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On a various noise source reconstruction algorithms in VVER-1000 reactor core
, Article Nuclear Engineering and Design ; Volume 261 , 2013 , Pages 132-143 ; 00295493 (ISSN) ; Vosoughi, N ; Sharif University of Technology
2013
Abstract
In present study, the neutron noise source is reconstructed using three different unfolding techniques in a typical VVER-1000 reactor core. In first stage, the neutron noise calculation based on Galerkin finite element method (GFEM) is performed; in which the neutron noise in two energy group due to the noise sources of type absorber of variable strength and vibrating absorber is calculated. The neutron noise due to inadvertent loading of a fuel assembly in an improper position (ILFAIP), as a new defined noise source in the neutron noise studies, is calculated as well. In the second stage, the inversion, zoning and scanning methods are applied for reconstruction of the noise source of type...
Calculation of the deterministic optimum loading pattern of the BUSHEHR VVER-1000 reactor using the weighting factor method
, Article Annals of Nuclear Energy ; Volume 49 , 2012 , Pages 170-181 ; 03064549 (ISSN) ; Pazirandeh, A ; Ghofrani, M. B ; Sadighi, M ; Sharif University of Technology
2012
Abstract
In the calculation of the optimum loading pattern for a nuclear reactor, several parameters, including the multiplication factor maximization, the raising of the desired safety threshold during the cycle length and the flattening of the radial power peaking factor should be taken into account. Our evaluation of numerous probable arrangements within the VVER-1000 reactor core (approx. 6 × 10 14 cases) indicated that a direct search of the optimum loading pattern is not feasible. Hence, in this study, the weighting factor method and some constraints on the arrangement of the fuel assemblies were applied so that the above cases were reduced to only 244 possible arrangements. Then, using a...
Calculation of VVER-1000 reactor scaling factor for inference of core barrel motion
, Article Annals of Nuclear Energy ; Vol. 63 , 2014 , pp. 205-208 ; ISSN: 03064549 ; Vosoughi, N ; Sharif University of Technology
Abstract
To quantify the core barrel motion (CBM) in a pressurized water reactor, a scaling factor can be calculated to convert the Root Mean Square (RMS) value of the ex-core signals (%) to the core barrel motion amplitude (mil) (Thompson et al., 1980). In the current paper, a scaling factor is calculated using the direct and adjoint methods for a typical VVER-1000 reactor. The scaling factor is calculated using the perturbed parameters that result from CBM. The results show that the calculated scaling factors are not the same in one and two-dimensional modeling, and strongly depend on the ex-core detector location. The linearity assumption of relative detector response versus the small displacement...
Study of fast transient pressure drop in vver-1000 nuclear reactor using acoustic phenomenon
, Article Science and Technology of Nuclear Installations ; Volume 2018 , 2018 ; 16876075 (ISSN) ; Rahgoshay, M ; Vosoughi, N ; Athari Allaf, M ; Sharif University of Technology
Hindawi Limited
2018
Abstract
This article aims to simulate the sudden and fast pressure drop of VVER-1000 reactor core coolant, regarding acoustic phenomenon. It is used to acquire a more accurate method in order to simulate the various accidents of reactor core. Neutronic equations should be solved concurrently by means of DRAGON 4 and DONJON 4 coupling codes. The results of the developed package are compared with WIMS/CITATION and final safety analysis report of Bushehr VVER-1000 reactor (FSAR). Afterwards, time dependent thermal-hydraulic equations are answered by employing Single Heated Channel by Sectionalized Compressible Fluid method. Then, the obtained results were validated by the same transient simulation in a...
Calculation of the fuel composition and the thermo-neutronic parameters of the Bushehr's VVER-1000 reactor during the initial startup and the first cycle using the WIMSD5-B, CITATION-LDI2 and WERL codes
, Article Annals of Nuclear Energy ; Volume 57 , 2013 , Pages 68-83 ; 03064549 (ISSN) ; Pazirandeh, A ; Ghofrani, M. B ; Sadighi, M ; Sharif University of Technology
2013
Abstract
In this paper, the concentrations of fission products and fuel isotopes as well as the changes of the thermo-neutronic parameters of the Bushehr's VVER-1000 reactor were studied during the initial startup and the first cycle. In order to perform the time-dependent cell calculations and obtain the concentration of fuel elements, the WIMSD5-B code was used. Besides, by utilizing the CITATION-LDI2 code, the effective multiplication factor and the thermal power distribution of the reactor were calculated. A computer program (WERL code) was designed in order to perform accurate calculation of the temperature distribution of the reactor core. For this purpose, the Ross-Stoute, Weisman, and...
