A Generalized Optimization Methodology for Isotope Management

Abstract

This research focuses on developing a new approach to studying the nuclear fuel cycle: instead of employing the trial and error approach currently used in actinide management studies in which reactors are designed and then their performance is evaluated, the methodology developed here first identifies relevant fuel cycle objectives–like minimizing decay heat production in a repository, minimizing Pu-239 content in used fuel, etc.–and then uses optimization to determine the best way to reach these goals. The first half of this research was devoted to identifying optimal flux spectra for irradiating used nuclear fuel from light water reactors to meet fuel cycle objectives like those mentioned above. This was accomplished by applying the simulated annealing optimization methodology to a simple matrix exponential depletion code written in Fortran using cross sections generated from the SCALE code system. Since flux spectra cannot be shaped arbitrarily, the second half of this research applied the same methodology to material composition of fast reactor target assemblies to find optimal designs for minimizing the integrated decay heat production over various timescales. The neutronics calculations were performed using modules from SCALE and ERANOS, a French fast reactor transport code.