As high-temperature reactor designs progress to demonstration, managing the radioactive wastes from these systems presents unique challenges. This work explores the irradiated graphite source term produced by three reactor designs: The Modular High Temperature Gas reactor (MHTGR), a pebble-bed High Temperature Gas Reactor (pb-HTGR), and a Fluoride-cooled High-temperature Reactor (FHR). We predicted a C-14 concentration of 4.3 Ci/m3 for the MHTGR, 1.2 Ci/m3 for the pebble bed HTGR, and 2.5 Ci/m3 for the gFHR after 20 years of operation. The final C-14 concentration highly depended on the graphite nitrogen impurity, a major precursor for C-14. The C-14 concentration in all reactor types exceeded the 0.8 Ci/m3 threshold, resulting in a Class C waste classification. The costs associated with accepting the graphite after 20 years in a low-level waste disposal facility were projected to be 255 $/kWe for the MHTGR, 248 $/kWe for the pb-HTGR, and 56.8 $/kWe for the FHR.

To ensure chemical stability in storage and transportation containers, graphite waste from Fluoride-cooled High-temperature Reactors (FHRs) requires treatment to remove residual fluorides and beryllium before disposal. This study demonstrates a simple, effective cleaning method for modern fine-grained nuclear graphite exposed to molten FLiBe salt under neutron irradiation. A straightforward soaking in deionized water successfully removed near-surface salt, confirmed through cross-sectional scanning electron microscopy, energy-dispersive spectroscopy mapping, and X-ray diffraction. Additional in situ electrochemical monitoring and ex situ fluorometer assays validated this approach. Post-cleaning analyses confirm the process effectively decontaminates graphite surfaces exposed to FLiBe during irradiation, offering a practical solution for FHR graphite waste management.

In this article, we present the creep characteristics of two reduced activation ferritic-martensitic steels of identical starting compositions formed by different fabrication routes: a nanostructured ferritic alloy commonly referred to as a castable nanostructured alloy (CNA) and a sintered nanostructured alloy (SNA) variant. Through a series of nanoindentation experiments spanning a temperature range of 25 degrees C to 650 degrees C, with a maximum load of 100 mN, we find creep behaviors in the cast and sintered materials to be remarkably similar. The creep stress exponent (n) for CNA and SNA were found to be in the range of 8-35 and the activation volume was similar to 14-42b(3), underscoring a dominance of dislocation-mediated mechanisms in both alloys. Notably, we observed a decline in the creep stress exponent with increasing temperature, attributable to the heightened influence of thermally activated dislocations. This phenomenon suggests a potential transition in the deformation mechanism towards a thermally activated dislocation climb process, significantly impacting the observed creep behavior.

As the US fusion materials community awaits the selection and design of a fusion prototypical neutron source (FPNS), a risk reduction exercise has been conducted to (i) provide an updated materials performance evaluation using state-of-the-art computational materials modeling, (ii) expand on legacy analysis based on pure Fe to other relevant fusion structural materials types, and (iii) ensure that materials response under FPNS operational conditions is consistent with referential fusion reactor conditions. The current paper describes the efforts undertaken to assemble a comprehensive computational methodology that includes neutronics, primary damage calculations, atomistic simulations of displacement cascades, chemical inventory evolution calculations, and a computational thermodynamic analysis of emerging phases during irradiation. Our work extends existing studies in pure Fe to reduced-activation ferritic/martensitic steels, tungsten, silicon carbide, and vanadium alloys. We focus on the single-beam deuteron/lithium-stripping neutron source behind the IFMIF-DONES concept, which we assess against ITER, two DEMO designs, and an ideal pure 14-MeV flux. Our analysis indicates that, within standard uncertainties inherent to the models employed, the DONES concept adequately captures fusion conditions in the four materials analyzed. Our work is intended as a comprehensive irradiation damage analysis of fusion-representative neutron sources, to be used for further neutron source evaluation and fusion facility operation.

Graphite's resilience to high temperatures and neutron damage makes it vital for nuclear reactors, yet irradiation alters its microstructure, degrading key properties. We used small- and wide-angle X-ray scattering to study neutron-irradiated fine-grain nuclear graphite (Grade G347A) across varied temperatures and fluences. Results show significant shifts in internal strain and porosity, correlating with radiation-induced volume changes. Notably, porosity volume distribution (fractal dimensions) follows non-monotonic volume changes, suggesting a link to the Weibull distribution of fracture stress.

