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... 

Thermal neutron analysis (TNA) technology has been used for the non-destructive detection of explosives. The system uses a relatively weak 252Cf neutron source (1.03.107 n/s) and two 3″3×″ NaI(Tl) detectors. The presence of explosives is confirmed via detection of the 10.83 MeV prompt gamma-ray associated with nitrogen decay. The MCNP4A code was used to simulate the neutron and gamma transport through the system. The thermal neutron flux in the activation position was measured using gold and indium foils. The measured thermal neutron flux was lower, by not more than 9.5%, than that of simulation. In this report the results of the preliminary tests on the system are described  

Knowledge of the complete multiaxial residual stress distribution in engineering components is essential for assessing their integrity. Often, however, only limited measurements are made. Here, an analysis is presented for determining the multiaxial distribution from a limited set of measurements. These measurements are used with an assumed plastic strain distribution. Residual stress measurements were made on hot forged and shot blasted steel bars using X-ray and neutron diffraction techniques. The residual stresses were measured on the surface and at selected interior points of the specimens. The predicted multiaxial distributions were compared with experimental measurements obtained using... 

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  

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... 

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... 

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 study, the SD-HACNEM (Sharif Dynamic - High order Average Current Nodal Expansion Method) neutron noise simulator in two energy groups using a second-order flux expansion method for two-dimensional rectangular X Y-geometry has been developed. In the first step, the calculations were performed for the steady state and results of ACNEM (Average Current Nodal Expansion Method) and HACNEM (High order Average Current Nodal Expansion Method) were examined and compared. To solve the problem, the power iteration algorithm has been used to calculate the distribution of neutron flux and neutron multiplication factor by considering the coarse-mesh (each fuel assembly one node). To validate 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... 

This paper presents a bit-flip tolerance in SRAM-based FPGAs which suffers from high energy particles, alpha and neutrons in the atmosphere. For each of protections, the applicability, efficiency and implementation issues are discussed. Moreover, the area, the power and the protection capability of the methods are mentioned and compared with previous work Based on the results of experiments and their analysis, one method is selected as best one. The selected method is much better than previous work e.g., duplication with comparison, triple modular redundancy which impose two and three area and power overheads, respectively. © 2007 IEEE  

In this paper, evaluation of the kinetic parameters (effective delayed neutron fraction and prompt neutron lifetime) in PWRs, using static cell and core calculation codes, is reported. A new software has been developed to link the WIMS, BORGES and CITATION codes in Visual C# computer programming language. Using the WIMS cell calculation code, multigroup microscopic cross-sections and number densities of different materials can be generated in a binary file. By the use of BORGES code, these binary-form cross-sections and number densities are converted to a format readable by the CITATION core calculation code, by which the kinetic parameters can be finally obtained. This software is used for... 

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... 

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 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...