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At least 739 records · Page 41
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Chemical and morphological evolution of hybrid conversion coatings in low-Earth orbit space environment

Understanding how protective coatings respond to the harsh low-Earth orbit (LEO) environment is essential for ensuring the safety, longevity, and cost-effectiveness of spacecraft. In particular, identifying environmentally friendly, non-chromate alternatives that can maintain performance under such conditions has both technological and regulatory significance. This study investigates the environmental stability of zirconium-based hybrid conversion coatings with Cu additives (Cu10 and Cu20) applied to cold-rolled steel, tested in the Materials International Space Station Experiment (MISSE) outside the International Space Station (ISS). Chemical and morphological analyses were carried out using a combination of electron microscopy and X-ray spectroscopy techniques, including scanning electron microscopy (SEM), scanning transmission electron microscopy with energy-dispersive X-ray spectroscopy (STEM-EDS), X-ray photoelectron spectroscopy (XPS), and X-ray absorption near-edge structure (XANES) spectroscopy. After exposure outside the ISS, all coatings remained structurally intact, with all exhibiting a uniform Zr-rich matrix and embedded Cu-rich clusters, while a thin Si-rich surface layer developed from interaction with space environments. Depth-resolved XPS showed a layered structure with CuO on the surface, Cu 2 O, and partial Zr(IV) reduction near Cu-rich sites, evidence of Atomic Oxygen (AO)-driven surface oxidation. These results demonstrate that Cu–Zr coatings maintain their chemical integrity and microstructure in harsh space environments, offering a non-chromate alternative for long-term aerospace protection. These insights provide valuable guidance for developing next-generation protective coatings that combine environmental sustainability with the reliability required for future aerospace and orbital applications.

36 MATERIALS SCIENCE↗
DOE OSTI2026

The stellar origins of 96 Zr excesses in presolar graphites from the Murchison meteorite

Context. Zirconium-96 is a stable isotope that can be synthesized under different neutron-rich nucleosynthetic conditions. Astrophysical models predict its production to occur in various stellar environments: from low-to-intermediate-mass asymptotic giant branch (AGB) stars to massive stars and core-collapse supernovae. Aims. Detections of 96Zr excesses, in combination with other isotopic measurements from presolar grains can provide unique constraints on its stellar origin. Presolar grains are microscopic particles found in primitive Solar System materials, which formed in stellar winds and supernova ejecta. The isotopic composition of each grain can provide us a snapshot of the nucleosynthetic processes that took place during the parent star’s lifetime. Methods. In this study, we measured the stable isotopes of C, N, O, Mo, Zr, and Ru in high-density presolar graphite grains from the Murchison meteorite and found four grains that contain positive isotopic anomalies in 96 Zr carried by their internal subgrains. We analyzed multi-element isotopic datasets from each grain to explore the source of the observed 96 Zr excesses. Results. Comparisons with stellar models indicate that two grains likely condensed in an intermediate-mass AGB star with initial metallicity of Z ≤ 0.014. Their 96 Zr/ 94 Zr ratios also match those predicted for born-again AGB stars undergoing a very late thermal pulse and rapidly accreting white dwarfs. After comparing the relative populations of the aforementioned dust-producing stars, we propose rapidly accreting white dwarfs as a new, and more likely, stellar source for one of the presolar grains. The remaining two grains could have originated in the supernova ejecta of massive stars, due to correlated excesses in the p-nuclides, 92, 94 Mo. Thus, grains with 96 Zr anomalies can have a variety of stellar origins, in agreement with theoretical studies. Conclusions. Our study highlights the importance of multi-element analysis in constraining the types of stars where presolar grains have condensed. These data will help improve our understanding of various nucleosynthesis processes in different stellar phases.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2020

Cubic crystals in an x-ray polarization-splitting geometry

Hexagonal and cubic crystals contain paired sets of internal planes that reflect the linearly polarized components of certain x rays into two separate, perpendicular directions. For the cubic crystals, two distinct crystal orientations provide the same polarization-splitting geometry. One of the orientations may have advantages for plasma spectroscopy by suppressing unwanted reflections. This paper demonstrates the two orientations with a germanium crystal and K characteristic lines from copper and zirconium.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Structural reversibility of Cu doped NU-1000 MOFs under hydrogenation conditions

