"PHOENIX: A Reactor Burnup Code with Uncertainty Quantification,"
Ph.D. Dissertation, Nuclear Engineering, Texas A&M University, College Station, TX (2014).
Codes for accurately simulating the core composition changes for
nuclear reactors have developed as computing technology developed.
The desire to understand neutronics, material compositions, and
reactor parameters as a function of time has been, and will
continue to be, an area of great interest in nuclear research.
Several methods have been developed to simulate reactor burnup;
however, quantifying the uncertainty in reactor burnup simulations
is in its relative infancy. This research developed a fundamentally
different approach to calculate burnup simulation uncertainty using
perturbations and regression methods. In this work, a computer
software package called PHOENIX was developed that simulates
reactor burnup and provides a quantitative prediction of the
systematic uncertainty associated with simulation modeling
parameters. PHOENIX is a "linkage" code that connects the Monte
Carlo N-Particle transport code MCNP6 to the buildup and depletion
A validation analysis was performed on four different reactor
configurations using PHOENIX. The validation analysis consisted of
two separate components: a code-to-code validation with MONTEBURNS
2.0, and a perturbation validation analysis using two different
perturbation methods. Each analysis observed differences in reactor
parameters and gram compositions for a selected isotopic suite and
compared them to a pre-determined validation criteria. For each
reactor configuration modeled, PHOENIX produced values that
successfully passed the pre-determined validation criteria.