SUMMARY

The magnetic fusion energy (MFE) tokamak reactor, which confines a high temperature plasma within a toroidal chamber by means of magnetic fields, is one of the most promising and best developed concepts for making nuclear fusion energy possible. One of the components of current MFE reactor chamber designs is the divertor, an essential element that removes the impurities from the core plasma, and is thus directly exposed to the plasma. The solid target plates of the divertor, which are exposed to very high heat fluxes of at least 10 MW/m^2, must therefore be cooled. Several designs for cooling divertors have been proposed. Most of these designs use helium to cool the back side of the target plates with impinging jets. In this proposed doctoral thesis, a number of helium cooled divertor designs including the helium-cooled flat plate (HCFP) divertor, helium-cooled modular divertor with multiple jets (HEMJ), and a “flat design” which is a simplified variant of the HEMJ, will be experimentally and numerically investigated to estimate their cooling capabilities and required pumping power under prototypical conditions. Experiments are proposed over a range of parameters in helium loops operating at the prototypical pressure of 10 MPa, and nearly prototypical helium inlet temperatures and incident heat fluxes. Numerical simulations, validated by these experimental data, will supplement these experimental studies and be used to estimate thermo-fluids performance under fully prototypical conditions and the robustness of these concepts with respect to dimensional variations.