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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Device for steam cladding oxidation testing at TREAT

To compare the chemical degradation of conventional zirconium alloy (Zry) cladding to advance silicon carbide (SiC) cladding in a post loss of coolant accident (LOCA) environment, new nuclear testing capabilities are necessary. The Transient Reactor Test (TREAT) Facility at Idaho National Laboratory (INL) has matured its transient fuel testing capabilities since its 2017 restart. The most recent experiment architecture is the Transient Water Irradiation System in TREAT (TWIST), which is designed to support qualification of accident tolerant fuels in light water reactors. INL has designed and analyzed a natural circulation steam flow modification for TWIST to produce prototypic conditions of cladding oxidation. The in-situ device will be electrically heated to drive natural circulation. Moreover, the SiC cladding requires heating above 1700 °C to observe failure, thus internal prototypic nuclear heating with radiation effects will be used. Thermal hydraulic analysis with RELAP5-3D (Reactor Excursion and Leak Analysis Program) estimated steam fluxes greater than 50 mg cm −2 s −1 can be achieved. These fluxes are adequate to test Zry cladding according to draft regulatory guides and to test SiC cladding according to past experiments.

Oxidation↗
DOE OSTI2024

UHT-CAMANCHE: Ultra-High Temperature Ceramic Additively Manufactured Compact Heat Exchangers

The conceptual basis for this project is the convergence of advanced ultra-high temperature ceramic materials and additive manufacturing technologies to produce compact ceramic heat exchangers with complex internal flow path geometries. Task areas were broadly divided into materials and manufacturing development, heat exchanger design, component testing, and techno-economic analysis. Technical challenges included the design and commissioning of new test facilities, improving feature resolution and deposition rate of ceramic additive manufacturing techniques, establishing process-structure-property relationships in additively manufactured ultra-high temperature ceramics, and assessing high temperature materials compatibility in CO 2 environments. The primary candidate material evaluated in this work is a composite comprising zirconium diboride (ZrB2) with 30 vol. % silicon carbide (SiC) which was selected based on its desirable combination of high temperature mechanical properties, high thermal and electrical conductivities, and oxidation resistance. High solids loaded ZrB2-SiC pastes suitable for extrusion-based additive manufacturing were developed for the first time as part of this work. Materials compatibility studies indicate this material oxidizes in CO 2 to form a protective borosilicate scale which transforms to pure silica above 1000°C. Parts made by additive manufacturing displayed enlarged grain sizes produced by pressureless sintering as compared to hot-press sintering. Increases in microstructural coarseness have outsized effect on oxidation performance up to 1400°C due to incomplete oxidation of coarse large diameter SiC particles resulting in lower amounts of silica that apparently inhibit protective scale formation. Additive manufacturing as a forming technique did not appear to significantly affect thermal conductivity, hardness, or elastic modulus, though flexural strength was reduced by half or more as compared to traditionally hot-pressed materials. This effect was attributed to the presence of strength-limiting flaws (ca. 40 microns in size) originating from extrudate inhomogeneities that could potentially be eliminated with further process improvements. Attempts to attain economies of scale for production of multi-kilowatt scale heat exchangers by ceramic additive manufacturing proved difficult. Lack of automation and a modest extrusion rate while retaining fine feature resolution made the overall process labor intensive and limited experimental throughput. A number of full-scale components were taken through post-process heat treatments including drying, binder burnout, and sintering; however, none survived without significant flaws or cracks. Therefore, no operational data from a newly installed heat exchanger test loop were able to be obtained during the performance period. Continued research and development is recommended to improve economic feasibility of ceramic additive manufacturing by standardizing the use of advanced sensors, artificial intelligence, and automation tools to reduce associated labor costs and accelerate production rates. The materials and manufacturing techniques demonstrated in this work are likely to find applications in defense and energy applications.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
NASA NTRS2001

Evaluation of Ultra-High Temperature Ceramics for Aeropropulsion Use

Among the ultra-high temperature ceramics (UHTC) are a group of materials consisting of zirconium diboride or hafnium diboride plus silicon carbide, and in some instances, carbon. These materials offer a good combination of properties that make them candidates for airframe leading edges on sharp-bodied reentry vehicles. These UHTC perform well in the environment for such applications, i.e., air at low pressure. The purpose of this study was to examine three of these materials under conditions more representative of a propulsion environment, i.e., higher oxygen partial pressure and total pressure. Results of strength and fracture toughness measurements, furnace oxidation and high velocity thermal shock exposures are presented for ZrB2 plus 20 volume % SiC, ZrB2 plus 14 volume % SiC plus 30 volume % C, and SCS-9a SiC fiber reinforced ZrB2 plus 20 volume % SiC. The poor oxidation resistance of UHTCs is the predominant factor limiting their applicability to propulsion applications.

