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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 361 records · Page 20
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Microstructure, Thermal, and Mechanical Properties Relationships in U and UZr Alloys (Final Report)

Uranium-zirconium (U-Zr) alloys are candidate fuel systems for transmutation based reactors that can be used to burn long-lived minor actinides and fission products in fast spectrum reactors. Metallic fuels have also been gaining more recent attention for applications as accident tolerant fuels, as well as for use in small modular reactors. This research focused on a “science-based” approach to capture the connections between U and U-Zr alloys’ three-dimensional (3-D) microstructure, thermal properties, and mechanical properties through closely coordinated experiments and modeling efforts from the unirradiated to the irradiated fuels. Advanced characterization and modeling techniques were used to understand irradiation-induced microstructural evolution and its direct impact on the thermal and mechanical properties of U and U-Zr fuel. Closely coordinated experiments and modeling were performed to provide crucial data that does not currently exist. Overall, this research spanned multiple length and time scales within the models and experiments. The scope of the research encompassed the understanding of the irradiation effects in U and various U-Zr alloys with particular attention paid to three task areas: (1) microstructural evolution, (2) in-situ/ex-situ thermal and mechanical properties, and (3) multiscale modeling connections to microstructure, thermal, and mechanical properties. This research resulted in (1) the 3-D characterization of neutron irradiated U-Zr fuel in multiple phase regions to better understand fission gas swelling and constituent redistribution, (2) development of a microstructural model linking thermal and mechanical properties via in situ Raman and nanoindentation, (3) and mesoscale phase field modeling was coupled with the AEH method in the MOOSE framework was used to calculate the effective thermal conductivities of U-Zr fuels consisting of α-U and δ-UZr 2 heterogeneous microstructures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Microstructural evolution of Mo-UO 2 cermets under high temperature hydrogen environments

Ceramic-metallic (cermet) materials show promise for use in nuclear thermal propulsion applications due to attractive thermophysical properties including high temperature stability and high thermal conductivity. In this work, molybdenum-uranium dioxide (Mo-UO 2 ) cermet fuel elements were fabricated by means of spark plasma sintering (SPS) and were subsequently exposed to hydrogen at high temperatures (2500 K). Mo-UO 2 samples pre- and post-exposure were characterized by means of optical microscopy, scanning electron microscopy, and X-ray diffraction (XRD). Microscopy analyses of the as-produced material displayed microscopic cracking on the interior of the spherical UO 2 fuel particles but confirmed that the fuel particles were fully encapsulated in the Mo matrix. The results further showed mass loss, macroscopic swelling, and cracking in the cermet samples which occurred during high temperature hydrogen testing. Nanoscale swelling was evidenced by XRD in the Mo matrix and UO 2 fuel structure due to the incorporation of defects and accompanied microstrain.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Feasibility of Recycling Discharged Microreactor Heavy Metal in Light Water and Sodium-Cooled Fast Reactors: A Neutronics Analysis

Nuclear microreactors (MRs) offer unique advantages, such as rapid deployment, potability, low maintenance requirements, and operational flexibility. Their compact size makes them a promising solution for decentralized power generation, particularly in remote areas, military bases, and disaster-stricken regions. However, MRs face challenges, including unutilized fissile material at the end of life, economic inefficiency, increased heavy metal (HM) waste complicating disposal, and the accumulation of plutonium (Pu) with high 239 Pu concentrations raising proliferation risks. Here, this study investigated the neutronics feasibility of a novel three-stage fuel cycle where discharged HM from MRs is recycled and burned in light water reactors and sodium-cooled fast reactors. This approach converts discharged HM into valuable fuel, enhancing the efficiency of MR deployments while improving the safeguardability of their final waste products. Neutronics analysis demonstrated that the safety characteristics of reactor designs in each stage were minimally impacted by the proposed cycle. For two representative MR designs, a fast-spectrum MR with solid pellet fuel and a thermal-spectrum MR with TRISO (TRi-structural-ISOtropic) fuel compacts, the proposed fuel cycle reduced the uranium disposal mass flow rate by ~60%, decreased the 235 U enrichment of the discharge fuel to ~1 wt%, eliminated plutonium disposal, and increased the cumulative fuel burnup to ~580 gigawatt-day per metric ton of initial heavy metal (GWd/t-iHM) or 60% fissions per initial metal atom. Despite the significant differences between the two MR designs, the performance and infrastructure requirements of the developed fuel cycles were remarkably similar, indicating its generalizability to a broader class of MRs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Actinide Separation Inspired by Self-Assembled Metal–Polyphenolic Nanocages

