Search NASA⌕ Search

SEARCH · Search NASA

Results for “Fuel Fabrication”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Electropolishing of a Full-Sized U-10Mo Plate

Electropolishing is used to remove material from the surface of a metal using electric potential and current. U-10Mo fuel is produced from a low enriched uranium plate alloyed with 10% molybdenum (U-10Mo). Electropolishing is being considered in two steps of the process of fabricating this U-10Mo fuel. First, it could be used as a means initially improving the surface finish of the cast plate while removing smut from the casting process from the surface prior to homogenization. Second, it could be used again after homogenization to remove oxidation and continue to improve the surface finish prior to hot rolling. The purpose of this work is to demonstrate the process on full-size ingots and optimize the process. A depleted uranium U-10Mo plate was electropolished three times: the first two times were prior to homogenization and the third after a homogenization step. The first two polishes decreased the overall roughness, as measured by laser confocal microscopy. After each polish, a smoother and shinier blue-gold surface was left on the plate. Upon homogenization, the surface had a more matte appearance and the surface roughness increased back to the pre-homogenization values obtained prior to the initial polish. After electropolishing the homogenized pieces, the changes in surface roughness increased, as measured by laser confocal microscopy. The final polishing step was not long enough to return the surface roughness to pre-homogenization conditions, implying less polishing is needed before the homogenization versus afterwards. Heat generation during the polish was addressed by circulating the electrolyte solution through an external heat exchanger; however, solubility and anodization challenges require further study.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Fabrication of MiniFuel Compacts for High-Power Irradiation Testing of TRISO Fuel

MiniFuel compacts containing 20 individual tristructural isotropic (TRISO)-coated fuel particles were characterized to provide supporting preirradiation characterization data. The preirradiated characterization data supports irradiation capsule design and safety analysis, as well as downstream postirradiation examination (PIE) for a planned MiniFuel irradiation to explore high particle powers. The analysis included dimensional inspection, x-ray radiography and tomography, as-fabricated defect fraction analysis, and a matrix impurity analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Aluminum-doped U 3 Si 2 composite fuels with enhanced oxidation resistance

Al-doped U 3 Si 2 composite fuels with controlled microstructure were fabricated by spark plasma sintering that display greatly-improved oxidation resistance as compared with monolithic and Al-doped silicides prepared by standard powder metallurgy or arc melting. The effects of Al additives on the thermal-mechanical properties and oxidation resisance of the micron- and nano-sized U 3 Si 2 composites were investigated. Additionally, a minimal addition of 1.8 at% Al is effective to increase the onset oxidation temperature of as-fabricated U 3 Si 2 pellets to 580 °C, which can be further increased to 610 °C by thermal annealing. The Al-doped U 3 Si 2 composite fuels also display simultaneously higher hardness and fracture toughness than un-doped U3Si2. These results highlight an effective strategy by integrating minimal Al additives, microstructure control and post-thermal annealing to design advanced silicide fuels with excellent oxidation resistance, desired thermal-mechanical properties and maintained high fissile element density.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Residual Stress Measurements in Extreme Environments for Hazardous, Layered Specimens

In nuclear fuel plates of low-enriched U-10Mo (LEU) clad with aluminum by hot isostatic pressing (HIP), post-irradiation stresses arising during reactor shutdown are a major concern for safe reactor operations. Measurement of those residual stresses has not previously been possible because the high radioactivity of the plates requires handling only by remote manipulation in a hot cell. The incremental slitting method for measuring through-thickness stress profiles was modified, and a system for automated, remote operation was built and tested. Experimental modifications consisted of replacing electric-discharge machining (EDM) with a small end mill and strain-gauge measurements with cantilever displacement measurements. The inverse method used to calculate stresses was the pulse-regularization method modified to allow discontinuities across material interfaces. The new system was validated by comparing with conventional slitting on a depleted U-10Mo (DU) fuel plate. The new system was applied to two measurements each on six as-fabricated (pre-irradiation) LEU miniature fuel plates. Variations between the measurements at two locations in the same plate were strongly correlated with measured geometrical heterogeneity in the plate—a tilt in the fuel foil. Compressive stresses in the U-10Mo were shown to increase from 20 to 250 MPa as the ratio of aluminum thickness to U-10Mo thickness increased causing increased constraint during cooling. Faster cooling rates during processing also increased stress magnitudes. The measurements trends agreed with data in the literature from similar plates made with DU, which further validates the method. Because other methods are impractical in a hot cell, the modified slitting method is now poised for the first measurements of post-irradiation stresses.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Evaluation of Irradiation Creep Effects in HT9 Cladding for FAST Experiments

