Measurement, simulation and uncertainty quantification of the neutron flux at the McMaster Nuclear Reactor

Neutron flux measurements in research reactors can be used for code validation and optimizing in-core activation procedures. Since the fuel adjacent to an irradiation site undergoes burnup , and may be shuffled, local flux measurements may be subject to an additional source of burnup-dependent uncertainty. It is unfeasible to perform these measurements for all core conditions; therefore, reactor physics codes may provide supplemental flux information. This work includes a validation study of the MCNP6 model of the McMaster Nuclear Reactor (MNR). Irradiations were performed over several months, with an emphasis on uncertainty quantification during data processing . No change in the local flux was measured over this period of operation, indicating that burnup effects may be insignificant compared to other sources of uncertainty. These results were validated by five sets of computational data from historical MNR cores. Burnup effects do not need to be accounted for in determining neutron flux uncertainties.