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 …