Development of high-strength materials often involves introduction of additional strengthening microstructures that also serve as tritium trapping sites. Such additions in fusion material development could degrade the fuel efficiency in fusion reactors and raise radiological concerns. The contribution of individual microstructure features in hydrogen trapping must be evaluated to ensure fuel efficiency and radiological safety. This study explores the mechanistic origins of deuterium trapping in reduced-activation ferritic-martensitic steels and its correlation to mechanical strengthening. A series of model alloys and engineering steels were fabricated and subjected to different heat treatments to control deuterium trapping site density. Deuterium retention was evaluated using D2 gas charging and thermal desorption spectroscopy, focusing on the role of grain boundary, dislocation, M23C6 precipitates, and TiC precipitates. Multiscale microstructure characterization and synchrotron X-ray diffraction were performed to characterize microstructure, which was correlated to the deuterium retention property. Results show that TiC precipitates exhibit the highest deuterium trapping capacity, followed by M23C6 precipitates. Dislocation and grain boundary demonstrate the lowest and similar efficiencies. The relationship of trapping quantity and mechanical strengthening of these microstructure features was quantified, demonstrating that TiC precipitates offer highest deuterium trapping per unit of mechanical strengthening.

Vanadium alloys and FeCrAl were investigated as interlayers between tungsten and reduced activation ferritic martensitic steel for fusion system components to avoid formation of intermetallic phase at operating temperatures between 550 and 1100 degrees C, while maintaining a body centered cubic phase throughout the interface. Physical and mechanical properties need to be graded between tungsten and steel, but recent results showed a significant hardness increase at the FeCrAl to vanadium alloy interface. Here, a sintered sample of these alloys was annealed for extended time, and the microstructure was investigated to provide a better understanding of the phenomena. A comparison with an additively manufactured interface of the same material is provided. An unexpected L21 intermetallic phase formation has been revealed using microscopy and synchrotron techniques and will inform future additive manufacturing approaches of the interface. A Cr layer interface as a preliminary solution was proposed between the Vanadium alloy and FeCrAl alloy interface.

Metal hydrides, including ZrHx and YHx, are of particular interest for advanced thermal fission reactors as they have high neutron moderating power and can be used at relatively high temperatures. They have direct applications as core components including as a moderating addition in nuclear fuel, and as neutron reflectors or moderators. Understanding their thermal and irradiation-induced property changes are important to their engineering application. Specifically, evolving metal hydrogen ratios are of critical importance. In this work we discuss the post-irradiation examination of neutron irradiated ZrH2-x and YH2-x specimens. We employ multiple characterization techniques including X-ray diffraction, scanning electron microscopy and thermophysical (thermal diffusivity) to determine the irradiation-induced macro- and microstructural evolution as a function of irradiation temperature. We readily quantify degradations in the thermal diffusivity, changes in lattice parameters, and an increase in metallic Zr indicative of hydrogen release in ZrH2-x specimens. Interestingly, minimal-tonil change in the metallic Y fraction was quantifiable in the YH2-x specimens and modest changes in the thermal diffusivity occur for the temperature and dose studied. The loss of hydrogen in the ZrH2-x samples is related to an apparent irradiation-accelerated desorption of hydrogen by the high ionizing radiation components (gamma, epithermal and fast neutron fluxes) from the in-core neutron irradiation. The most apparent feature from the microstructural analysis for both metal hydrides was a temperature-dependent decrease in the X-ray diffraction peak broadening, attributable to changes in the number and makeup of the two-dimensional defects. These results and trends improve both the fundamental understanding of neutron-solid interactions, and the development of such an important class of core materials.

Graphite has been used in large volumes as a structural material and neutron moderator since the earliest days of nuclear fission. However, no international consensus exists on the disposal of irradiated graphite, leaving much of the historic radioactive graphite inventory in interim vault or silo storage. With several new graphite-moderated reactors planned or under construction, the issue of graphite waste management is becoming increasingly urgent.This paper reviews and quantifies impurities in both historic and modern nuclear graphite, with emphasis on nitrogen—responsible for much of the 14C inventory—and chlorine, which plays a critical role in repository performance and design. Modern graphites, benefitting from stringent quality-control measures developed for non-nuclear industries, meet or exceed the ASTM Ultra-High Purity nuclear standards, even without halide purification. Both chlorine and nitrogen concentrations have declined over time. For chlorine, identified as a key impurity influencing U.S. waste repository design, we propose a target of 0.1 appm in as-fabricated billets as a reasonable benchmark. Nitrogen sources are traced throughout the graphite production process, with surface and bulk concentrations characterized for all materials studied. Modern graphites commonly exhibit nitrogen levels below 5 appm, with values approaching 1 appm achievable. Using such reduced-nitrogen grades is critical to keeping graphite-induced radioactivity below the greater-than-Class-C waste threshold, thereby avoiding disposal cost penalties of nearly an order of magnitude.