The separation of actinides has a vital place in nuclear fuel reprocessing, recovery of radionuclides, and remediation of environmental contamination. Here we propose a new paradigm of nanocluster-based actinide separation, namely, nanoextraction, that can achieve efficient sequestration of uranium in an unprecedented form of giant coordination nanocages using a cone-shaped macrocyclic pyrogallol[4]arene as the extractant. The U 24 -based hexameric pyrogallol[4]arene nanocages with distinctive [U 2 (PG) 2 ] binuclear units (PG = pyrogallol) that rapidly assembled in situ in monophasic solvent were identified by single-crystal X-ray diffraction, MALDI-TOF mass spectrometry, NMR spectroscopy, and small-angle X-ray and neutron scattering. Furthermore, comprehensive biphasic extraction studies showed that this novel separation strategy has enticing advantages such as fast kinetics, high efficiency, and good selectivity over lanthanides, thereby demonstrating its potential for efficient separation of actinide ions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Hyperfine structures and isotopic shifts of uranium transitions using tunable laser spectroscopy of laser ablation plumes

We report isotopic shifts and hyperfine structures of selected U transitions employing tunable spectroscopy viz: laser-induced fluorescence and laser absorption spectroscopy of laser ablation plumes. The laser-produced plasmas were produced during ns laser ablation on a natural U metal which contains 0.7% 235U. Our results show that isotopic shifts between 238U and 235U are entangled with hyperfine structures of 235U. The isotopic ratio obtained using laser-induced fluorescence is affected by optical saturation effect. Time-resolved laser absorption spectroscopy is carried out for evaluating the optical absorption and estimating the hyperfine constants.

(020.3260) Isotope shifts↗
DOE OSTI2020

Structural features of solid-solid phase transitions and lattice dynamics in U 3 O 8

Triuranium octoxide (U 3 O 8 ) undergoes an orthorhombic to hexagonal structural phase transition near Ts=305°C, and a separate nonstructural phase transition at Tc=210°C. The later transition has previously been associated with temperature-induced fluctuations in the uranium oxidation state. A discontinuity in the slope of electrical conductivity versus temperature measurement at 210°C has supported this idea. The orthorhombic phase has three crystallographic sites in two distinct oxidation configurations [2 U(V) and 1 U(VI)], whereas the hexagonal phase has one distinct uranium site. High-resolution x-ray diffraction measurements eliminate the possibility of superlattice Bragg reflections to less than 0.2 e - scattering power and U 3 O 8 is not metallic; consequently, the presence of oxidation fluctuations is required for charge balancing. Interestingly, the order-to-disorder transition occurs at a much lower temperature than the structural transition. Using temperature-dependent x-ray diffraction and Raman spectroscopy, we show anisotropic lattice expansion in the in-plane b and c lattice constants. A specific discontinuity in the temperature derivatives of the b and c lattice constants are the first reported structural signatures of the order-to-disorder transition, suggestive of a change in local U–O coordination. Phonon frequencies of U 3 O 8 measured by Raman spectroscopy show significant temperature-dependent dynamics. Redshifting of several modes between 40 and 300°C cannot be explained by unit cell expansion alone because the unit cell volume decreases in this region. Instead, we show that phonon frequencies are highly correlated with the anisotropic lattice expansion/contraction along specific crystallographic directions.

