Monte Carlo with burn-up calculation method for research reactors

A method (MCQ) has been developed by introducing a microscopic burn up scheme which uses the Monte Carlo calculated fluxes and microscopic reaction rates as a basis for solving nuclide material balance equations for each spatial region of the particular nuclear device of interest. Energy dependent cross section libraries and full 3D geometry of the system can be input for the calculations. The resulting predictions at successive burn up time steps are thus based on a calculation route where both geometry and cross sections are accurately represented with continuous energy data, providing an independent approach for benchmarking other methods. The growing speed of computers will also allow the use of this method for solving complex problems related with utilization of research reactors. A description of the main characteristics of the method and preliminary tests and applications are included here. (author)