Levine, Stanley R.↗
NASA NTRS2002

Evaluation of Ultra-High Temperature Ceramics for Propulsion Application

Ultra-high temperature ceramic (UHTC) materials consisting of zirconium diboride or hafnium diboride plus silicon carbide offer a combination of properties that make them candidates for airframe leading edges on sharp-bodied reentry vehicles. These UHTC perform well in the environment for such applications, i.e. low pressure air. The purpose of this study was to examine three of these materials under conditions more representative of a propulsion environment, i.e. higher oxygen and total pressure. Results of stren$h and fracture toughness measurements, furnace oxidation, and thermal shock exposures are presented. The poor oxidation resistance of UHTCs is the predominant factor limiting their applicability to propulsion.

Levine, Stanley R.↗
NASA NTRSDate not supplied

Unified Material-Environment Interaction Model for Binary UHTC Composites

A new model for the surface chemistry of zirconium and hafnium diborides containing silicon-carbide when exposed to high temperature conditions in air is formulated. Oxidation dynamics are simplified by considering the diffusion-limited equilibrium regime within the material at elevated temperatures. This model is first assessed against experimental mass-gain data taken from UHTC samples heated to high temperature in an oxygen environment. Then, further evaluations of the new model are conducted using hypersonic CFD simulations to analyze conditions experienced by HfB2-SiC in arc jet experiments conducted at NASA Ames Research Center. Measurements of stagnation point heat transfer and pressure are then used to calibrate the simulations. Results of coupled CFD-material response simulations are then compared to the surface temperatures measured during the Ames arc jet test. Results support the use of the proposed model while highlighting the need for improved gas-phase data on the additional constituents considered.

ultra high temperature ceramic↗
DOE OSTI2021

Dynamic Probabilistic Risk Assessment Based Response Surface Approach for FLEX and Accident Tolerant Fuels for Medium Break LOCA Spectrum

After the Fukushima Daiichi Accident, the safety features such as accident tolerant fuel (ATF) and diverse and flexible coping strategies (FLEX) for existing nuclear fleets are being investigated by the US Department of Energy under the Light Water Reactor Sustainability Program. This research is being conducted to quantify the risk-benefit of these safety features. Dynamic probabilistic risk assessment (DPRA)-based response-surface approach has been presented to quantify the FLEX and ATF benefits by estimating the risk associated with each option. ATFs with multilayered silicon carbide (SiC), iron-chromium-aluminum, and chromium-coated zirconium cladding were considered in this study. While these ATF candidates perform better than the current zirconium cladding (Zr), they may introduce additional failure modes in some operating conditions. The fuel failure analysis modules (FAMs) were developed to investigate ATF performance. The dynamic risk assessments were performed using RAVEN, a DPRA tool, coupled with RELAP5 and FAMs. A cumulative distribution function-based index provided a mean of comparing the benefits of safety enhancements. For medium break loss of coolant accidents, FLEX operational timing window for each fuel type was estimated. Among these ATF candidates, SiC-type ATF was the most beneficial candidate for an increased safety margin than Zr-based fuel and was found to complement FLEX strategies in terms of risk and coping time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Phase Equilibria and Thermochemistry of Advanced Fuels: Modeling Burnup Behavior (Final Report)

