Speaker
Description
Reactor system codes use reduced-order models to represent each key subsystem of the fusion power plant, such that rapid self-consistent evaluation of full fusion power plant concepts can be performed. Traditionally system codes must rely on fitted empirical data to estimate neutron transport and deposition. PROCESS is the primary systems code used by UKAEA, which could benefit from the addition of a semi-analytical neutronics model. Finding the solution to the full Boltzmann transport equation for neutrons is expensive, therefore a diffusion approximation is applied, and the geometry is simplified to only account for variation in materials and neutron flux in the radial direction. Materials are assumed to be layered perpendicular to the direction at which neutrons diffuse out of the plasma, and the multi-group neutron flux is solved for all points along the radial direction, such that neutron heating, tritium breeding rate, neutron leakage rate, and neutron damage can be calculated.
This model is applied onto a generic fusion reactor’s geometry, and then evaluated against an OpenMC simulation of an analogous geometry. The discrepancies between the two are explored, and the effect on reactor design optimization is discussed. Ultimately, the addition of this neutronics model allows for more realistic constraints and optimization objectives to be calculated: for example, PROCESS can then be ran with a constraint on the coolant pumping power such that the heat removal rate is at least equal to the neutron heating rate; or optimized for a lower DPA on the central solenoid such that the reactor’s down time may be minimized.
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