Neutron moderators, such as nuclear-grade graphite, are essential components of high-temperature nuclear reactors. Their primary functions are to degrade the energy of fission-born neutrons into an energy range that promotes further fission and to serve as the central core structure. This paper will discuss historic reactor moderator materials, which are relatively simple monolithic materials each having intrinsic benefits and limitations. Additionally, a new class of engineered composite moderators is presented for which two examples are fabricated: magnesia-matrix composite systems with either beryllium-based or metal-hydride entrained phases. In additions to presenting their route to fabrication effectiveness as advanced moderators is discussed.

With significant improvement in High Temperature Superconductors (HTS), several projects are adopting HTS technology for fusion power systems. Compact HTS tokamaks offer potential advantages including lower plant costs, enhanced plasma control, and ultimately lower cost of electricity. However, as compact reactors have a reduced radial build to accommodate shielding, HTS degradation due to radiation damage or heating is a significant and potentially design limiting issue. Shielding must mitigate threats to the superconducting coils: neutron cascade damage, heat deposition and potentially organic insulator damage due x-rays. Unfortunately, there are currently no hi-performance shielding materials to enable the potential performance enhancement offered by HTS. In this work, we present a manufacturing method to fabricate a new class of composite shields that are high performance, high operating temperature, and simultaneously neutron absorbing and neutron moderating. The composite design consists of an entrained metal-hydride phase within a radiation stable MgO ceramic host matrix. We discuss the fabrication, characterization, and thermophysical performance data for a series of down-selected composite materials inspired by future fusion core designs and their operational performance metrics. To our knowledge these materials represent the first ceramic composite shield materials containing significant metal hydrides.

A series of monometallic thermal convection loops (TCLs) has been conducted to determine the maximum temperature where FeCrAl alloys have compatibility with eutectic lead lithium (Pb-Li) for a dual-coolant fusion blanket. Pre-oxidizing commercial alloy APMT (Fe-21Cr-5Al-3Mo) for 2 h at 1000°C to form a surface α-Al₂O₃ layer was very effective in reducing the mass loss of specimens in the hot and cold legs of the most recent TCL that was operated for 1000 h with a peak temperature of 700°C. However, unlike previous experiments, the postexposure room temperature ductility of many of the APMT specimens was degraded to 680°C) were severely damaged or were not recovered. Wide-angle X-ray scattering found that the pre-formed α-Al₂O₃ scale transformed to a mixture of trigonal and tetragonal LiAlO₂. The overall results suggest that the maximum temperature for FeCrAl is limited to <700°C.

Here, we employ a combination of advanced synchrotron-based scattering characterization techniques to understand and unravel the atomic origins of the colossal stored energy release in neutron-irradiated silicon carbide. The quantification of the neutron irradiation-induced defects and their impact on the structure-property relationship are important for the design and safe operation of advanced fission and fusion reactors. Our experimental results show that the atomic structure in the as-irradiated samples is significantly perturbed due to a large fraction of vacancy- and interstitial-type defects that lead to complex microstructures and additional components in the x-ray diffraction and pair distribution function (PDF) results. We directly correlate the stored energy release to the recovery of the sublattices with PDF analysis, highlighting that the carbon interstitial- and vacancy-type defects contribute to stored energy more than those of silicon. We find these results to be striking and believe our discoveries to be timely and noteworthy given the technological importance of silicon carbide to the nuclear fission and fusion communities.

The chemical interactions in Fe–HfH2 metal matrix composites (MMCs) are studied across multiple length scales to elucidate the decomposition of the parent phases and corresponding reaction zone physics during direct current sintering. Fe–HfH2 composites were synthesized with increasing as-mixed hydride contents of Fe–25% HfH2, Fe–40% HfH2, Fe–55% HfH2, and Fe–70% HfH2 (all in vol. %) to demonstrate the ability to achieve sintered MMCs with target hydride contents. Samples were probed across multiple length scales through a multi-modal workflow employing x-ray diffraction, scanning electron microscopy and segmentation analysis, and synchrotron techniques including hard x-ray fluorescence mapping and nanoprobe x-ray absorption near-edge structure measurements. Under the selected sintering temperature and pressure conditions, hydrogen evolution is seen to evolve through parallel paths: thermal decomposition from during the transformation of HfH2 to HfHx<2 and through subsequent reaction with the Fe matrix leading to intermetallic phase formation. Specifically, HfFe and HfFe2 intermetallic formation accelerates the release of hydrogen with a subsequent HfO2 phase forming at grain boundaries. For this MMC, the consumption or loss of hydrogen can be considerable in compacts with initial hydride loading of 25%–40% HfH2 approaching 83% hydrogen loss for the lower volume fraction composites. Increasing the volume fraction of HfH2 to 70% enhanced the retained hydrogen content to 53% and attributed to the reduced interfacial area intrinsic to the increased HfH2 loading in this MMC.