Sharif Digital Repository

The objective of this paper is to introduce a new numerical method for neutronic calculation in a reactor core. This method can produce the final finite form of the neutron diffusion equation by classifying the neutronic variables and using two kinds of cell complexes without starting from the conventional differential form of the neutron diffusion equation. The method with linear interpolation produces the same convergence as the linear continuous finite element method. The quadratic interpolation is proven; the convergence order depends on the shape of the dual cell. The maximum convergence order is achieved by choosing the dual cell based on two Gauss' points. The accuracy of the method... 

The Lateral deformation of Fuel Assembly (FA) under the operational conditions of the reactor cores is called FA bowing. This phenomenon is caused by factors such as thermal and hydraulic loads on FAs in the reactor core. It can lead to disturbances in the movement of control rods inside of FAs, cross-contact of FAs in refueling, and also changes in power distribution. Changing the distance between the fuels along the assemblies due to bowing, leads to non-uniform distribution of water (coolant) around the FAs and results in neutronic perturbation. In this research, by developing a calculation model for the bowed FAs of the VVER-1000 reactor, based on the distribution of water around FA at... 

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 highly reliable tool for transient simulation is vital in the safety analysis of a nuclear reactor. Despite this fact most tools still use diffusion theory and point-kinetics that involve many approximation such as discretization in space, energy, angle and time. However, Monte Carlo method inherently overcomes these restrictions and provides an appropriate foundation to accurately calculate the parameters of a reactor. In this paper fundamental parameters like multiplication factor (K eff ) and mean generation time (t G ) are calculated using Monte Carlo method and then employed in transient analysis for computing the neutron population, proportional to K eff , during a generation time... 

Transient analysis which is vital in safety analysis requires a reliable calculation method. Most valid tools use diffusion theory with many approximations by now. However, the Monte Carlo method inherently overcomes these approximations and accurately calculates the parameters of a reactor. In this paper, a new time-dependent transport approach is described to simulate the nuclear reactor dynamic correctly using the MCNPX code. In this approach the fundamental parameters of a nuclear reactor like multiplication factor (K eff ) and mean generation time (t G ) are calculated using MCNPX code. They are then employed in the formulas to compute neutron population, proportional to K eff , during... 

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.... 

In this paper, a simpler approach compared to the existing approaches is developed to analyze nuclear reactor dynamics based on the explicit Monte Carlo method. A new population control method is also introduced to prevent neutron population growth and consequent computer memory shortages, which also increases simulation speed. The scheme is applied for time-dependent particle tracking in three-dimensional arbitrary geometries in the presence of feedbacks through a code named MCSP-Explicit. Changes in material density, as well as geometry dimensions, are also considered during simulation. MCSP-Explicit can be run with either continuous or multigroup data libraries, and it is further boosted... 

In this paper, a simpler approach compared to the existing approaches is developed to analyze nuclear reactor dynamics based on the explicit Monte Carlo method. A new population control method is also introduced to prevent neutron population growth and consequent computer memory shortages, which also increases simulation speed. The scheme is applied for time-dependent particle tracking in three-dimensional arbitrary geometries in the presence of feedbacks through a code named MCSP-Explicit. Changes in material density, as well as geometry dimensions, are also considered during simulation. MCSP-Explicit can be run with either continuous or multigroup data libraries, and it is further boosted... 

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... 

The objective of this paper is to introduce a new direct method for neutronic calculations. This method, called Direct Discrete Method (DDM), is simpler than the Neutron Transport Equation and more compatible with the physical meanings of the problem. The method, based on the physics of the problem, initially-runs through meshing of the desired geometry. Next, the balance equation for each mesh interval is written. Considering the connection between the mesh intervals, the final discrete equation series are directly obtained without the need to first pass through the set-up of the neutron transport differential equation. In this paper, a single and multigroup neutron transport discrete... 

