactor core is considered a porous medium subject to high temperature and high pressure helium flow. The porosity and permeability parameters are calculated and utilized. The coolant compressibility has been introduced as an effective parameter in the thermal-hydraulic analysis. Accordingly, using the ANSYS CFX code, which is capable of simulating compressible flow in porous media, the reactor core is simulated and thermal-hydraulic parameters of the core are obtained through Computational Fluid Dynamic (CFD) approach. The heat flux in the core is first obtained in axial and radial coordinates by MCNP code and is then used in CFD simulation as a semi sine and an algebraic function. The major characteristics of the flow field have been calculated whereby the thermal-hydraulic parameters such as temperature and pressure profiles have been obtained and compared with other data. Comparing the results obtained with other codes and software, the outcomes show that the inclusion of compressibility is reasonable and will lead to a slight difference between the measured and actual temperature, pressure and velocity. In another stage, pressure drop, flow vortices and helium flow lines are explored for two fuel complexes. The empirical formula of pressure drop presented by Kugeler and Schulten is modified and gas density is considered as a function of the core length. Fuel complexes in the reactor core are randomly arranged. However, because Body Center Cubic (BCC) is the closest arrangement to the random distribution, flow parameters are obtained using the BCC arrangement and they are found to deviate very slightly when compared with predictions of other codes. © 2013 Elsevier Ltd. All rights reserved.