The metal-organic framework (MOF), NU-1000, and its metalated counterparts have found proof-of-concept application in heterogeneous catalysis and hydrogen storage amongst others. A vapor-phase technique, akin to atomic layer deposition (ALD), is used to selectively deposit divalent Cu ions on oxo, hydroxo-bridged hexa-zirconium(IV) nodes capped with terminal –OH and -OH 2 ligands. Subsequent reaction with steam yields node-anchored, CuII-oxo,hydroxo clusters. We find that cluster installation via AIM (= ALD In MOFs) is accompanied by an expansion of MOF mesopore (channel) diameter . We investigated the behavior of the cluster-modified material, termed Cu-AIM-NU-1000, to heat treatment up to 325 °C, at atmospheric pressure with a low flow of H2 into the reaction cell. The response under these conditions revealed two important results: (1) Above 200 °C, the initially installed few-metal-ion clusters reduce to neutral Cu atoms. The neutral atoms migrate from the nodes and aggregate into Cu nanoparticles. While the size of particles formed in the MOF interior is constrained by the width of mesopores (ca. 3 nm), those formed on the exterior surface of the MOF can grow as large as ca. 8 nm. (2) Reduction and release of Cu atoms from the MOFs nodes is accompanied NU-1000 undergoes dynamic structural transformation as it reverts back to its original dimension following the release. These results show while the MOF framework itself remains intact at 325 °C in an H 2 atmosphere, the small, AIM-installed Cu II -oxo,hydroxo clusters are stable with respect to reduction and conversion to metallic nanoparticles only up to ~200 °C.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Wake-up and fatigue mechanisms in ferroelectric Hf 0.5 Zr 0.5 O 2 films with symmetric RuO 2 electrodes

In this work, the mechanisms leading to wake-up and fatigue in ferroelectric hafnium zirconium oxide thin film devices with symmetric RuO 2 electrodes are investigated via polarization, relative permittivity, dielectric nonlinearity, pyroelectric coefficient, and microfocus X-ray diffraction (XRD) measurements. The devices are observed to wake-up for up to 10 3 bipolar pulsed field cycles, after which fatigue occurs with the polarization approaching zero following 10 8 cycles. Wake-up is accompanied by a decrease in both high-field permittivity and hysteresis loop pinching and an increase in pyroelectric coefficient, indicating that the wake-up process involves a combination of transformations from the tetragonal to the orthorhombic phase and domain depinning from defect redistribution. Fatigue is observed to coincide with an increase in irreversible domain wall motion and a decrease in pyroelectric coefficient. Finite pyroelectric coefficients are measured on fully fatigued devices, indicating that domain pinning is a strong contributor to fatigue and that fatigued devices contain domain structures that are unable to switch under the fields applied for measurement. Microfocus XRD patterns measured on each device reveal that the phase constitution is qualitatively unaffected by the field cycling and resultant polarization fatigue. This data indicates that the wake-up process has contributions from both phase transformations and domain depinning, whereas the fatigue process is driven primarily by domain pinning, and the near-zero measured switchable polarization is actually a poled device with immobile domains. These observations provide insight into the physical changes occurring during field cycling of HfO 2 -based ferroelectrics while examining a possible oxide electrode material for silicon CMOS device implementation.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2022

Thermite and intermetallic projectiles examined experimentally in air and inert gas environments

Intermetallic (aluminum and zirconium) and thermite (aluminum and molybdenum trioxide) projectiles were launched using a high velocity impact ignition testing system. The experiments were designed to simulate reactivity in high (argon) and low (air) altitude environments. The projectiles were launched into a chamber that included a steel target plate for projectile penetration before impacting a rear witness plate. The chamber was semi-sealed and instrumented for quasi-static pressure data. The results provide an understanding of energy release from the projectile materials and of the environmental influence on performance. The transient pressure traces provide insight into reaction kinetics. A bifurcation in transient pressure rise was an indication of a shift in reaction kinetics from the inherent reactive material to metal oxidation with the environment. The bifurcation was delayed by about 0.15 ms for the intermetallic relative to the thermite, evidence that the thermite reaction proceeded faster upon impact than the intermetallic. The two-step process (impact ignition of the reactive material followed by metal oxidation) was shown to produce higher energy conversion efficiencies than projectiles composed of pure fuel (i.e., aluminum) reported previously. Both reactive materials showed energy conversion efficiencies greater than 30% (for air) and 50% (for argon), and an explanation of underestimated efficiency and energy losses is provided. These results have implications for advancing formulations for ballistic applications. Structural reactive materials can be used to modify the effective reactivity of metal-containing formulations in varied atmospheric environments.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Thermal conductivity evaluation of ion irradiated Si 3 N 4 and ZrN ceramics using spatial domain thermoreflectance

