Sharif Digital Repository

In pebble bed reactors, the core is filled with thousands of graphite and fuel pebbles. Fuel pebbles in these reactors consist of TRISO particles, which are embedded in a graphite matrix stochastically. The reactor core is also stochastically filled with pebbles. These two stochastic geometries comprise the so-called double heterogeneous nature of this type of reactor. In this paper, a pebble bed reactor, the HTR-10, is used to demonstrate a treatment of this double heterogeneity using the MCNP5 Monte Carlo code and MATLAB programming. In this technique, TRISO particles are modeled in a pebble using the expanded FILL and LATTICE features of MCNP5. MATLAB is used to generate random numbers... 

Applying the available neutron flux for medical and industrial purposes is the most important application of research reactors. The neutron radiography system is used for non-destructive testing (NDT) of materials so that it is one of the main applications of nuclear research reactors. One of these research reactors is the 5 MW pool-type light water research reactor of Tehran (TRR). This work aims to investigate on materials and location of the beam tube (BT) of the TRR radiography system to improve the index parameters of BT. Our results showed that a through-type BT with 20 cm thick carbon neutron filter, 1.2 cm and 9.4 cm of the diameter of inlet (D1) and output (D2) BT, respectively... 

An embedded time interval data acquisition system (DAS) is developed for zero power reactor (ZPR) noise experiments. The system is capable of measuring the correlation or probability distribution of a random process. The design is totally implemented on a single Field Programmable Gate Array (FPGA). The architecture is tested on different FPGA platforms with different speed grades and hardware resources. Generic experimental values for time resolution and inter-event dead time of the system are 2.22 ns and 6.67 ns respectively. The DAS can record around 48-bit × 790 kS/s utilizing its built-in fast memory. The system can measure very long time intervals due to its 48-bit timing structure... 

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... 

In this study, new method for NAA purposes at 30 kW Isfahan MNSR is suggested. An algorithm based on ANN is proposed to quantitatively predict the unknown elements with no need standard sample. A three-layer feed-forward ANN with back-propagation algorithm has been used to determine concentration of selenium and fluorine in Multiple Sclerosis patients and healthy people blood samples. Predicted concentration of elements show good agreement between new method and experiment results. The correlation coefficient between the experimentally determined and predicted values are 0.99104 and 0.99364, respectively. This method is a rapid and precise approach for elemental analysis. © 2022, Akadémiai... 

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 stochastic behavior of a point reactor is modeled with a system of Ito stochastic differential equations. This new approach does not require computing the square root of a matrix which is a great computational advantage. Moreover, the derivation procedure clearly demonstrates the mathematical approximations involved in the final formulation. Three numerical benchmarks show the accuracy of this model in predicting the mean and variance of the neutron and precursor population in a point reactor  

An improvement to the Direct Discrete Method (DDM), also known as the Cell Method, has been discussed. The improvement is based on a duality theorem between the primal and dual complexes. Also, the analog counterpart of the Integral operator has been derived in this paper. The multi-group neutron diffusion is then derived, directly in a discrete algebraic form, according to this procedure. A numerical example has shown that this method would yield a high order of convergence (approximately 4.6) if its parameters are adjusted suitably. Finally, the method is applied to the 2D IAEA benchmark problem, and has shown to yield accurate solutions with a reasonably low number of unknowns  

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... 

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  

In an attempt to explore the significance of transport theory in neutron noise, localization of a noise source by Green's function based on transport theory is investigated. There are considerable differences between Green's functions based on diffusion and transport, such as small dimensions, near edges, high heterogeneity medium and high-frequency source of perturbation. These differences are expected to significantly impact unfolding, reconstruction, and identification of the neutron noise source. Improvement in noise source unfolding methods is essential in terms of safety aspects and reactor performance enhancement. Since gaining the ability to monitor nuclear reactor based on noise... 

The alkaline treatment condition plays a crucial role in governing the ultimate properties of flax fibers. In this study, flax fibers were modified with mild alkalization and severe mercerization conditions to give fundamental insight into how the molecular-scale changes in cell wall fine structure and cellulose supramolecular structure can affect the macroscopic properties of fibers. SEM, FTIR, XRD, TGA, and DSC techniques were employed to characterize the variations in morphology, composition, crystalline structure, and thermal properties of fibers. Also, tensile tests evaluated their reinforcing performance in polypropylene-based composites. The results indicated that alkalization in 5%... 

Various methods have been used for solving the neutron transport equation in the past, and a number of computer codes have been developed based on these solution methods. This paper describes a novel method for the solution of the steady-state and time-dependent neutron transport equation using the duality between neutronic parameters in the method of characteristic (MOC) and the electrical parameters in the cellular neural networks (CNN). The relevant electrical circuit can be simulated by professional electrical circuit simulator software, HSPICE. This software is used for numerical solution of the transport equation only by preparation of appropriate inputs. This method does not need... 

In this paper, we describe an innovative method to model and solve spatio-temporal behavior of nuclear reactor cores via three-dimensional multilayer cellular neural networks. This method uses electrical elements and the existing duality between neutronic and thermal-hydraulic parameters of nuclear reactors. The relevant electrical circuit can be simulated by existing professional electrical circuit software. This research goes beyond our previous efforts to use a neural computing approach in the nuclear field. Modeling and solving simple nuclear reactor kinetic equations is now expanded to a complete dynamic calculation, integrating the core thermal-hydraulic models and the relevant... 

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... 

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... 

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... 

Residual stresses, created in a steel beam by elastic-plastic bending, are predicted using an approximate analysis and the finite element method. The predictions are compared to experimental measurements obtained from the application of incremental centre hole drilling, deep hole drilling and neutron diffraction methods. Finite element simulations of the incremental centre hole drilling and deep hole drilling methods applied to the predicted residual stresses permitted an assessment of their ability to reconstruct the stresses. An analytical reconstruction analysis using an Airy stress function together with boundary and equilibrium conditions is developed and applied to the predictions and... 

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... 

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...