Analysis of in-core experiment activities for the MIT Research Reactor using the ORIGEN computer code

Abstract

The objective of this study is to devise a method for utilizing the ORIGEN-S computer code to calculate the activation products generated in in-core experimental assemblies at the MIT Research Reactor (MITR-II). ORIGEN-S is a nuclear depletion and decay analysis code. It accounts for all types of nuclear reactions and eliminates the need for selection of the dominant reactions that will occur in a given experiment, as must be done with the existing activity calculation method. It is expected that the new approach will be easy to use, and will produce radioactivity estimations that are generally more accurate than those produced by the existing method. The ORIGEN-S method has been developed and tested for four experiments that have been or are scheduled to be irradiated in the MITR. These experiments are the Advanced Cladding Irradiation (ACI), High Temperature Irradiation Facility (HTIF), Electric Power Research Institute Electro-Chemical Potential (EPRI ECP) loop, and Annular Fuel Test Rig (AFTR). The method has also been used to perform activation analyses for ten individual elements (plus U-235 and U-238) that are commonly found in MITR in-core experiment (ICE) assemblies. The ORIGEN-S analyses for the ACI, HTIF, and EPRI ECP experiments produced results that were relatively similar to the results produced by previous analyses that utilized the current method of activation estimation. This is because the thermal neutron capture reactions, which are major contributors to the activation of these experiments, are already well accounted for in the existing method. The results of the ORIGEN-S analysis for the AFTR, which contains fissile material, were also very similar to the results of the previous analysis, despite the fact that the previous analysis accounted for changes in flux due to fissile nuclide depletion during irradiation and the current analysis did not.

It is concluded that the activation calculation method developed should be generally adequate for all experiments irradiated in the MITR core. A possible exception involves experiments containing quantities of fissile material larger than the quantities contained in the AFTR, as these experiments could produce significant changes in neutron flux levels that would render this method inadequate.