The push for advanced reactor fuels for improved reactor safety and efficiency had led to a renewed interest in metallic fuel for nuclear reactor applications. Experimental investigation is necessary to ensure a robust understanding of the thermomechanical properties of new metallic fuel designs. Unfortunately, with the current experimental facilities, thoroughly investigating the responses of metallic fuel burnup would take a prohibitively long time. To alleviate this, the Fission Accelerated Steady State Test (FAST) was developed to accelerate the irradiation testing while simultaneously decreasing the sensitivity to fabrication tolerances by reducing the fuel diameter and scaling the experiment. This method successfully scales the radiation effects on the fuel, but the HT9 cladding is not exposed to prototypic radiation conditions. This raises questions on whether the FAST experiment results are truly indicative of the HT9 cladding performance due to radiation induced creep effects not being appropriately accounted for. Using BISON fuel performance code, the simulated FAST cladding strain is compared to simulated EBR-II cladding strain. This is done through a sensitivity study of input parameters and scaling of neutron fluence on the cladding. This allows a parametric comparison of physical phenomena on the effective difference between cladding strains between FAST and equivalent burnup EBR-II fuel pins. The results show that the irradiation induced deformation (creep or swelling) is insignificant compared to the thermal-mechanical deformation. Therefore, the difference between the FAST experiment cladding and the EBR-II experiment cladding is negligible and comparison of fuel system performance between the two experiments is appropriate.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Dissolution Flowsheet for Skull Oxide Generated during U-Mo Alloy Casting for High Performance Research Reactor Fuel

The Savannah River National Laboratory was requested to develop dissolution flowsheets for high assay low enriched U scrap generated during the fabrication of high performance research reactor fuel. The scrap streams include U-10Mo and U-10Mo-Zr foils, rejected Al-clad fuel plates, and skull oxide from casting molds. Flowsheets for the dissolution of the U-10Mo-Zr foils and Al-clad U-10Mo-Zr mini-plates received from BWX Technologies, Inc were developed and demonstrated in the first phase of this project. In the second phase (this work), the skull oxide from a U-10Mo casting mold was obtained from the Y-12 National Security Complex (Y-12) and small-scale experiments were performed to demonstrate dissolution flowsheets.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Molybdenum Sleeve Experiments in Fully-Reflected Water-Moderated Triangular-Pitched U(6.90)O 2 Fuel Rod Lattices (1.55 cm Pitch)

