Sharif Digital Repository

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... 

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... 

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 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... 

Accurate calculation of kinetic parameters is of utmost importance in the safety analysis of a nuclear reactor. In the current paper, two approaches are investigated to evaluate these parameters in energy phase space. In the first approach, these parameters are derived from an energy-continuous form of the forward and adjoint transport equations and then integrals with respect to the energy variable are replaced by weighted summations over the energy groups, while in the second approach these parameters are extracted from the multi-group forward equation and its associate adjoint equation in which their multigroup constants are weighted by forward spectrum. The difference of weighting... 

In this research, effective delayed neutron fraction (βeff) and neutron generation time (A) of the Tehran Research Reactor (TRR) are calculated for different uranium enrichments from 14.84 w/o to 96.56 w/o U 235 in two states of the TRR, (cold fuel, clad and coolant temperature of 20°C; and hot fuel, clad and coolant temperature of 65, 49 and 44°C, respectively) using the MTR-PC computer code. Comparative analysis shows that both βeff and Λ increase as fuel enrichment decreases. However, variation rate βeff is not the same in two conditions. βeff in the hot state is larger than those calculated in the cold state when fuel enrichment goes more than 83.91%, while the situation is reverse for... 

Effective delayed neutron fraction βeff and neutron generation time Λare important factors in reactor physics calculation and transient analysis. In first stage of this research, these kinetics parameters have been calculated for two states of Tehran research reactor (TRR), i.e. cold (fuel, clad and coolant temperature 20°C) and hot (fuel, clad and coolant temperature 65, 49 and 44°C, respectively) using MTR-PC code. In second stage, these parameters have been measured with experimental method based on Inhour equation. For cold state, calculated βeff and Λ by MTR-PC are 0.008315 and 30.190 μsec, respectively. Same parameters in hot state are 0.008303 and 33.828 μsec, respectively. The... 

In this research, the characteristics of the prepared samples in epoxy matrix by means of X-ray diffraction (XRD), energy dispersive X-ray spectroscopy (EDS), as well as scanning electron microscope (SEM) are evaluated. Meanwhile, the obtained mechanical properties of the specimen are investigated. Thermogravimetric analysis (TGA) is also employed to evaluate the thermal degradation of manufactured nanocomposites. The thermal neutron absorption properties of nanocomposites containing 3 wt% of montmorillonite nanoclay (closite30B) have been studied experimentally, using an Am-Be point source. Mechanical tests reveal that the higher B4C concentrations, the more tensile strengths, but lower... 

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... 

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... 

The neutron noise is defined as the stationary fluctuation of the neutron flux around its mean value due to the induced perturbation in the reactor core. The neutron noise analysis may be useful in many applications like noise source reconstruction. To identify the noise source, calculated neutron noise distribution of the detectors is used as input data by the considered unfolding algorithm. The neutron noise distribution of the VVER-1000 reactor core is calculated using the developed computational code based on Galerkin Finite Element Method (GFEM). The noise source of type absorber of variable strength is considered in the calculation. The computational code developed based on An Adaptive... 

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... 

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... 

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... 

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... 

Reactivity Controlled Compression Ignition can be extended over a wide spectrum of fuels and is anticipated as a promising strategy in meeting current and future emission regulations. In this study, the effect of n-butanol addition on combustion characteristics and emissions in a reactivity controlled engine was investigated experimentally. Different ratios of butanol-diesel blends at different settings of EGR and premixed ratios were applied to a light duty diesel engine. The butanol-diesel blends were directly injected into the combustion chamber while gasoline was injected at the intake port. Combustion phasing was maintained at 2.7 °CA for all of test points by adjusting fuel injection... 

In this paper, the neutron noise based on transport theory and diffusion noise theory using Green's function technique is calculated. As the neutron noise is used for core diagnostic, surveillance and monitoring, calculation of neutron noise precisely can play an important role in monitoring and safety. We compare the accuracy of Green's function based on transport and diffusion theory in order to survey the differences between these theories. In this study some deviation between results obtained two theories are observed, and the impact of dimensions, cross sections and frequency on the results investigated. © 2017  

This paper presents a comparative study of various soft computing algorithms for reconstruction of neutron noise sources in the nuclear reactor cores. To this end, the computational code for reconstruction of neutron noise source is developed based on the Adaptive Neuro-Fuzzy Inference System (ANFIS), Decision Tree (DT), Radial Basis Function (RBF) and Support Vector Machine (SVM) algorithms. Neutron noise source reconstruction process using the developed computational code consists of three stages of training, testing and validation. The information of neutron noise sources and induced neutron noise distributions are used as output and input data of training stage, respectively. As input... 

In the present study, the development of a neutron noise simulator, DYN-ACNEM, using the Average Current Nodal Expansion Method (ACNEM) in 2-G, 2-D hexagonal geometries is reported. In first stage, the static neutron calculation is performed. The neutron/adjoint flux distribution and corresponding eigen-values are calculated using the algorithm developed based on power iteration method by considering the coarse meshes. The results of the static calculation are validated against the well-known IAEA-2D benchmark problem. In the second stage, the dynamic calculation is performed in the frequency domain in which the dimension of the variable space of the noise equations is lower than the time... 

Purpose: The purpose of this paper is to investigate the impacts of electric vehicles' (EVs) charging/discharging decisions in energy resources scheduling problem of active distribution networks. Design/methodology/approach: The problem under study is modelled as a two-stage optimisation problem in which the main requirements of EV owners are introduced as an objective function of the first stage. The total energy costs and the emission factor are considered as the main criteria of the second stage. The output generation schedules of distributed generation (DG) technologies together with the charging/discharging schedule of EVs are proposed as decision variables of the energy scheduling...