Nitride ceramics have been investigated for different applications in the nuclear industry, such as space nuclear power, fusion reactor diagnostics and plasma heating, inert matrix fuels, and accident tolerant fuels. Although thermal conductivity remains one of the most important properties to track following irradiation, traditional techniques such as laser flash and xenon flash are limited to bulk sample characterization, which requires lengthy and cost-consuming neutron irradiation. This work used spatial domain thermoreflectance (SDTR) for the micrometer-scale measurement of thermal conductivity in 15 MeV Ni ion-irradiated silicon nitride and zirconium nitride from 1 to 50 dpa and 300 to 700 °C. Here, the SDTR-measured unirradiated thermal conductivity was found to be consistent with the published data on bulk samples. Electrically conductive ZrN exhibits modest reduction after irradiation which is minimal at the highest irradiation temperatures. In electrically insulating Si 3 N 4 , the reduction is more significant and unlike ZrN, the reduction remains significant even at a higher irradiation temperature. The thermal resistance evolution following irradiation was compared with lattice swelling, which was determined using grazing incidence x-ray diffraction, and radiation-induced defects were observed using transmission electron microscopy. A saturation value was observed between 15 and 50 dpa for thermal conductivity degradation in both nitride ceramics and a direct correlation with high-temperature defect recombination was observed, as well as the potential presence of additional carrier scattering mechanisms.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Physics-based model of irradiation creep for ferritic materials under fusion energy operation conditions

Irradiation creep is known to be an important process for structural materials in nuclear environments, potentially leading to creep failure at temperatures where thermal creep is generally negligible. While there is a great deal of data for irradiation creep in steels and zirconium alloys in light water reactor conditions, much less is known for first wall materials under fusion energy conditions. Lacking suitable fusion neutron sources for detailed experimentation, modeling, and simulation can help bridge the dose-rate and spectral-effects gap and produce quantifiable expectations for creep deformation of first wall materials under standard fusion conditions. In this paper, we develop a comprehensive model for irradiation creep created from merging a crystal plasticity representation of the dislocation microstructure and a defect evolution simulator that accounts for the entire cluster dimensionality space. Both approaches are linked by way of a climb velocity that captures dislocation-biased defect absorption and a dislocation strengthening term that reflects the accumulation of defect clusters in the system. Here, we carry out our study in Fe under first wall fusion reactor conditions, characterized by a fusion neutron spectrum with average recoil energies of 20 keV and a damage dose rate of ≈3×10 –7 dpa/s at temperatures between 300 and 800 K.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Machine learned interatomic potential for dispersion strengthened plasma facing components

Tungsten (W) is a material of choice for the divertor material due to its high melting temperature, thermal conductivity, and sputtering threshold. However, W has a very high brittle-to-ductile transition temperature, and at fusion reactor temperatures (≥1000 K), it may undergo recrystallization and grain growth. Dispersion-strengthening W with zirconium carbide (ZrC) can improve ductility and limit grain growth, but much of the effects of the dispersoids on microstructural evolution and thermomechanical properties at high temperatures are still unknown. We present a machine learned Spectral Neighbor Analysis Potential for W–ZrC that can now be used to study these materials. In order to construct a potential suitable for large-scale atomistic simulations at fusion reactor temperatures, it is necessary to train on ab initio data generated for a diverse set of structures, chemical environments, and temperatures. Further accuracy and stability tests of the potential were achieved using objective functions for both material properties and high temperature stability. Additionally, validation of lattice parameters, surface energies, bulk moduli, and thermal expansion is confirmed on the optimized potential. Tensile tests of W/ZrC bicrystals show that although the W(110)–ZrC(111) C-terminated bicrystal has the highest ultimate tensile strength (UTS) at room temperature, observed strength decreases with increasing temperature. At 2500 K, the terminating C layer diffuses into the W, resulting in a weaker W–Zr interface. Meanwhile, the W(110)–ZrC(111) Zr-terminated bicrystal has the highest UTS at 2500 K.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Response of ZrC to swift heavy ion irradiation

