Sharif Digital Repository

Noise analysis techniques including Feynman-α (variance-to-mean) and Rossi-α (correlation) have been simulated by MCNP computer code to calculate the prompt neutron decay constant (α0), effective delayed neutron fraction (βeff) and neutron generation time (Λ) in a subcritical condition for the first operating core configuration of Tehran Research Reactor (TRR). The reactor core is considered to be in zero power (reactor power is less than 1 W) in the entire simulation process. The effect of some key parameters such as detector efficiency, detector position and its dead time on the results of simulation has been discussed as well. The results of proposed method in the current study are... 

Based on the determination of protons fluence at the phantom's surface, a 3D dose distribution is calculated inside a water phantom using a fast method. The dose contribution of secondary particles, originating from inelastic nuclear interactions, is also taken into account. This is achieved by assuming that 60% of the energy transferred to secondary particles is locally absorbed. Secondary radiation delivers approximately 16.8% of the total dose in the plateau region of the Bragg curve for monoenergetic protons of energy 190 MeV. The physical dose beyond the Bragg peak is obtained for a proton beam of 190 MeV using a Geant4 simulation. It is found that the dose beyond the Bragg peak is <... 

Three distinct methods, namely, the spectral density, the multifractal random walk approach, and the multifractal detrended fluctuation analysis are utilized to study the properties of four distinct types of well logs from three oil and gas fields, namely, the natural gamma ray emission, neutron porosity, bulk density, and the sonic transient time logs. Such well logs have never been analyzed by the methods that we utilize in the present study. The results indicate that the well logs exhibit multifractal characteristics, and the estimated Hurst exponents by the three methods are close to each other. Using multifractal detrended fluctuation analysis and the shuffled and surrogated data, we... 

The time-dependent diffusion of neutrons in a spheroid as a function of the focal distance has been studied. The solution is based on an orthogonal basis and an extrapolation distanced related boundary condition for the spheroidal geometry. It has been shown that spheres and disks are two limiting cases for the spheroids, for which there is a smooth transition for the systems properties between these two limits. Furthermore, it is demonstrated that a slight deformation from a sphere does not affect the fundamental mode properties, to the first order. The calculations for both multiplying and non-multiplying media have been undertaken, showing good agreement with direct Monte Carlo... 

The High-Performance Light Water Reactor (HPLWR) is the European version of the advance nuclear energy power plant at Supercritical-pressure. A light water reactor at supercritical pressure, being currently under design, is the new generation of nuclear reactors. The aim of this study is to predict the HPLWR neutronic behavior of the axial enrichment profile with an average enrichment of 5 w/o U-235. Neutronic calculations are performed using WIMS and CITATION codes. Changes in neutronic parameter, such as Power Peaking Factor (PPF) are discussed in this paper  

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... 

Abstract In this paper, the solution of multi-group neutron/adjoint equation using Finite Element Method (FEM) for hexagonal and rectangular reactor cores is reported. The spatial discretization of the neutron diffusion equation is performed based on two different Finite Element Methods (FEMs) using unstructured triangular elements generated by Gambit software. Calculations are performed using Galerkin and Generalized Least Squares FEMs; based on which results are compared. Using the power iteration method for the neutron and adjoint calculations, the neutron and adjoint flux distributions with the corresponding eigenvalues are obtained. The results are then validated against the valid... 

In this paper, the development of the MCNPX-ESUT (MCNPX-Energy Engineering of Sharif University of Technology) computer code for simulation of neutron/gamma pulse height distribution is reported. Since liquid organic scintillators like NE-213 are well suited and routinely used for spectrometry in mixed neutron/gamma fields, this type of detectors is selected for simulation in the present study. The proposed algorithm for simulation includes four main steps. The first step is the modeling of the neutron/gamma particle transport and their interactions with the materials in the environment and detector volume. In the second step, the number of scintillation photons due to charged particles such... 

Neutron noise induced by propagating disturbances in VVER-type reactor core is addressed in this paper. The spatial discretization of the governing equations is based on the box-scheme finite difference method for triangular-z geometry. Using the derived equations, a 3-D 2-group neutron noise simulator (called TRIDYN-3) is developed for hexagonal-structured reactor core, by which the discrete form of both the forward and adjoint reactor dynamic transfer functions (in the frequency domain) can be calculated. In addition, both types of noise sources, namely point-like and traveling perturbations, can be modeled by TRIDYN-3. The results are then benchmarked in different cases. Considering the... 

Neutron noise induced by propagating thermal-hydraulic disturbances (propagation noise for short) in pressurized water reactors is investigated in this paper. A closed-loop neutron-kinetic/thermal-hydraulic noise simulator (named NOISIM) has been developed, with the capability of modeling the propagation noise in both Western-type and VVER-type pressurized water reactors. The neutron-kinetic/thermal-hydraulic noise equations are on the basis of the first-order perturbation theory. The spatial discretization among the neutron-kinetic noise equations is based on the box-scheme finite difference method (BSFDM) for rectangular-z, triangular-z and hexagonal-z geometries. Furthermore, the finite... 

