Many technological hurdles remain before a viable commercial magnetic fusion energy reactor can be constructed, including the development of plasma-facing components with long lifetimes that can survive the harsh environment inside a reactor. One such component, the divertor, which maintains the purity of the plasma by removing fusion byproducts from the reactor, must be able to accommodate very large incident heat fluxes of at least 10 MW/m^2 during normal operation. Modular helium-cooled tungsten divertors are one of the leading divertor designs for future commercial fusion reactors, and a number of different candidates have been proposed including the modular He-cooled divertor concept with pin array (HEMP), the modular He-cooled divertor concept with multiple-jet-cooling (HEMJ), and the helium-cooled flat plate (HCFP). These three designs typically operate with helium coolant inlet temperatures of 600 °C and inlet pressures of 10 MPa. Performing experiments at these conditions to evaluate the thermal performance of each design is both challenging and expensive. An alternative, more economical approach for evaluating different designs exploits dynamic similarity. Here, geometrically similar mockups of a single divertor module are tested using coolants at lower temperatures and pressures. Dynamically similar experiments were performed on an HEMP-like divertor with helium and argon at inlet temperatures close to room temperature, inlet pressures below 1.4 MPa, and incident heat fluxes up to 2 MW/m^2. The results are used to predict the maximum heat flux that the divertor can accommodate, and the pumping power as a fraction of incident thermal power, for a given maximum tungsten temperature. A new nondimensional parameter, the thermal conductivity ratio, is introduced in the Nusselt number correlations which accounts for variations in the amount of conduction heat transfer through the walls of the divertor module. Numerical simulations of the HCFP divertor are performed to investigate how the thermal conductivity ratio affects predictions for the maximum heat flux obtained in previous studies. Finally, a helium loop is constructed and used to perform dynamically similar experiments on an HEMJ module at inlet temperatures as high as 300 °C, inlet pressures of 10 MPa, and incident heat fluxes as great as 4.9 MW/m^2. The correlations generated from this work can be used in system codes to determine optimal designs and operating conditions for a variety of fusion reactor designs.

Divertors are in-vessel, plasma-facing, components in magnetic-confinement fusion reactors. Their main function is to remove the fusion reaction ash (α-particles), unburned fuel, and eroded particles from the reactor, which adversely affect the quality of the plasma. A significant fraction (~15 %) of the total fusion thermal power is removed by the divertor coolant and must, therefore, be recovered at elevated temperature in order to enhance the overall thermal efficiency. Helium is the leading coolant because of its high thermal conductivity, material compatibility, and suitability as a working fluid for power conversion systems using a closed high temperature Brayton cycle. Peak surface heat fluxes on the order of 10 MW/m^2 are anticipated with surface temperatures in the region of 1,200°C to 1,500°C. Recently, several helium-cooled divertor designs have been proposed, including a modular T-tube design and a modular finger configuration with jet impingement cooling from perforated end caps. Design calculations performed using the FLUENT® CFD software package have shown that these designs can accommodate a peak heat load of 10 MW/m^2. Extremely high heat transfer coefficients (~50,000 W/(m^2 K)) were predicted by these calculations. Since these values of heat transfer coefficient are considered to be outside of the experience base for gas-cooled systems, an experimental investigation has been undertaken to validate the results of the numerical simulations. Attention has been focused on the thermal performance of the T-tube and the finger divertor designs. Experimental and numerical investigations have been performed to support both divertor geometries. Excellent agreement has been obtained between the experimental data and model predictions, thereby confirming the predicted performance of the leading helium-cooled divertor designs for near- and long-term magnetic fusion reactor designs. The results of this investigation provide confidence in the ability of state-of-the-art CFD codes to model gas-cooled high heat flux plasma-facing components such as divertors.