Multi-Group Neutron Cross Sections and Scattering Matrix Calculations Via Monte Carlo Method

Abstract

This study describes a new methodology to estimate multi-group neutron cross sections and scattering matrix elements from a Monte Carlo simulation, particularly from MCNPX 2.7 code. The geometric flexibility associated with the Monte Carlo method makes it very suitable for the generation of highly accurate multi-group constants. While the deterministic lattice codes are not capable of dealing with energy as a continuous variable the Monte Carlo codes such as MCNPX make use of a continuous energy cross sections data for neutron transport calculations. To determine the scattering matrix, an output file of a MCNPX run (socalled PTRAC file) with all relevant source, collision and terminal events of the simulation is used. First, by a separately special program in MATLAB, the PTRAC file is read and tracks are selected in the geometrical region for which one wants to calculate the multi-group constants. Then, information such as coordinates of the particle position, particle direction with coordinates axes, energy and weight of the particle are extracted. The multi-group scattering matrix elements are generated via the weight-to-flux ratio method using the above data available in the PTRAC file. The proposed method is validated using three benchmark problems involve a slab, a pin cell, and a fuel assembly of Tehran research reactor (TRR). The generated multi-group constants via presented method and multiplication factor calculation presents a good comparison to the reference values