PERFORMANCE OF ADVANCED ALLOY MATERIALS EXPOSED TO EXTREME FUSION AND FISSION ENVIRONMENTS

Fusion and fission materials are exposed to a wide variety of extreme conditions in which in situ characterization is difficult. The work presented here focuses on alloyed or special materials that are projected to be used in future DEMO fusion reactors and GEN IV high temperature gas reactors (HTGR). Due to the nature of many various extreme exposures that are found in these types of reactors, this work focused on three important conditions: high temperatures, ion irradiations, and varied heat load deposition (through laser exposures). Fusion plasma facing materials (PFMs) experience high flux deuterium (D⁺) and helium (He⁺) irradiations with high heat loads during steady state operation. When off-normal events such as edge-localized modes (ELMs) occur, a much higher heat load impinges the surface of the PFM, which can cause severe cracking and erosion/splashing of the material. High-Z materials such as tungsten have gained popularity due to their superior thermal properties, low sputtering yield, and low fuel retention, but as little as 10^-5 concentration can quench the plasma and terminates the fusion reaction. Ion irradiations affect the surface in different ways depending on temperature and species, where He⁺ can cause nano-tendril growth (fuzz) on the surface and D⁺ can cause blistering on the surface. The goal of this dissertation is to further investigate promising PFM alternatives, PFM degradation over time, and in situ characterization techniques.

W-Ta alloys studies indicate severe erosion due to the lowered mechanical and thermal properties during synergistic studies. Growth and annealing of the surface were also observed such that with increasing heat loading, increased the pore size, and migration to grain boundaries. Pure W will transmute to Re and Os during operation because of the highly energetic neutrons, thus W-Re alloys were studied. W-Re studies have also shown to increase ductility, but suffer from reduced thermal properties compared to W. It is observed that W-Re alloys experience less surface modification due to ion irradiation than the pure W and W-Ta alloys at high temperature, lower flux and longer high-temperature tests. W-Re and pure W fuzz growth was systematically analyzed at a wide range of temperatures that PFMs will endure, where W-Re showed suppression signs of early-stage fuzz growth at the expected divertor-like temperatures. W-Re alloys did experience more macro size cracking due to the applied transient heating events which indicate its drawbacks as a PFM, lower mechanical and thermal properties than pure W and W-Ta alloys. This study indicates the need for testing PFM at a wide range of temperatures, and ELM-like transients, as well as the need to test the evolving material concentration (irradiation damage, impurity transport, and gas absorption), thermal and mechanical properties.

Due to the harsh environment, fuzz analysis is difficult to perform on site. A novel detection and cleaning method using laser-induced breakdown spectroscopy (LIBS) showed promising detection of W-fuzz relating to the reduced optical reflectivity and geometry of the surface. Furthering LIBS ability in future reactors to study on-site methods of material properties, INCONEL 617, a Ni-based superalloy, was investigated due to its recent spotlight in HTGR reactors and it is easier to reach reactor-like conditions in the laboratory. Applications of INCONEL for DEMO-like reactor structural materials indicate the need for on-site material property identifications, which may be difficult due to its crowded optical emission spectrum. Alloy 617 was characterized using LIBS at high sample temperatures where an increase in electron density and ablation depth is observed while the plasma temperature remained the same. Surface oxides were also characterized indicating LIBS ability to diagnose nuclear structural material degradation over time, in which this method could be the most effective in speed, safety, and cost.