Feasibility Study on MCNPX-based Burnup Calculation for CANDU Reactor Analysis

In recent years, particle transport codes based on Monte Carlo method have been widely used for analysis of various reactors such as PWR, FBR, and LMR, etc. Especially, the MCNP code, which is developed in Los Alamos National Laboratory (LANL), is a very powerful in solving gamma, neutron, and electron transport equation, and provides exact solutions (i.e., particle flux and dose rate) for a wide range of energy level and for various geometries. Besides, one of the latest extension versions of the MCNPX code (i.e., MCNPX 2.6.0) was integrated with CINDER90 depletion code. Hence, burnup can be calculated for the various combinations of materials defined on the user input, as well as for each individual material containing a fissile actinide. However, the neutronic analysis of the domestic CANDU reactors has been performed only by using the deterministic codes, such as WIMS, Dragon and RFSP. In this study, therefore, the performances of the MCNPX 2.6.0 code were evaluated to investigate the suitability and possibility of MCNPX-based burnup calculation in CANDU reactor. For this reason, the multiplication factor and physical parameters (η, f, p, and ε) as a function of burnup were calculated for an infinite lattice of natural uranium fuel. The results were then compared with those from the WIMS-IST code which is a multi-group transport code for lattice calculations