Additive manufacturing (AM) rapidly produces complex shapes crucial for energy technologies and engineering designs. In this work, the microstructure and mechanical properties of additive manufactured Grade 91 (modified 9Cr–1Mo) ferritic/martensitic (FM) steel were investigated. Computational thermodynamics, synchrotron X-ray diffraction, scanning electron microscopy, and microhardness testing were utilized to quantify microstructure (crystallographic phases, microstrain), lattice parameters, and mechanical properties as functions of the build directions and post-build heat-treatment. The as-built microstructures were found to contain a large fraction of microstrain due to two-dimensional defects, and a low-carbon martensite phase (maximum value 38.4 vol%), as quantified from the X-ray diffraction analysis. It was shown that a post-build heat-treatment can effectively remove this martensitic phase and promote the formation of desired microstructures with dispersed M23C6 carbides which was consistent with Thermo-Calc predictions. From the X-ray diffraction analysis, an ∼82.5% reduction in microstrain, a 61.5% increase in the coherent grain size and complete elimination of body-centered tetragonal (BCT) phase demonstrated that an appropriate heat-treatment of AM built FM steel yields microstructures and hardness comparable to conventionally processed steel. However, based on lattice parameter and hardness map statistics, heat-treatment on the as-manufactured steel did not eliminate anisotropy.

Structural materials for laser-based inertial fusion energy (IFE) reactor concepts are expected to operate under pulsed irradiation conditions, with cycles consisting of microsecond-long neutron bursts followed by inter-pulse periods of up to one second in duration. During each laser shot, irradiation damage is introduced at dose rates that are up to six orders of magnitude higher than those in their magnetic fusion energy (MFE) counterparts. Under certain conditions, the inter-pulse periods may last an amount of time sufficient to anneal much of the damage introduced during each shot. This phenomenon is highly temperature dependent, with pulsed heating directly linked to the pulsed damage, large surface temperature spikes may also occur. As such, this intermittent mode of operation has the potential to lead to fundamental differences in how irradiation damage accumulates in structural reactor materials. However, damage to structural materials under IFE conditions has received comparatively much less attention than in MFE, as pulsed conditions add yet an extra dimension to the already extremely challenging problem of microstructural evolution under fusion neutron irradiation in structural materials. In this work we use the stochastic cluster dynamics (SCD) method to simulate the evolution with time of defect cluster concentrations under IFE conditions. We consider the Laser Inertial Fusion Energy (LIFE) reactor concept as the representative IFE design for our study, for which detailed spectral information is available, including gas transmutant production. We simulate several pulse frequencies and three different temperatures, and compare the results with continuous irradiation cases under identical average dose rates. The simulations are run in Fe-9Cr system as a model alloy for reduced-activation ferritic/martensitic (RAFM) steels, which are the leading structural material candidates for first-wall structures in MFE and IFE devices. We find that, in practically all scenarios, pulsed irradiation restricts the formation of helium-vacancy clusters relative to the levels seen under equivalent steady irradiation conditions. As well, although self-interstitial atom clusters do accumulate under pulsed operation, their number densities remain up to an order of magnitude lower than in continuous irradiation conditions. Based on the SCD results, we provide a temperature-pulse rate map to identify regions where pulsed irradiation may lead to larger defect accumulation than under continuous irradiation.

We describe our recent efforts demonstrating direct current sintering parameters appropriate to mimic near-identical microstructure to optimize reduced activation ferritic martensitic "castable nanostructured alloy". The fabrication process is presented, and through a combination of computational thermodynamics, multimodal characterization, and mechanical testing we confirm that sintering may be used to produce relevant castable nanostructured alloy (CNA). Our success in demonstrating the applicability of sintering to CNA fabrication opens the opportunity to fabricate functionally graded first wall tile structures or other complicated structures with demanding high-temperature performance, as example fusion high heat flux components.