Zirconium carbide (ZrC) is commonly used for energy sector research, as well as a surrogate for the proposed advanced nuclear fuel candidate uranium carbide. Here this study investigates structural modifications to nanocrystalline and microcrystalline ZrC resulting from dense electronic excitations induced by swift heavy ion exposure. Samples were irradiated with 946 MeV Au ions to various fluences up to 6 × 10 13 ions cm –2 and characterized using synchrotron-based x-ray diffraction. The evolution of the unit-cell parameter and heterogeneous microstrain were evaluated as a function of fluence and compared with those of nanocrystalline and microcrystalline CeO 2 (a surrogate for UO 2 fuel) irradiated under identical conditions. Distinct differences were observed in the radiation responses of the carbide and oxide across both grain sizes. Most notably, microcrystalline ZrC exhibits swelling characterized by two distinct regimes, which does not result in saturation at the ion fluences achieved. This contrasts with CeO 2 , which exhibits the well-documented direct-impact defect accumulation mechanism, reaching a steady-state saturation of swelling at higher fluences. Nanocrystalline CeO 2 undergoes more pronounced swelling compared with microcrystalline CeO 2 , in contrast to nanocrystalline ZrC, which exhibits only minimal unit-cell changes. These results demonstrate that swift heavy ion-induced structural changes can be quite different in carbides and oxides, which must be considered when extrapolating fission-fragment type damage in current fuels to advanced fuels.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

New dynamic diamond anvil cell for time-resolved radial x-ray diffraction

The dynamic diamond anvil cell (dDAC) is a recently developed experimental platform that has shown promise for studying the behavior of materials at strain rates ranging from intermediate to quasi-static and shock compression regimes. Combining dDAC with time-resolved x-ray diffraction (XRD) in the radial geometry (i.e., with incident x-rays perpendicular to the axis of compression) enables the study of material properties such as strength, texture evolution, and deformation mechanisms. This work describes a radial XRD dDAC setup at beamline P02.2 (Extreme Conditions Beamline) at DESY’s PETRA III synchrotron. Time-resolved radial XRD data are collected for titanium, zirconium, and zircon samples, demonstrating the ability to study the strength and texture of materials at compression rates above 300 GPa/s. In addition, the simultaneous optical imaging of the DAC sample chamber is demonstrated. The ability to conduct simultaneous radial XRD and optical imaging provides the opportunity to characterize plastic strain and deviatoric strain rates in the DAC at intermediate rates, exploring the strength and deformation mechanisms of materials in this regime.

47 OTHER INSTRUMENTATION↗
DOE OSTI2024

Uniaxial compressive creep tests by spark plasma sintering of 70% theoretical density α -uranium and U-10Zr

Metallic fuels hold numerous advantages over conventional uranium dioxide fuels and are a key component of several liquid metal-cooled advanced reactor concepts including sodium fast reactors. These fuels undergo rapid swelling during early burnup; consequently, they spend most of their reactor lifetime in a porous state. The presence of this porosity alters many of the mechanical properties of the fuel including creep impacting fuel deformation during axial swelling. This work investigates the creep behavior of the porous fuel using a spark plasma sintering technique. Creep tests were performed for the first time on porous α-phase uranium and uranium with 10 wt. % zirconium (U-10Zr) samples. The samples of α-phase uranium and U-10Zr were fabricated from depleted uranium by spark plasma sintering and subjected to uniaxial compressive creep testing. Calculated stress exponents were found to be 2.6±1.6 and 5.7±1.4 for α-U and U-10Zr, respectively, and calculated activation energies were found to be 61.6±1.1kJ/mol for α-U. The creep data were also used to evaluate existing porosity inclusive in creep models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Multistate resistance in TaN/(Hf,Zr)O 2 /Ta ferroelectric tunnel junctions

Ferroelectric tunnel junctions (FTJs) utilizing hafnium zirconium oxide (HZO) have emerged as promising non-volatile memory elements for microelectronics, compatible with back end of line (BEOL) complementary–metal–oxide semiconductor fabrication. This study investigates asymmetric electrode TaN/HZO/Ta devices with a 6 nm thick HZO layer as FTJs for multistate resistive memory applications. The individual FTJs exhibit a resistance ratio exceeding 10× when utilized as a binary state device, with pulsing between −1.7 and +1.4 V to set the high resistance state (HRS) and low resistance state (LRS), respectively. Following with reduced write voltage pulses allows the ferroelectric device to operate with a selection of over 32 distinct resistance states (2 5 bits) between the LRS and HRS. This work then explores the stability of the resistance states during write/read pulse cycling, along with the stability of the state after multiple read pulses. Accessing the multibit state shows stability within 50 reads with the binary state remaining stable for more than 4000 reads pulses. With their multistate tunability and versatility, FTJs hold promise as BEOL memory elements for compute-in-memory (CiM) arrays, binary digital memory, or weighted vector matrix multiplication applications with low power consumption during computations.

