Statistical error in calculating the nuclide composition of low-burnup fuel

The error arising in the change of the ^sup 235^U and ^sup 239^Pu concentrations as a result of the statistical error in the microscopic cross sections during a computational fuel-run simulation with the MCU and MCNP programs is investigated. The analysis is limited to the thermal neutron spectrum and low fuel burnup. A simplified model simulating a fuel-run calculation using MCU and MCNP type statistical programs is constructed. This model is used to analyze for a commercial uranium-graphite reactor the effect of the rate of recalculation of and the statistical error in the microscopic cross sections over a run on the calculation of the ^sup 235^U and ^sup 239^Pu concentrations. The results show that the influence of the statistical error on the computed ^sup 235^U and ^sup 239^Pu concentration is negligible even with 10^sup 5^ neutron histories in the statistical computational sample over a run.[PUBLICATION ABSTRACT]