Validating a Monte Carlo Model of the University of Wisconsin Nuclear Reactor with Operational Data

Abstract

A model of the University of Wisconsin Nuclear Reactor has been developed using MCNP5. Benchmarking of the model has centered on available reactor operations data from the original loading of the current TRIGA Fuel Life Improvement Program core, including control element differential worth curves and axial flux maps. By simulating the experimental measurement procedure for control element worth, integral worth values obtained for three control blades are within 6% of measured values. The comparison of simulated and measured axial thermal flux profiles suggests the need to improve the definition of the core temperatures and detailed isotopics. Future plans include modeling full-power (1 MW), high-temperature operation and burnup calculations to obtain current fuel definitions.