CMOS↗
DOE OSTI2022

Discrete scale invariance of the quasi-bound states at atomic vacancies in a topological material

Recently, log-periodic quantum oscillations have been detected in the topological materials zirconium pentatelluride (ZrTe 5 ) and hafnium pentatelluride (HfTe 5 ), displaying an intriguing discrete scale invariance (DSI) characteristic. In condensed materials, the DSI is considered to be related to the quasi-bound states formed by massless Dirac fermions with strong Coulomb attraction, offering a feasible platform to study the long-pursued atomic-collapse phenomenon. Here, we demonstrate that a variety of atomic vacancies in the topological material HfTe 5 can host the geometric quasi-bound states with a DSI feature, resembling an artificial supercritical atom collapse. The density of states of these quasi-bound states is enhanced, and the quasi-bound states are spatially distributed in the “orbitals” surrounding the vacancy sites, which are detected and visualized by low-temperature scanning tunneling microscope/spectroscopy. By applying the perpendicular magnetic fields, the quasi-bound states at lower energies become wider and eventually invisible; meanwhile, the energies of quasi-bound states move gradually toward the Fermi energy (E F ). These features are consistent with the theoretical prediction of a magnetic field–induced transition from supercritical to subcritical states. The direct observation of geometric quasi-bound states sheds light on the deep understanding of the DSI in quantum materials.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2023

Designing Nuclear Fuels with a Multi-Principal Element Alloying Approach

Previous research has shown that multi-principal element alloys (MPEAs) using chromium, molybdenum, niobium, tantalum, titanium, vanadium, and zirconium can form stable body-centered-cubic (BCC) structures across a large temperature region (25°C to 1000°C). This is the same crystal structure as γ-uranium (U), which has shown desirable thermal and irradiation behavior in previous alloy fuel research. It is hypothesized then that the MPEA alloying approach can be used to produce a stable BCC uranium-bearing alloy and to retain its stability throughout anticipated operating regimes of power-producing reactors. Candidate elements were assessed using Monte Carlo N-Particle (MCNP) analysis to determine uranium densities necessary to make the alloy an economically viable fuel compared to conventional fuel forms. Following neutronic considerations, materials property databases and empirical predictors were used to determine the compositions with a high potential to form a BCC solid solution alloy. The final four alloys were MoNbTaU 2 , MoNbTiU 2 , NbTaTiU 2 , and NbTaVU 2 , which were cast using arc melting of raw elemental foils and chunks. Characterization of the fabricated alloys included scanning electron microscopy, X-ray diffraction, energy dispersive X-ray spectroscopy, and transmission electron microscopy. The results showed a two-phase system with a U-rich matrix phase surrounding the refractory precipitates. The U phase was found to contain varying concentrations of the alloying elements and was a BCC γ-U phase. These results warrant further research to identify ideal compositions for use as an advanced alloy fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Preliminary Neutronics Design and Analysis of the Fast Modular Reactor

General Atomics is developing a new 100-MW(thermal) fast modular reactor (FMR) that provides safe, carbon-free electricity and is capable of incremental capacity additions. The modular design allows it to be factory built and assembled onsite to keep the capital cost low, while the use of dry cooling facilitates siting to complement renewables in nearly any location. The FMR uses high-assay low-enriched uranium-dioxide fuel encapsulated by recognized irradiationresistant silicon carbide composite (SiGA®) cladding that is derisked in the current accident-tolerant fuel program. The FMR fuel assembly is a hexagonal fuel bundle of 120 fuel rods. The total length of the fuel assembly is less than 4 m, with an active fuel length of 1.8 m. The fuel assemblies are configured in an annular core that is located and supported by the reactor internals. The coolant material is helium at a normal operating pressure of 7 MPa. The core is surrounded by zirconium silicide (Zr 3 Si 2 ) and graphite reflector blocks. The fuel, coolant, internals, and reflectors are contained within a reactor pressure vessel. Here, the preliminary nuclear design and analysis established the arrangement of the active core and reflector blocks. The nuclear design analyses of the FMR defined the design parameters, such as fuel enrichments, excess reactivity, fueling scheme, fuel cycle, power distribution, and control rod worth. The preliminary conceptual design determined the three-batch fueling scheme with the allowable total power peaking factor of 1.5. The average discharge burnup is 100 GW days per ton of uranium.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Optimizing Hydride Stability in U-ZrH x Nuclear Fuel: The “Goldilocks Radius”

