Abstract
The error arising in the change of the 235U and 239Pu 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 235U and 239Pu concentrations. The results show that the influence of the statistical error on the computed 235U and 239Pu concentration is negligible even with 105 neutron histories in the statistical computational sample over a run.
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