SUMMARY

The magnetic fusion energy (MFE) tokamak reactor, which confines a high temperature plasma within a torus shaped 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 challenges of current long-pulse MFE reactor chamber is overcoming the extreme conditions at the first wall of the reactor chamber, where plasma facing surfaces are subject to very high heat fluxes. Among the plasma facing components is the divertor, an essential element that removes the impurities and fusion products 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/m2, must therefore be cooled. In addition, about 20% of the total energy produced by the fusion reaction is absorbed by plasma facing surfaces including the divertor, and therefore it is essential to have a cooling system that can recover this energy. Several designs for cooling the divertor targets have been proposed, and most use helium to cool the back side of the target plates with impinging jets. This doctoral thesis details experimental and numerical studies to estimate the cooling capabilities and required pumping power under prototypical conditions of a number of helium cooled divertor designs, specifically 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. Experiments were performed over ranging He mass flow rate of 3 – 10 g/s in helium loops operating at the prototypical pressure of 10 MPa, and elevated helium inlet temperatures as high as 400°C and incident heat fluxes as high as 8.1MW/m2. Numerical simulations using commercial software, validated by these experimental data, complement these experimental studies and are used to extrapolate the thermo-fluids performance to fully prototypical conditions.