The current and future need for high-energy neutrons has been a subject of increasing discussion and concern. Immediate applications for such an intense neutron source include medical isotope production, high-energy physics (HEP) research, and for materials development and to support qualification for fission reactors. Also, and of the utmost importance, is the need for such a source to inform critical gaps in our understanding of the transmutation materials science issues facing fusion power reactors. A 14 MeV fusion prototypical neutron source (FPNS) has been a critical, yet unresolved need of the fusion program for more than 40 years. Given the narrowing timeline for construction of pilot and fusion power plants the urgency and necessity of such a neutron source has become increasingly time sensitive. One possibility to address this need is a scaled-down version of IFMIF technology ("IFMIF-Lite"), operating at 125 mA with the beam and target technology leveraging technology developed under the IFMIF/EVEDA program. Within this white paper, a blueprint of necessary R&D to enable a transformational change in both the capital and operating cost of this IFMIF-Lite driver concept is presented. Enabling this transformation is the replacement of the historic RFQ/LINAC components with multiple compact 35+ MeV D+ drivers, based on compact cyclotrons.

The plasma-facing components (PFCs) of future fusion reactors will have intricate structures and require multiple materials because no one material can simultaneously satisfy all the requirements of the component. The dissimilar material joints in PFCs must withstand extreme thermal and stress gradients under neutron irradiation. The Fusion Research Oriented to Neutron Irradiation and Tritium Behavior at Material Interfaces (FRONTIER) U.S.-Japan collaboration seeks to explore and explain the behavior of internal solid interfaces in PFCs under neutron irradiation. The first step of the collaboration was to identify the leading PFCs that should be studied further and prepare them for the next step, which will include neutron irradiation. Different strategies for material development are being pursued worldwide to produce robust PFCs. Here, an overview is presented of some of the most promising materials in the areas of copper alloys, tungsten-copper composites, tungsten-steel composites, additively manufactured tungsten, particle-reinforced tungsten, and tungsten and SiC fiber composites. Each material's fabrication and benefits are described, and some discussion of remaining questions is given.

In this work, an approach is demonstrated for designing a microreactor based on a High Temperature Gas-Cooled Reactor (HTGR) platform intended for use in remote regions or for emergency response. The pre-conceptual microreactor design generated in this study is based on the Idaho National Laboratory's High Temperature Gas-Cooled Test Reactor (HTGTR) point design. Major design changes to the HTGTR were made to transform the design to a microreactor based on constraints inspired by open literature information from the Office of Secretary of Defense (OSD)/Strategic Capabilities Office (SCO) Pele Program. The primary purpose of this study is to compare reactor performance by evaluating key figures of merit for systems employing two-phase composite moderators (magnesia matrix with either an entrained beryllium or hydride containing moderating phase) relative to a reference graphite moderated case. Design parameters were selected to maximize reactor cycle performance for a given microreactor design with additional performance metrics such as natural resource requirement and fuel cycle costs also calculated. Our results demonstrate that a microreactor design based on the HTGR can be achieved by implementing a high fuel density. By using two-phase composite beryllium- or hydride-based moderators instead of graphite, substantial improvements can be realized in the reactor cycle performance, natural resource utilization, and fuel cycle cost.

This work expands on a pre-conceptual microreactor design based on a High Temperature Gas-Cooled Reactor (HTGR) platform generated in the companion paper of this study. Here, power, coolant inlet temperature, and coolant outlet temperature are varied to outline an operable design space associated maximum allowable component temperatures and coolant pressure drop for a given microreactor design. Calculations show that beryllium- and hydride-based composite moderators have a smaller operating design space than graphite because of a lower allowable moderator temperature under the conservative upper limits employed for this study. Power below 6MWth was considered to be undesirable for all designs because of a limited inlet/outlet temperature range. Reactivity Temperature Coefficients (RTC) are calculated to illustrate an inherent safety feature of HTGRs and are fed into a Depressurized Loss of Forced Coolant (DLOFC) without active intervention accident simulation, considered to be a bounding case for fuel temperatures in a HTGR. RTC calculations showed that all designs would expect to have a negative isothermal coefficient through their entire operating cycle and DLOFC calculations showed no case exceed their maximum temperature limits. An additional accident that simulates the release of fission products is performed to compare expected minimum Emergency Planning Zone (EPZ) size for all designs. EPZ calculations showed a marginally smaller achievable size for the composite moderators because of a lower concentration of relevant nuclides due to a more thermal spectrum