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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.

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