Development of a VVER-1000 core loading pattern optimization program based on perturbation theory
, Article Annals of Nuclear Energy ; Volume 39, Issue 1 , 2012 , Pages 35-41 ; 03064549 (ISSN) ; Vosoughi, N ; Sharif University of Technology
Abstract
In-core nuclear fuel management is one of the most important concerns in the design of nuclear reactors. Two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor in order to extract the maximum energy, and keeping the local power peaking factor lower than a predetermined value to maintain the fuel integrity. Because of the numerous possible patterns of fuel assemblies in the reactor core, finding the best configuration is so important and challenging. Different techniques for optimization of fuel loading pattern in the reactor core have been introduced by now. In this study, a software is programmed in C# language to find an...
Evaluating the effect of using different sets of enrichment for FAs on fuel management optimization using CA
, Article Annals of Nuclear Energy ; Volume 38, Issue 4 , 2011 , Pages 835-845 ; 03064549 (ISSN) ; Fadaei, A. H ; Zahedi, E ; Sharif University of Technology
Abstract
In nuclear reactor core design, achieving the optimized arrangement of fuel assemblies (FAs) is the most important step towards satisfying safety and economic requirements. In most studies, nuclear fuel optimizations have been performed by using a finite number of different types of FAs. However the effect of FA numbers with different enrichments and the difference between their maximum and minimum enrichment values can be important and should be evaluated in the optimization process. This research is aimed at evaluating the effect of using different enrichment values for FAs. This issue has been investigated by focusing on two parameters, namely, the initially selected enrichment and the...
Localization of a noise source in VVER-1000 reactor core using neutron noise analysis methods
, Article International Conference on Nuclear Engineering, Proceedings, ICONE, 17 May 2010 through 21 May 2010 ; Volume 2 , May , 2010 ; 9780791849309 (ISBN) ; Vosoughi, N ; Zahedinejad, E ; Nuclear Engineering Division ; Sharif University of Technology
2010
Abstract
In this paper, localization of a noise source from limited neutron detectors sparsely distributed throughout the core of a typical VVER-1000 reactor is investigated. For this purpose, developing a 2-D neutron noise simulator for hexagonal geometries based on the 2-group diffusion approximation, the reactor dynamic transfer function is calculated. The boxscheme finite difference method is first developed for hexagonal geometries, to be used for spatial discretisation of both 2-D 2-group static and noise diffusion equations. The dynamic state is assumed in the frequency domain which leads to discarding of the time disrcetisation. The developed 2-D 2- group neutron noise simulator calculates...
Design and analysis of a thermal hydraulic integral test facility for bushehr nuclear power plant
, Article Progress in Nuclear Energy ; Volume 51, Issue 3 , 2009 , Pages 443-450 ; 01491970 (ISSN) ; Jafari, J ; Sohrabpour, M ; Sharif University of Technology
2009
Abstract
In this paper, design and analysis of a thermal hydraulic integral test facility for Bushehr Nuclear Power Plant (NPP) is presented. The Bushehr Integral Test Facility (BITF) is a test facility designed to model the thermal-hydraulic behaviours of the Bushehr NPP (VVER-1000) pressurized water reactors currently in use in IRAN. These reactors have unique features that differ from other PWR designs. The BITF simulates the major components and systems of the reference NPP, making it possible to examine postulated small and medium break a loss of coolant accidents (LOCAs) and operational transients. The BITF is a volume-scaled model (1:1375). To ensure that gravitational forces remain equal to...
Fuel management optimization based on power profile by Cellular Automata
, Article Annals of Nuclear Energy ; Volume 37, Issue 12 , 2010 , Pages 1712-1722 ; 03064549 (ISSN) ; Moghaddam, N. M ; Zahedinejad, E ; Fadaei, M. M ; Kia, S ; Sharif University of Technology
Abstract
Fuel management in PWR nuclear reactors is comprised of a collection of principles and practices required for the planning, scheduling, refueling, and safe operation of nuclear power plants to minimize the total plant and system energy costs to the extent possible. Despite remarkable advancements in optimization procedures, inherent complexities in nuclear reactor structure and strong inter-dependency among the fundamental parameters of the core make it necessary to evaluate the most efficient arrangement of the core. Several patterns have been presented so far to determine the best configuration of fuels in the reactor core by emphasis on minimizing the local power peaking factor (Pq). In...
Three-dimensional simulation of turbulent flow in 3-sub channels of a VVER-1000 reactor
, Article Scientia Iranica ; Volume 17, Issue 2 B , 2010 , Pages 83-92 ; 10263098 (ISSN) ; Firoozabadi, B ; Sharif University of Technology
2010
Abstract
In this study, the fluid dynamics and convective heat transfer for turbulent flows through a 3-sub channel of a rod bundle, which is representative of those used in VVER-1000, are examined. The rod bundle is constructed from parallel rods in a hexagonal array. The rods are on constant pitch by spacer grids spaced axially along the rod bundle. The geometry details of the bundle and heat flux from the fuel rod are similar to that of the Iranian nuclear reactor under construction. A numerical study using Computational Fluid Dynamics (CFD) was carried out to estimate the flow field, pressure loss and heat transfer coefficients in spacer grids and rod bundles. Turbulence has been modeled using...