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Nuclear Engineering and Technology (17385733)(8)pp. 1603-1610
Due to ever-growing advancements in computers and relatively easy access to them, many efforts have been made to develop high-fidelity, high-performance, multi-physics tools, which play a crucial role in the design and operation of nuclear reactors. For this purpose in this study, the neutronic Monte Carlo and thermal-hydraulic sub-channel codes entitled MCNP and COBRA-EN, respectively, were applied for external coupling with each other. The coupled code was validated by code-to-code comparison with the internal couplings between MCNP5 and SUBCHANFLOW as well as MCNP6 and CTF. The simulation results of all code systems were in good agreement with each other. Then, as the second problem, the core of the VVER-1000 v446 reactor was simulated by the MCNP4C/COBRA-EN coupled code to measure the capability of the developed code to calculate the neutronic and thermohydraulic parameters of real and industrial cases. The simulation results of VVER-1000 core were compared with FSAR and another numerical solution of this benchmark. The obtained results showed that the ability of the MCNP4C/COBRA-EN code for estimating the neutronic and thermohydraulic parameters was very satisfactory. © 2020
Kerntechnik (9323902)(1)pp. 15-25
Since estimating the minimum departure from nucleate boiling ratio (MDNBR) requires complex calculations, an alternative method has always been considered. One of these methods is neural network. In this study, the Back Propagation Neural network (BPN) and Radial Basis Function Neural network (RBFN) are introduced and compared in order to estimate MDNBR of the VVER-1000 light water reactor. In these networks, the MDNBR were predicted with the inputs including core mass flux, core inlet temperature, pressure, reactor power level and position of the control rods. To obtain the data required to design these neural networks, an externally coupledcode was developed and its ability to estimate the thermo-hydraulic parameters of the VVER-1000 reactor was compared with other numerical solutions of this benchmark and the Final Safety Analysis Report (FSAR). After ensuring the accuracy of this coupled-code, MDNBR was calculated for 272 different conditions of reactor operating, and it was used to design BPN and RBFN. Comparison of these two neural networks revealed that when the output SMEs of the two systems were approximately the same, the training process in RBFN was much faster than in BPN and the maximum network error in RBFN was less than in BPN. © Carl Hanser Verlag, München.
Kerntechnik (9323902)(5)pp. 351-358
One of the most important issues in nuclear reactor operation and its designing is considering the interaction between thermal hydraulics and neutronics physics because there is an important relationship between the states of fluid and neutron spectrum and power distribution. In this research, the MCNP4C and COBRA-EN nuclear codes were coupled with each other to precisely analyze the fuel assembly of the light water reactor core. This coupling was carried out using iterative processes between the linked neutronic and thermal-hydraulic codes applying successive procedures while the desired convergence was made in both. The newly designed code was checked for three test problems, and the obtained results showed the improvement of the computations procedures by the developed code. © 2020 Carl Hanser Verlag. All rights reserved.
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.