The US Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories. The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5 % 235 U. The 7uPCX assembly is a water-moderated and -reflected array of aluminum-clad U(6.90 %)O 2 fuel rods. Other critical experiments performed in the 7uPCX assembly are documented in LEU-COMP-THERM-078, LEU-COMP-THERM-080, LEU-COMP-THERM-096, LEUCOMP-THERM-097, LEU-COMP-THERM-101, and LEU-COMP-THERM-102. The purpose of these experiments was to measure the effects of molybdenum in nearly-critical systems. The molybdenum was introduced into the fuel arrays as tubular sleeves that surrounded some of the fuel rods in the fuel arrays measured. Four hundred molybdenum tubes nominally 12.7 mm outside diameter, 498 mm long, with 0.762 mm wall thickness were provided for the experiments by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN). Small polyethylene adapters at each end of the tubes were used to center each tube on a fuel rod in the assembly. The critical experiments were done using a set of triangular-pitched grid plates fabricated for these experiments. The grid plate set accommodated a fuel array of a total of 1261 fuel rod positions on a pitch of 0.610 in (1.5494 cm) in a series of 20 hexagonal rings surrounding the central fuel rod. The fuel used in these experiments was fabricated using unirradiated 6.90 % enriched UO 2 fuel pellets from fuel elements designed to be used in the internal nuclear superheater section of the Pathfinder boiling water reactor operated in South Dakota by the Northern States Power Company in the 1960s. The fuel elements were obtained from The Pennsylvania State University where they had been stored for many years. The fuel pellets in those fuel elements were removed from the original Incoloy cladding and reclad in 3003 aluminum tubes and end caps for use in the experiments reported here. The five critical experiments in this series were performed in August through December 2022, in the Sandia Critical Experiments (SCX) at the Sandia Pulsed Reactor Facility. Case 1 had no molybdenum sleeves, Case 2 had 208 molybdenum sleeves clustered at the center of the array, Case 3 had 397 molybdenum sleeves clustered at the center of the array, Case 4 had 175 molybdenum sleeves in the central position and in five alternating hexagonal rings, and Case 5 had 331 molybdenum sleeves in the central position and in seven alternating hexagonal rings. All five critical experiments are judged to be acceptable as benchmark experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Survey of Uranium Nitride and Mixed Oxide Fuels for Microreactor Applications

Different fuel systems are being proposed for use in microreactors including TRISO fuel and metallic U-Zr based fuel. TRISO fuel is considered the main fuel option for most of the industry teams. Given the need for compact size core and potential for long fuel life of a microreactor, high density fuels such as metallic and nitride fuels are potential options. Of interest here is the uranium nitride (UN) option, which can allow for higher fissile material content and higher thermal conductivity compared to conventional uranium oxide fuel, and also has higher melting temperature compared to metallic fuel. Meanwhile, the current availability is limited for high assay low enriched uranium (HALEU) that is needed for high density fuels, which motivates the consideration of using Pu as a potential replacement for HALEU until adequate production capacity is in place. Current Pu availability is mainly attributed to the inventory of excess weapons Pu rather than through reprocessing of spent nuclear fuel. This inventory of excess Pu can be used in both metallic and oxide fuel systems to replace HALEU. Of interest here, the oxide form, that is the mixed oxide form of PuO 2 and UO 2 (MOX). In this report, the options of using UN or MOX, in both pellet and TRISO fuel forms, in microreactors are evaluated in relation to their properties, fuel performance and irradiation data, as well as fabrication. Gaps related to those areas are identified for both fuel systems, to guide future activities by DOE programs such as the advanced fuels campaign (AFC), to enable their use in microreactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Transport Modeling of As-Run ATR Cycles to Support U-10Mo Research Reactor Fuel Qualification Experiment

The Department of Energy’s (DOE) Office of Materials Management and Minimization has been tasked with converting the five remaining United States High Performance Research Reactors (U.S. HPRR) from highly enriched uranium to low-enriched uranium. The Nuclear Regulatory Commission (NRC) regulates Massachusetts Institute of Technology Reactor (MITR), Missouri University Research Reactor (MURR), and National Bureau of Standards Reactor (NBSR); and DOE regulates the High Flux Isotope Reactor and Advanced Test Reactor (ATR). To meet the high demands of these reactors, the U.S. HPRR program has chosen to use 90% uranium - 10% molybdenum (U-10Mo) monolithic fuel. This plate-type fuel will undergo multiple irradiation experiment campaigns in ATR, from mini-plates to full element tests, over a large range of operating conditions. This will provide data in support of the fuel qualification of each reactor. This summary focuses on the as-run neutronic analysis of the first series of mini-plate (MP-1) experiments, which have been irradiated in the ATR. MP-1 experiment’s main goal is to demonstrate the fabrication process and meet the fuel irradiation performance requirements with primary focus on the fuel plates associated with the three NRC reactors, with a small focus on the low power requirements of the ATR fuel. The MP-1 experiments are planned to be the basis of the monolithic U-10Mo fuel for qualification through the NRC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Transport Modeling of As-Run ATR Cycles for U-10Mo Fuel Qualification Experiment