36 MATERIALS SCIENCE↗
DOE OSTI2023

A chemistry-informed hybrid machine learning approach to predict metal adsorption onto mineral surfaces

Historically, surface complexation model (SCM) constants and distribution coefficients (K d ) have been employed to quantify mineral-based retardation effects controlling the fate of metals in subsurface geologic systems. Our recent SCM development workflow, based on the Lawrence Livermore National Laboratory Surface Complexation/Ion Exchange (L-SCIE) database, illustrated a community FAIR data approach to SCM development by predicting uranium(VI)-quartz adsorption for a large number of literature-mined data. Here, we present an alternative hybrid machine learning (ML) approach that shows promise in achieving equivalent high-quality predictions compared to traditional surface complexation models. At its core, the hybrid random forest (RF) ML approach is motivated by the proliferation of incongruent SCMs in the literature that limit their applicability in reactive transport models. Our hybrid ML approach implements PHREEQC-based aqueous speciation calculations; values from these simulations are automatically used as input features for a random forest (RF) algorithm to quantify adsorption and avoid SCM modeling constraints entirely. Named the LLNL Speciation Updated Random Forest (L-SURF) model, this hybrid approach is shown to have applicability to U(VI) sorption cases driven by both ion-exchange and surface complexation, as is shown for quartz and montmorillonite cases. The approach can be applied to reactive transport modeling and may provide an alternative to the costly development of self-consistent SCM reaction databases.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Amidoxime-functionalized materials and their use in extracting metal ions from liquid solutions

A method of producing a functionalized material that extracts metal ions from solution, the method comprising: (i) providing a precursor material having nitrile groups appended to its surface; and (ii) reacting said nitrile groups with hydroxylamine or a derivative thereof in the presence of a polar aprotic solvent at a temperature of 60-80° C. for at least 1 hour, to convert at least a portion of said nitrile groups to amidoxime and imide dioxime groups, followed by reaction with a base capable of hydrolyzing any remaining nitrile groups to carboxylic acid groups; wherein said functionalized material has a higher uranium absorption capacity than a functionalized material produced under same conditions except that the nitrile groups are reacted with hydroxylamine in only a protic solvent. The invention is also directed to functionalized materials produced by the above-described method, and methods for using the functionalized material for extracting metal ions from metal-containing solutions.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Amidoxime-functionalized materials and their use in extracting metal ions from liquid solutions

A method of producing a functionalized material that extracts metal ions from solution, the method comprising: (i) providing a precursor material having nitrile groups appended to its surface; and (ii) reacting said nitrile groups with hydroxylamine or a derivative thereof in the presence of a polar aprotic solvent at a temperature of 60-80° C. for at least 1 hour, to convert at least a portion of said nitrile groups to amidoxime and imide dioxime groups, followed by reaction with a base capable of hydrolyzing any remaining nitrile groups to carboxylic acid groups; wherein said functionalized material has a higher uranium absorption capacity than a functionalized material produced under same conditions except that the nitrile groups are reacted with hydroxylamine in only a protic solvent. The invention is also directed to functionalized materials produced by the above-described method, and methods for using the functionalized material for extracting metal ions from metal-containing solutions.

Janke, Christopher J.↗
DOE OSTI2020

Fabrication of UN-Mo CERMET Nuclear Fuel Using Advanced Manufacturing Techniques

Ceramic-metallic nuclear fuels are a candidate fuel for nuclear thermal propulsion systems due to their high heat transport properties, which are necessary in very high-temperature environments. The conventional fabrication of uranium nitride–molybdenum fuel has been thoroughly studied in the past, but modern manufacturing techniques have presented a unique opportunity for further development within this field. This work demonstrates the use of advanced manufacturing techniques to produce nuclear fuel pellets composed of uranium nitride microspheres encased in a molybdenum matrix. Binder jetting is used to print molybdenum disks that are filled with uranium nitride microspheres and afterward sintered using spark plasma sintering. Two fuel pellets were fabricated to demonstrate the methodology and to provide a baseline analysis of the effects of temperature and pressure processing conditions. Characterization of the sintered fuel pellets includes detailed microstructural analysis and thermal conductivity measurements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Design of a first-of-A-kind instrumented advanced test reactor irradiation Capsule experiment for In situ thermal conductivity measurements of metallic fuel

Metallic fuel undergoes dramatic microstructural changes early in life due to fission gas swelling until ~2–3 at% burnup which affects the conductivity of the material, however the evolution of metallic fuel thermal conductivity during this early phase burnup has never been successfully measured in situ. The Irradiated Material Properties Accelerated Characterization Test (IMPACT) experiment will be the first in a series of experiments to irradiate advanced nuclear metallic fuel specimens with novel embedded thermal conductivity probes in ATR. In the current work the IMPACT experiment final design and supporting analysis is reported in detail. Results are evaluated for various reactor operational conditions to meet the functional requirements of the experiment. Finally, the first iteration of this IMPACT experiment will provide data regarding thermal properties evolution in uranium-zirconium (U10Zr) fuel, but this experiment vehicle is envisioned for future advanced fuels and structural materials irradiations in ATR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Groundwater characterization and monitoring at a complex industrial waste site using electrical resistivity imaging