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... 

In this research, new software package for neutronic calculations, especially kinetic parameters of PWR reactors, has been developed. The program used to link the WIMS-D5, BORGES and CITVAP nuclear codes has been written in Visual C# programming language. This software was used for calculation of kinetic parameters of WER-1000 and NOK Beznau reaction The ration (βeff)i/(βeff)corewhich are an important input data for the reactivity accident analysis, were also calculated. The results were compared with final safety analysis report (FSAR) and published documents. Copyright  

Effective delayed neutron fraction βeff and neutron generation time Λ are important factors in reactor physics calculation and transient analysis. In the 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) states using MTR-PC computer code. The ratio of (βeff) i/(βeff)core plays an important role in reactivity accident analysis codes. This parameter and its contribution to effective delayed neutron fraction from each nucleus have been calculated in cold and hot reactor states. Uncertainty of... 

The accuracy of static neutronic parameters in the nuclear reactors depends upon the determination of group constants of the diffusion equation in the desired geometry. Although several methods have been proposed for calculating these parameters, there is still the need for more reliable methods. In this paper a powerful and innovative method based on Spatial Homogenization and Temperature Variation (SHTV) of physical properties of a WWER-1000 nuclear reactor core for calculating the relative power distribution of Fuel Assemblies (FA) and the hot fuel rod, is presented. The method is based on replacing the heterogeneous lattices of materials with different properties by an equivalent... 

In this paper, the development of a neutron noise simulator for hexagonal-structured reactor cores using both the forward and the adjoint methods is reported. The spatial discretisation of both 2-D 2-group static and dynamic equations is based on a developed box-scheme finite difference method for hexagonal mesh boxes. Using the power iteration method for the static calculations, the 2-group neutron flux and its adjoint with the corresponding eigenvalues are obtained by the developed static simulator. The results are then benchmarked against the well-known CITATION computer code. The dynamic calculations are performed in the frequency domain which leads to discarding of the time... 

The analytical evaluation of the capability of Bremsstrahlung highly directional energetic γ-beam to induce photo transmutation of 90Sr (γ,n) 89Sr is presented. Photo transmutation of hazardous nuclear waste of 90Sr, one of the two main sources of heat and radioactivity in spent fuel into valuable nuclear medicine radioisotope of 89Sr is explained. Based on the calculations, a fairly decent fraction of gamma rays in this range are used in transmuting of 90Sr into 89Sr where according to the available experimental data it is shown that by irradiating a 1-cm thick 90Sr sample with lasers of intensity of 1021 W/cm2 and repletion rate of 100Hz for an hour, the reaction activity would be 1.45... 

Effective delayed neutron fraction βeff and neutron generation time Λ are important factors in reactor physics calculation and transient analysis. In the first stage of this research, these kinetic parameters have been calculated for two states of Tehran Research Reactor (TRR), i.e. cold (fuel, clad and coolant temperatures equal to 20 °C) and hot (fuel, clad and coolant temperatures of 65, 49 and 44 °C, respectively) states using MTR_PC code. In the second stage, these parameters have been measured with an experimental method based on Inhour equation. For the cold state, calculated βeff and Λ by MTR_PC code are 0.008315 and 30.190 μs, respectively. In the hot state, these parameters being... 

Obtaining a reliable reservoir permeability map that is consistent with all available data is of great important for reservoir engineers. However, there is not enough core from existing wells to estimate the reservoir permeability directly but Well log data are more widely available. This study aims to model permeability within the reservoir while there is no enough data. In particular, we use artificial neural networks to estimate permeability using four input logs of sonic, gravity, porosity and neutron logs in six existing wells. In order to eliminate the correlated data, we have done Principal Component Analysis on selected input logs. Collocated cokriging is considered as a valuable... 

In the present study, the neutron noise, i.e. the stationary fluctuation of the neutron flux around its mean value is calculated based on the 2G, 3D neutron diffusion theory. To this end, the static neutron calculation is performed at the first stage. The spatial discretization of the neutron diffusion equation is performed based on linear approximation of Galerkin Finite Element Method (GFEM) using unstructured tetrahedron elements. Using power iteration method, neutron flux and corresponding eigen-value are obtained. The results are then benchmarked against the valid results for VVER-1000 (3D) benchmark problem. In the second stage, the neutron noise equation is solved using GFEM and... 

We carry out a series of cross-correlation analysis of raw well-log data, in order to study the possible connection between natural gamma ray (GR) logs and other types of well logs, such as neutron porosity (NPHI), sonic transient time (denoted usually by DT), and bulk density (RHOB) of oil and gas reservoirs. Three distinct, but complementary, methods are used to analyze the cross correlations, namely, the multifractal detrended cross-correlation analysis (MF-DXA), the so-called Qcc(m) test in conjunction with the statistical test-the χ2(m) distribution-and the cross-wavelet transform (XWT) and wavelet coherency. The Qcc(m) test and MF-DXA are used to identify and quantify the strength of...