The Department of Energy’s (DOE) Office of Materials Management and Minimization has been tasked with converting the five remaining United States High Performance Research Reactors (U.S. HPRR) from highly enriched uranium to low-enriched uranium. The Nuclear Regulatory Commission (NRC) regulates Massachusetts Institute of Technology Reactor (MITR), Missouri University Research Reactor (MURR), and National Bureau of Standards Reactor (NBSR); and DOE regulates the High Flux Isotope Reactor and Advanced Test Reactor (ATR). To meet the high demands of these reactors, the U.S. HPRR program has chosen to use 90% uranium - 10% molybdenum (U-10Mo) monolithic fuel. This plate-type fuel will undergo multiple irradiation experiment campaigns in ATR, from mini-plates to full element tests, over a large range of operating conditions. This will provide data in support of the fuel qualification of each reactor. This summary focuses on the as-run neutronic analysis of the first series of mini-plate (MP-1) experiments, which have been irradiated in the ATR. MP-1 experiment’s main goal is to demonstrate the fabrication process and meet the fuel irradiation performance requirements with primary focus on the fuel plates associated with the three NRC reactors, with a small focus on the low power requirements of the ATR fuel. The MP-1 experiments are planned to be the basis of the monolithic U-10Mo fuel for qualification through the NRC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Effects of a refractory metal protective barrier on uranium carbonitride ceramic–metallic composites exposed to a high temperature hydrogen environment

Uranium nitride has been under consideration at NASA and the Department of Energy for nuclear thermal propulsion applications intended for deep space exploration vehicles. It is critical to the success of the technology development that the behavior of the fuel system is properly characterized in operational environments. This body of work fabricated ceramic metal composites consisting of uranium ceramic fuel in a molybdenum and molybdenum–tungsten metal matrix, both with and without an exterior molybdenum barrier. Specimens were introduced to a high temperature hydrogen environment and characterized through x-ray diffraction, scanning electron microscopy, and energy dispersive spectroscopy to identify characteristic changes. A portion of the specimens was fabricated with a 1.27 mm thick molybdenum barrier in order to isolate the material from the hydrogen environment. Instability of the uranium nitride occurred as low as 2073 K, leading to aggressive instability approaching 2658 K. Carbon reactions with the metal matrix were observed as well as alloying between the uranium and metal matrix constituents. The liquid uranium resulting from the instability of the uranium nitride was observed to attack the grain boundaries of the refractory metal, resulting in the formation of a liquid uranium molybdenum eutectic melt. Specimens encased in a barrier experienced similar results. The authors conclude this is a significant challenge to applications of this cermet at these temperatures. Alternate configurations with resilient hermetic seals or lower temperature applications such as power reactors may have more success.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Destructive Analysis of TRISO Particles: Crush/Burn/Leach Followed by Davies-Gray Titration and IDMS

