Sharif Digital Repository

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... 

Several parameters, one of which is the Dancoff Factor (DF), are used to calculate the resonance integral (RI) and reduced flux in the resonance region of heterogeneous systems as well as to accurately determine the group constants for criticality calculations. This paper is a report on the development of a program to calculate the DF correction factor using Monte Carlo method and collision probability definition in three-dimensional (3-D) geometries and with multi energy groups. Hence, the DF for any arbitrary arrangement of cylindrical and slab fuel elements is hereby calculated. The fuel elements are monitored and kept at equal levels, though different material compositions and formations... 

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 the present study, the neutron noise, i.e. The stationary fluctuation of the neutron flux around its mean value, is calculated in 2-group forward and adjoint diffusion theory for both hexagonal and rectangular reactor cores. To this end, the static neutron calculation is performed at the first stage. The spatial discretization of equations is based on linear approximation of Galerkin Finite Element Method (GFEM) using unstructured triangle elements. Using power iteration method, forward and adjoint fluxes with the corresponding eigenvalues are obtained. The results are then benchmarked against the valid results for BIBLIS-2D and IAEA-2D benchmark problems and DONJON computer code. The...