A contaminated industrial waste site in Washington State (USA) containing buried, metallic-waste storage tanks, pipes, and wells, was evaluated to determine the feasibility of monitoring groundwater remediation activities associated with an underlying perched aquifer system using electrical resistivity tomography. The perched aquifer, located ~65 m below ground surface and ~10 m above the regional water table, contains high concentrations of nitrate, uranium, and other contaminants of concern from past tank leaks and intentional releases of wastes to surface disposal sites. The extent of the perched water aquifer is not well known, and the effectiveness of groundwater extraction for contaminant removal is uncertain, so supplemental characterization and monitoring technologies are being evaluated. Numerical simulations of subsurface flow and contaminant transport were performed with a highly resolved model of the hydrogeologic system and waste site infrastructure, and these simulations were used as the physical basis for electrical resistivity tomography modeling. The modeling explicitly accounted for metallic infrastructure at the site. The effectiveness of using surface electrodes versus surface and horizontal subsurface electrodes, for imaging groundwater extraction from the perched water aquifer, was investigated. Although directional drilling is a mature technology, its use for electrode emplacement in the deep subsurface under a complex industrial waste site via horizontal wells has not yet been demonstrated. Results from this study indicate that using horizontal subsurface electrode arrays could significantly improve the ability of electrical resistivity tomography to image deep subsurface features and monitor remediation activities under complex industrial waste sites.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

A Uranium(II) Arene Complex That Acts as a Uranium(I) Synthon

We report two-electron reduction of the amidate-supported U(III) mono(arene) complex U(TDA) 3 (2) with KC 8 yields the anionic bis(arene) complex [K[2.2.2]cryptand][U(TDA) 2 ] (3) (TDA = N-(2,6-di-isopropylphenyl)pivalamido). EPR spectroscopy, magnetic susceptibility measurements, and calculations using DFT as well as multireference CASSCF methods all provide strong evidence that the electronic structure of 3 is best represented as a 5f 4 U(II) metal center bound to a monoreduced arene ligand. Reactivity studies show 3 reacts as a U(I) synthon by behaving as a two-electron reductant toward I 2 to form the dinuclear U(III)–U(III) triiodide species [K[2.2.2]cryptand][(UI(TDA) 2 ) 2 (μ-I)] (6) and as a three-electron reductant toward cycloheptatriene (CHT) to form the U(IV) complex [K[2.2.2]cryptand][U(η 7 -C 7 H 7 )(TDA) 2 (THF)] (7). The reaction of 3 with cyclooctatetraene (COT) generates a mixture of the U(III) anion [K[2.2.2]cryptand][U(TDA) 4 ] (1-crypt) and U(COT) 2 , while the addition of COT to complex 2 instead yields the dinuclear U(IV)–U(IV) inverse sandwich complex [U(TDA) 3 ] 2 (μ-η 8 :η 3 -C 8 H 8 ) (8). Two-electron reduction of the homoleptic Th(IV) amidate complex Th(TDA) 4 (4) with KC 8 gives the mono(arene) complex [K[2.2.2]cryptand][Th(TDA) 3 (THF)] (5). The C–C bond lengths and torsion angles in the bound arene of 5 suggest a direduced arene bound to a Th(IV) metal center; this conclusion is supported by DFT calculations.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2026