To solve neutron transport equation in multigroup approach, in addition to weighting function and number of energy groups, proper selection of the group boundaries have high importance for the accuracy of the calculations. In the current paper, the bilinear combination of forward and adjoint neutron spectra is used for the optimization of 69 energy group structure of WIMSD5 lattice physics code. To remedy the energy self-shielding effect, homogeneous adjoint and forward BN equations on an ultrafine energy group structure have been solved to obtain the ultrafine forward and adjoint spectra. The coarse group intervals are selected to have equal values of bilinear function in each... 

Objective: To develop and validate a versatile Monte Carlo (MC)-based dose calculation engine to support MC-based dose verification of treatment planning systems (TPSs) and quality assurance (QA) workflows in proton therapy. Methods: The GATE MC toolkit was used to simulate a fixed horizontal active scan-based proton beam delivery (SIEMENS IONTRIS). Within the nozzle, two primary and secondary dose monitors have been designed to enable the comparison of the accuracy of dose estimation from MC simulations with respect to physical QA measurements. The developed beam model was validated against a series of commissioning measurements using pinpoint chambers and 2D array ionization chambers (IC)... 

The present research aims at depositing molybdenum on stainless steel substrate using dense plasma focus (DPF) device. The varying parameter was the angular position of substrate surface from the anode tip. The deposited layers were characterized using various analyses such as X-ray diffraction (XRD), scanning electron microscopy (SEM) equipped with energy dispersive X-ray spectroscopy (EDS), atomic force microscopy (AFM), and microhardness measurements. Microstructural observations revealed that the deposited coatings had rough surfaces containing holes as well as a stripe-featured structure covering the whole surface including the interior regions of the holes and the regions among them.... 

Inferior temporal (IT) cortex is the most important part of the brain and plays an important role in response to visual stimuli. In this study, object decoding has been performed using neuron spikes in IT cortex region. Single Unit Activity (SUA) was recorded from 123 neurons in IT cortex. Pseudo-population firing rate vectors were created, then dimension reduction was done and an Artificial Neural Network (ANN) was used for object decoding. Object decoding accuracy was calculated for various window lengths from 50 ms to 200 ms and various window steps from 25 ms to 100 ms. The results show that 150 ms length and 50 ms window step size gives an optimum performance in average accuracy  

Inferior temporal (IT) cortex is the most important part of the brain and plays an important role in response to visual stimuli. In this study, object decoding has been performed using neuron spikes in IT cortex region. Single Unit Activity (SUA) was recorded from 123 neurons in IT cortex. Pseudo-population firing rate vectors were created, then dimension reduction was done and an Artificial Neural Network (ANN) was used for object decoding. Object decoding accuracy was calculated for various window lengths from 50 ms to 200 ms and various window steps from 25 ms to 100 ms. The results show that 150 ms length and 50 ms window step size gives an optimum performance in average accuracy  

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... 

Background: in this study, two proton beam delivery designs, i.e. passive scattering proton therapy (PSPT) and pencil beam scanning (PBS), were quantitatively compared in terms of dosimetric indices. The GATE Monte Carlo (MC) particle transport code was used to simulate the proton beam system; and the developed simulation engines were benchmarked with respect to the experimental measurements. Method: A water phantom was used to simulate system energy parameters using a set of depth-dose data in the energy range of 120-235 MeV. To compare the performance of PSPT against PBS, multiple dosimetric parameters including Bragg peak width (BP W50), peak position, range, peak-To-entrance dose ratio,... 

Block method (BM) is a simple and fast method to solve inverse and forward problems in Electrical Impedance Tomography (EIT). In BM, at first tissue is modeled by some blocks and it is assumed that each block has a specific conductivity. Then a medical image is constructed by calculation of its conductivity. Recently, a non-iterative linear inverse solution is presented for block method which we name 2D BM. In this paper, an efficient algorithm with new formulation is proposed to improve the 2D BM, and then several examples have been investigated to examine the proposed method. Results show that suggested algorithm achieves better outcomes in all situations, although its run time is...