Citation:
G. Spence, “PHOENIX: A Reactor Burnup Code with Uncertainty Quantification”, Ph.D. Dissertation, Nuclear Engineering, Texas A&M University, College Station, TX (2014).
Abstract:
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 code ORIGEN-S. 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.