Radiation Effects in Next Generation Used Nuclear Fuel Reprocessing Strategies

With the global community committed to significantly expanding nuclear energy capacity, the development of efficient used nuclear fuel (UNF) management strategies has become more critical than ever. These strategies are vital to fostering the widespread adoption of closed fuel cycles, which are essential for sustainable nuclear energy production and security. Achieving this ambitious goal necessitates a comprehensive understanding of radiation effects on next-generation technologies, as radiolysis can often limit the longevity and performance of these systems. This seminar will provide an overview of next-generation UNF reprocessing strategies, highlighting the latest advancements and innovative approaches in the field. Particular attention will be given to two key areas of recent research: 1. Radiation robustness and performance of advanced sulfur chloride-based chlorination technologies. We will explore the efficacy of sulfur chloride-based chlorination processes in the presence of surrogate cladding materials, specifically aluminum. These processes have shown promise in the dissolution, decontamination, and recovery of cladding materials for reuse. Detailed findings on how the composition and performance of these sulfur chloride solvents respond to radiation exposure will be discussed. 2. Impacts of metal ion complexation and direct dissolution conditions on monoamide-based reprocessing strategies. We will delve into the time-resolved and dose accumulation effects of irradiation on the direct dissolution of voloxidized uranium and rhenium using N,N-di-(2-ethylhexyl) butyramide (DEHBA) or N,N-di-(2-ethylhexyl)isobutyramide (DEHiBA) in pre-equilibrated n-dodecane solvent. The implications of these interactions on dissolution efficiency, radiolytic stability, and overall process performance will be examined. These studies aim to underscore the importance of understanding radiation effects in the development of next-generation UNF reprocessing technologies and the global transition towards more sustainable and efficient nuclear energy systems.

38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗
DOE OSTI2023

Pros and Cons Analysis of HALEU Utilization in Example Fuel Cycles

The Systems Analysis and Integration campaign assessed the pros and cons of high-assay low-enriched uranium (HALEU) utilization in advanced reactors and associated fuel cycles. The assessment was done for three example fuel cycles at equilibrium states: once-through, limited recycle, and continuous recycle (CR) starting with HALEU. Front- and back-end fuel cycle parameters and the Levelized Cost of Fuel (LCF), which is the Levelized Cost of Electricity excluding reactor cost, of the three example fuel cycles were calculated using a single Analysis Example Reactor. The pros and cons of HALEU utilization were assessed by normalizing the fuel cycle parameters and LCF to a unit of electricity generation (GWe-year) and comparing them with a Basis of Comparison. In this study, a sodium-cooled reactor with sodium-bonded metallic fuel having a burnup of ~100 GWd/t was used as the Analysis Example Reactor because its technology readiness level is high, and the burnup and fuel enrichment are in the middle of those ranges of advanced reactor concepts that are under development. The current once-through Light Water Reactors (OT-LWRs) with <5% low-enriched uranium and 50 GWd/t burnup were used as the Basis of Comparison. In addition, a series of sensitivity analyses was conducted by varying burnup, enrichment, fuel forms, and reactor types to capture the design variations in two once-through Advanced Reactor Demonstration Program (ARDP) reactors, Natrium with sodium-free metallic fuel having a burnup of ~150 GW/t and Xe-100 with Tristructural-Isotropic (TRISO) pebble fuel having a burnup of ~168 GWd/t.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Comparison of Nuclear Thermal Propulsion Reactor Cores Using Different Fuel Forms, Neutron Spectra, and Uranium Enrichments

A comparison of 300 MWt HEU and HALEU NTP cores using different fuel forms is presented to demonstrate trends and potentially inform fuel development. Compared to HEU, HALEU cores require additional in-core moderation. Beryllium was chosen as the reference in-core moderator material; metal hydrides may also be considered. Typical NTP fuel forms such as carbide and carbonitride solid solutions, carbide-graphite composites, CERMETs and CERCERs were considered together with less typical fuel forms such as UC and U metal. Be-moderated HALEU cores are heavier than HEU cores (1.9-2.7 and 0.9-2.0 metric tons, respectively). Furthermore, HALEU loading is very sensitive to the amount of in-core moderation which translates into tight fabrication tolerances. HALEU cores requires more reactivity control mechanisms than HEU cores to compensate for temperature feed-back. Using enrichments only slightly above HALEU (e.g., 25%) may provide a compromise between the administrative preference for using HALEU and the engineering preference for simplifying the reactor systems by, for example, not using in-core moderators.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Impact of UV-light and pH on the Fate of Tc, I, and U in Wetlands at Savannah River Site - 20230

