Validation of a fuel management code MCODE-FM against fission product poisoning and flux wire measurements of the MIT reactor

Abstract

The Massachusetts Institute of Technology Research Reactor (MITR), with thermal power of 6 MW, is one of the five high-performance research reactors operated in the U.S. An in-house program MCODE-FM (MCNP-ORIGEN Coupled Depletion Program – Fuel Management) as the reference code supporting routine operation is currently developed by the MITR staff. Adopting Monte Carlo methods enables exact modeling of the MITR core geometry with use of continuous-energy nuclear data. A criticality search algorithm to track control blade movement is implemented in MCODE-FM. The code also features automation of input file generation, data manipulation, and post-processing of output data for fuel cycle analysis. Some verification and validation runs have been carried out by comparisons with the multiplication factor and the reactivity worth induced by control blades. In this study, further validation runs are performed, with two sets of measured data: 1) reactivity effects of fission product poisoning and fuel depletion and 2) reaction rate based thermal and fast neutron flux. Fission product poisoning effects calculated by MCODE-FM are in good agreement with the measured data obtained during the reactor start-up process (within 10%). The trend of peak xenon reactivity after the reactor shutdown is predicted accurately as well. As compared to the neutron activation experiment, MCODE-FM is found to slightly over-estimate the thermal flux with less than 10% discrepancy and the fast neutron flux with less than 20% discrepancy.