A Generalized Optimization Methodology for Isotope Management
Name
NFC-122.pdf
Description
NFC-122.pdf
Size
1.44 MB
Format
Adobe PDF
Checksum (MD5)
1f1d70b22a7a5b06daff281a173a6408
Author(s)
Massie, Mark Edward
Forget, Benoit
Date Issued
September 2010
Publisher
Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Nuclear Fuel Cycle Program
Series/Report no.
MIT-NFC;TR-122
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.
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