The Savannah River Site (SRS) is one of the major nuclear facilities owned by the U.S Department of Energy. During the Cold War, these facilities produced large amounts of radioactive and hazardous waste. On site, three unlined seepage basins in the F-Area received approximately 1.8 billion gallons of low-level waste containing nitric acid, radionuclides, and dissolved metals due to plutonium and tritium production operations. The acidic nature of the waste created a source of relatively mobile radionuclides below the basins. Radionuclides previously disposed of within the F-Area, including uranium isotopes (U), technetium-99 (Tc-99), and iodine-129 (I-129), are moving with groundwater towards Four Mile Branch Wetland, where they are subsequently upwelling and interacting with natural organic matter (NOM). Many environmental factors including redox conditions, porewater composition, pH, light, and temperature affect the degradation of organic matter as well as interactions with Tc-99, U, and I-129. In particular, the high concentrations of nitrate from the acidic waste may increase the formation of reactive oxygen species (ROS) that impact both degradation of NOM and behavior of contaminants. In the presence of sunlight, I-129, Tc-99, and U speciation and complexation behavior may be affected by ROS and NOM degradation products in addition to the factors generally considered in subsurface systems in the absence of light. This research aims to determine whether the interactions between radionuclides, NOM, and nitrate affect the fate of I-129, Tc-99, and U and which processes are controlling their behavior. Experiments were conducted at variable pH in the presence of NaNO{sub 3} in order to determine the impact of light and pH on NOM degradation and to evaluate the impact on the fate of contaminants of concern. Soil samples high in NOM collected from two sites in the Southeastern United States (Savannah River Site and the Everglades) were studied. Batch experiments were conducted with NOM and the following aqueous contaminants, U, I-129, Tc-99 with exposure to ultraviolet (UV) light in an environmental chamber. This study was conducted in a sterile environment in order to exclude the potential for microbial degradation of organic matter. Results indicated photodegradation of NOM and significant interaction of radionuclides with NOM. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Advanced Fuels Campaign Execution Plan

The Advanced Fuels Campaign (AFC) Execution Plan details the strategy, mission, scope, and goals—both near-term and long-term—along with the structure and organization of nuclear fuels and materials research, development, and demonstration (RD&D) activities within the Fuel Cycle Technologies (FCT) program. The FCT program, tasked by the U.S. Department of Energy (DOE), employs a science-based approach to advance fuel technologies. This approach integrates theory, experiments, and multi-scale modeling and simulation (M&S) to develop a predictive understanding of fuel fabrication processes and fuel/cladding performance under irradiation, moving beyond traditional empirical methods. The long-term goals of the AFC are guided by the AFC Strategic Plan and align with the DOE Office of Nuclear Energy (NE) Roadmap [1], which outlines a multi-decade vision for demonstrating and qualifying advanced fuel forms to support diverse fuel cycle options. Near-term goals focus on enhancing accident tolerant fuels (ATF) for Light Water Reactors (LWR), a significant challenge that demands balancing immediate objectives with ongoing progress toward advanced reactor missions. Accelerating the traditional fuel qualification process to meet ATF objectives is another critical challenge. A detailed set of 5-year goals, summarized below, has been developed in line with the overarching science-based fuel development approach: • Advanced LWR Fuel Technologies: By 2027, support the development of advanced LWR fuel technologies with improved performance and enhanced accident tolerance. This includes high burnup (HBu), low enriched uranium (LEU)+, coated cladding, and doped fuel, aimed at complementing industry-led significant LWR uprates and plant refurbishments. • Tristructural Isotropic (TRISO) Fuel: Achieve qualification by 2028 and develop improved designs for emerging markets. • Metal Fuel: Achieve qualification by 2028 and develop improved designs for emerging markets. • Molten Salt Fuel: By 2027, deploy a robust program that enables fuel salt qualification technologies needed to support fuel salt research and development (R&D), focusing on emergent needs to derisk fuel salt production and utilization in advanced reactors. • Long-Term ATF: Develop fuel technologies that enable significant power uprates (~50%) in refurbished or new LWRs while optimizing fissile material utilization and waste disposal. The 5-year milestones in the AFC Execution Plan are contingent on an assumed budget. This Execution Plan will be updated annually to reflect actual funding profiles as budget guidance becomes available, ensuring milestones are adjusted accordingly. In summary, the AFC Execution Plan presents a comprehensive strategy to advance nuclear fuel technologies through a science-based approach, addressing both near-term and long-term goals while adapting to funding realities.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