The accurate accounting of nuclear materials is a cornerstone of international nuclear safeguards. One emerging challenge in this domain is the fabrication of TRIstructural ISOtropic (TRISO) particle fuels. Although these innovative fuel forms are critical for advanced reactor applications, their robust refractory ceramics and coating compositions present significant obstacles to destructive analysis (DA) methods. Ensuring full and quantitative recovery from these particles is essential for accurate mass accountancy. The current study was initiated to address these challenges, first by validating a previously established destructive method developed by Oak Ridge National Laboratory (ORNL) for the quantitative recovery of uranium from TRISO particles and then following that process with uranium content determination through isotope dilution mass spectrometry (IDMS) and Davies-Gray titration. This study expands on the scope of a digestive method that was developed under the Advanced Gas Reactor Fuel Development and Qualification program and is currently implemented in both the Coated Particle Fuel Development Laboratory and Irradiated Fuels Examination Laboratory at ORNL. The success of the previous Advanced Gas Reactor work relied on developing a DA method to evaluate the fabrication process and reactor experiments. The methodology described in this report was designed to rigorously investigate the efficacy of the crush/burn/leach sample preparation of TRISO particles; it aims to quantify uranium recovery while also assessing the effects of TRISO constituents (e.g., silicon and zirconium) on analytical precision and accuracy. By comparing the results from the titration method and IDMS, we sought to determine whether existing analytical procedures accepted by the International Atomic Energy Agency (IAEA) could be effectively translated to TRISO fuel forms. The team employed an approach that involved processing replicate TRISO samples, optimizing the milling (i.e., crushing) step and performing serial leaches. The elemental composition of the analytical samples was examined to prepare for interference studies in the second year of this project. The integration of gamma spectrometry to verify residual uranium activity further strengthened the validation. Statistical methods were applied to the collected data to evaluate the uncertainties arising from sampling, sample preparation and uranium quantification. These uncertainties were then compared to the IAEA’s international target values (ITVs). Additional data collected in upcoming project work will strengthen the uncertainty estimates. Ultimately, it is hoped that this project will contribute materially to the body of work related to characterization of TRISO based fuels for the purpose of material accountancy and its applications to international safeguards.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Preliminary Benchmark Uncertainties for Deimos, a HALEU-Fueled and Graphite-Moderated Advanced Reactor Testbed

Many advanced reactor concepts will make use of various uranium fuels with levels of enrichment higher than previously seen in current light water reactors. In particular, High-Assay Low Enriched Uranium (HALEU), that is uranium enriched to 235 U ≈ 20 w/o%, is planned to be used in over ten new reactor concepts. HALEU is attractive for advanced reactors as it enables longer intervals between refueling. Unfortunately, little to no experience with HALEU is available in experimental literature raising concerns for not only licensing advanced re actors but also fabrication and transportation of HALEU fuels. This is where Deimos, a Los Alamos National Laboratory internal project, comes in. Deimos is a new critical experiment scheduled for FY24 at the National Criticality Experiments Research Center (NCERC). Deimos is a graphite moderated, graphite and beryllium reflected critical experiment making use of HALEU TRi-structural ISOtropic (TRISO) fuel from the Compact Nuclear Power System (CNPS). This transaction entails a brief description of efforts to benchmark Deimos for inclusion into the International Criticality Safety

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Low-Cost, Scalable Sorbents with Balanced Capacity-Kinetics-Thermodynamics for H2 Storage

The goal of this project is to utilize cutting-edge materials science, chemistry, and process engineering to demonstrate the feasibility of a new class of scalable sorbents with unprecedented balance of capacity-kinetics-thermodynamics for H 2 storage in fossil fuel power plants . The core fabrication leverages the unique exotic properties (e.g., high H 2 affinity, ultrahigh surface area, and exceptional thermal, mechanical and chemical stability) of the emerging 2D materials i.e. hexagonal boron nitride (hBN) followed by strategic coupling with alkali metals (Li, Na) dopants and optimal interlayer distance to deliver a stable, long-term energy storage for the grid. Figure 1 shows how this technology works.

08 HYDROGEN↗
DOE OSTI2025

Fabrication and Characterization of MiniFuel Specimen

Burnable absorbers (BAs) are used in UO 2 fuels to tune the reactivity in a light water reactor (LWR) particularly at the beginning of life of the fuel. BAs are elements that have high neutron absorption cross sections and because of this property can prevent power peaking. The addition of BAs reduces the reactivity at the beginning of life, and this in turn allows for the increase of enrichments from > 5 % 235 U up to ~ 10 %. This also enables the increase of the burnup limit to 75 GWd/MTU and the cycle length to 24 months. The BA are eventually consumed under irradiation, and their concentration is diminished at roughly 10-20 GWd/MTU. The objective of this report is to fabricate UO 2 fuels that have been doped with BAs (Gd 2 O 3 in this case) and subsequently irradiate them in the high flux isotope reactor (HFIR) at Oak Ridge National Laboratory. The test specimen that are fabricated include 4 and 8 weight % Gd 2 O 3 doped UO 2 to investigate the influence of the BAs on the thermophysical properties and microstructure of the fresh and irradiated fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Pore-Filled PEMs from Poly(Phenylene Sulfonic Acid)s and Electrospun Poly(Phenylene Sulfone) Fiber Mats