Nuclear-powered microreactors show great promise for opening new nuclear energy markets due to the flexibility offered by their rapid/streamlined in-factory fabrication, transportability, and self-regulating nature. The economic benefits of any commercialized nuclear reactor, however, rely on the system’s ability to produce large amounts of heat and efficiently convert that heat into electrical power reliably for long periods of time. Uranium-zirconium hydride (U-ZrH x ) is currently being considered for compact reactor designs because it is a well-known nuclear fuel system that is self-moderating, but this fuel, which has historically been used for research reactors, has not been optimized for commercial power production. Here, this paper analyzes the hydride stability of standard 304 stainless steel–clad U-ZrH x fuel under commercially relevant conditions. Fuel element design parameters, including physical dimensions, as-fabricated hydrogen content, burnup, peak fuel temperature, temperature gradient, operational fuel cycle duration, and volumetric heat generation rate, are discussed with a focus on hydrogen distribution and phase stability within the fuel element. Hydride stability declines more rapidly as the coolant temperature, burnup, and fuel cycle duration increase. Using a fuel-cladding gap material with heat transfer properties superior to air, such as helium or sodium, is essential to prolonging fuel hydride stability. The fuel’s physical dimensions are also important. At very small fuel diameters, the H/Zr ratio in the fuel meat decreases too rapidly due to the hydrogen content’s dependence on fuel meat volume. Conversely, the fuel meat temperature and temperature gradient exacerbate hydrogen loss at very large fuel diameters. We find that the most important parameter to consider when optimizing the hydride stability of U-ZrH x fuel is the relationship between the fuel meat radius and the power density in the fuel. A simple equation is empirically determined that relates the “Goldilocks radius,” that is, the fuel radius for which the H/Zr ratio is most stable, to the power density in the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Ceramic Composite Inert Matrix Fuel Forms in High-Temperature Gas-Cooled Microreactors

Here, this work optimizes micro-prismatic high-temperature gas reactor (HTGR) designs to reduce the energy-normalized mass of spent nuclear fuel (SNF) and high-level waste (HLW) produced. The optimization was performed for the current graphite moderator and an inert matrix fuel (IMF) concept employing different composite moderators in a prismatic design architecture. The fuel matrix is magnesium oxide (MgO) with entrained tristructural-isotropic (TRISO) fuel. The moderator materials, including beryllium oxide (MgO-BeO) and beryllium (MgO-Be) at 40 vol % loading and yttrium hydride (MgO-YH x=1.9 ) and zirconium hydride (MgO-ZrH x=1.9 ) at 15 vol % loading, were entrained within the MgO host matrix. A generic graphite micro-prismatic HTGR is used as the baseline point design where the external dimensions are held constant. The composite moderator designs use 19.9% enriched uranium nitride TRISO fuel and hexagonal assemblies. For each IMF concept, an optimization study was performed to maximize the discharge burnup of the fuel by varying the TRISO packing fraction and the lattice pitch of the assemblies. The mass of SNF and HLW, other waste metrics, fuel cost, environmental impact metrics, and the activity of the SNF and HLW at 100 years and 100 000 years were calculated for the optimized IMF and graphite reference designs. The IMF results were subsequently compared to those of the graphite reference and the values for a light water reactor (LWR) and a small modular LWR. For the SNF and HLW, all the IMF concepts and the graphite reference produced less waste compared to the traditional LWR designs. However, the IMF concepts outperformed the graphite reference regarding the mass of SNF and HLW. For the other waste metrics, the IMF concepts showed reductions in fuel cost with improved environmental metrics relative to the graphite reference. Overall, the IMF concepts significantly reduced the SNF and HLW produced per unit of energy generated compared to traditional LWR designs.

TRISO↗