Thermal-Hydraulics of Helium-Cooled Divertors
Providing energy from fusion is one of the National Academy of Engineering’s Grand Challenges. Magnetic fusion typically uses toroidal magnetic fields to confine a plasma of hydrogen isotopes at the extremely high temperatures and pressures necessary to achieve fusion. About 20% of the energy produced by the burning plasma will be dissipated on the solid plasma-facing surfaces, which is enough to erode and melt these surfaces. By confining the core plasma within a separatrix with a single or double null(s) in the poloidal magnetic field, this energy can be “diverted” away from the first wall of the reactor, so that it instead impinges upon, and is removed by, the target surfaces of a “divertor” near the null. The target surfaces of the divertor are, however, subject to very high steady-state heat fluxes—estimated to be at least 10 MW/m2.
Over the last decade, our group has experimentally and numerically studied the thermal-hydraulics of a variety of proposed helium-cooled solid-tungsten divertor designs. Our current efforts are focused on the helium-cooled multi-jet (HEMJ) modular divertor, one of the leading candidates for the demonstration fusion power plant DEMO, which has been proposed as the “next step” after ITER. Models of a HEMJ module are being tested in a high-pressure, high-temperature (as great as 10 MPa and 400°C) helium loop, and computational fluid dynamics and finite-element simulations are being performed with commercial software to determine the thermal performance of this design. The goal of this work is to develop design correlations to estimate the thermal-hydraulics performance of current divertor designs, and to improve these designs.
Collaborators: Said Abdel-Khalik, School of Mechanical Engineering
For more information on ITER: https://www.iter.org/