A novel, composite, non-PFSA-based fuel cell membrane has been fabricated using a pore filling technique. The membrane consists of a mechanically stabilizing skeleton from an electrospun poly(phenylene sulfone) (PPSU) fiber mat and a thermally crosslinkable poly(phenylene sulfonic acid) (cPPSA) proton conducting ionomer that fills the interfiber voids. cPPSA copolymer was synthesized using Ullmann coupling copolymerization of 4,4-dibromobiphenyl 3,3-disulfonic acid with 1,4-dibromobenzene-2,5-disulfonic, followed by grafting a certain fraction of backbone sulfonic acid groups with biphenyl linker. The PPSU fiber mat was electrospun from NMP/acetone solution. Pore-filling was carried out by pouring a solution of cPPSA in methanol over the mat, followed by heating at 70°C to evaporate solvent. The cPPSA was crosslinked by an additional heating step, in a vacuum oven at 210°C for 5 hours. Here the resultant membrane had excellent proton conductivity, 5 times greater than that of Nafion® 211 in the 40-90% RH range at 80°C.

25 ENERGY STORAGE↗
DOE OSTI2024

Unraveling the conundrum of electronic leakage in protonic ceramic cells: Operation-specific insights and rational design strategies

Electronic conduction through proton-conducting electrolytes significantly impairs the efficiency of protonic ceramic cells (PCCs). Here, in this study, we explore the electron and ion mixed transport properties of four common protonic ceramics, BaZr 0.8 Y 0.2 O 3-δ (BZY82), BaZr 0.7 Ce 0.2 Y 0.1 O 3-δ (BZCY721), BaZr 04 Ce 0.4 Y 0.1 Yb 0.1 O 3-δ (BZCYYb4411), and BaZr 0.1 Ce 0.7 Y 0.1 Yb 0.1 O 3-δ (BZCYYb1711). It marks the first instance of investigating these properties under operation-specific scenarios: fuel side of electrolysis cell, air side of electrolysis cell, fuel side of fuel cell, and air side of the fuel cell. BZCYYb1711 exhibits the highest ionic conductivity, but two to three times higher electronic leakage when exposed to oxygen-containing environments than the others. BZY82 exhibits approximately two times higher electronic leakage in a hydrogen-containing environment. BZCY721 demonstrates excellent ion transport numbers (~0.95) across these four operating conditions. BZCYYb4411 behaves quite similarly to BZCY721. The most challenging operating environment for all candidates is the air side of fuel cell mode. This mode leads to a high initial electronic leakage, followed by a significant increase with polarization. The probable cause for this behavior is a H 2 -free, polarization-induced reduction that leads to the formation of V$^•_O$. The electron small polaron associated with V$^•_O$ is released by the electrical field due to the Poole-Frenkel effect. ZnO and NiO sintering aids are found to be detrimental to the ionic conductivity of the electrolytes. In particular, NiO substantially lowers the ion transport number. The correlation of the operation-specific electronic leakage to full cells is discussed. It is suggested that a rational PCC design should synergistically couple BZCYYb1711 at fuel side with BZCY4411 at air side to deliver a well-balanced performance and faradaic efficiency simultaneously, and the high temperature sintering process with a NiO fuel electrode should be shortened or replaced by ultra-fast sintering techniques or using a fuel electrode scaffold-infiltration fabrication strategy.

42 ENGINEERING↗