On the accuracy of modeling tokamak neutron sources for Monte Carlo transport simulations

A seminar by Ethan Peterson

Abstract: The starting point of all nuclear analysis for fusion reactor design is the neutron source -- defined by the physics inside the plasma. In the case of Monte Carlo neutron transport simulations, the results of these simulations and our subsequent ability to interpret diagnostic responses or design for adequate shielding or thermal management are limited by the accuracy of these neutron source definitions. In the development of a parametric tokamak neutron source for integration with the open source Monte Carlo code, OpenMC, our team has discovered a number of flaws in existing methodologies. This talk presents the proper methodology for sampling tokamak neutron sources and quantifies the impact of this new method on metrics including the neutron source site distribution, neutron flux distribution, and neutron wall loading.

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