Achieving the goal of developing advanced fuel concepts that meet the DOE objectives of being robust, demonstrating high performance, and are more tolerant of accident conditions than current fuel systems will require a thorough understanding of the thermophysical and thermochemical properties of the constituent materials. Non-oxide fuel systems are being explored under the Advanced Fuels Program that hold significant promise for improved performance and accident tolerance, including the uranium silicide-based system considered in the current work. Prospective cladding materials currently considered that contribute to improved accident tolerance include silicon carbide composites and ferritic alloys (Fe-Cr-Al base compositions). Thus, the effort developed thermochemical models and values, supported with targeted experiments, to evaluate the ferritic alloy and silicon carbide composite cladding systems in contrast to current zirconium alloy cladding. The developed detailed understanding will serve to aid in relatively early screening of candidate systems to avoid wasted effort, guide development of new fuel forms, and to provide a basis for predicting and modeling fuel performance. Major deliverables for the project included: Thermochemical assessment and models of phases in the U-Si and U-Si-N systems; thermochemical evaluation supported by experimental measurements of fuel-cladding interactions of silicide fuel with baseline zirconium, ceramic composite, and ferritic alloy cladding; thermochemical assessment and models of phases supported by experimental measurements for silicide fuel with key fission products provided in a dataset and reported in refereed publications. Within the project a significantly refined U-Si phase diagram was developed and reported that now includes homogeneity ranges for key phases, such at the U 3 Si 2 proposed fuel phase, and settles issues with regard to uncertainty in the stability of some phases. Computational efforts together with key experiments has determined phase formation in interactions between U 3 Si 2 and Zircaloy-4 cladding material, a ferritic FeCrAlY alloy of interest as an advanced cladding material, and silicon carbide, also of interest as a fiber-reinforced composite cladding. As expected, very significant reactions occur between U3Si2 and Zircaloy-4, with much less interaction at higher temperatures for the ferritic alloy, and finally interactions with SiC only in the region of contact. A potentially major issue is the stability of U 3 Si 2 fuel that has undergone significant burnup. The result is the loss of the uranium metal, liberating silicon, and the formation of concomitant fission product elements that either dissolve in the U 3 Si 2 phase or form independent, and possibly silicide phases. A combination of experimental determinations of phase formation of U 3 Si 2 reacted with representative fission products yttrium, gadolinium, cerium, zirconium, and molybdenum and first principles calculations has helped understand the fuel chemistry. The behavior of the U 3 Si 2 phase and the partitioning of silicon to possible fission product phases with burnup was thus determined, with significant dissolution in U 3 Si 2 of cerium, gadolinium, zirconium, and plutonium predicted along with independent phase formation of a U-Mo-Si ternary phase, yttrium silicide, and elemental selenium. It can be concluded that at significant burnup there will be a very minor amount of the U 3 Si 2 fuel phase that will decompose to a lower silicide or a uranium alloy as silicon preferentially forms a secondary phase.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Other Power Reactor Fuels

While oxide fuels have received the bulk of historic interest as power reactor fuels, a range of other uranium compounds and deployment architectures have also been considered and demonstrated to varying degrees. Metal alloys, technical ceramics, and oxycarbide or oxynitride particle fuels are candidates for various power reactor applications. Although in general these families of nuclear fuels have received more limited study than oxides, the benefits of uranium density, favorable thermophysical properties, and other factors have resulted in interest from both the traditional nuclear power industry as well as more recent entrepreneurial efforts. The dominant fuel chemistry issues for these fuel forms differ substantially from those of oxide systems. Synthesis routes, factors that affect fresh fuel properties, and critical aspects of fuel performance are unique to these systems and often dominated by the desired chemical behavior of uranium in each system. This chapter will briefly introduce metallic uranium alloys, specifically uranium-zirconium, non-traditional ceramic fuels such as carbides and nitrides, and oxycarbide/oxynitride particle fuels most familiar as the basis for TRISO fuel designs. A survey of major fuel chemistry issues is provided for each of these families of fuels, with a focus on contemporary challenges and active research avenues.

Nelson, Andrew↗
NASA NTRS2003

Ultra-High-Temperature Ceramics Evaluated for Aeropropulsion Use

Ultra-high-temperature ceramics (UHTC) are a group of materials consisting of zirconium diboride (ZrB2) or hafnium diboride (HfB2) plus silicon carbide (SiC), and in some instances, carbon (C). They offer a combination of properties that make them candidates for airframe leading edges on sharp-bodied reentry vehicles. These UHTCs perform well in the environment for such applications (i.e., air at low pressures). The purpose of this study at the NASA Glenn Research Center was to examine three of these materials under conditions more representative of a propulsion environment: that is, higher oxygen partial pressure and total pressure. Relatively long, multiple-exposure cycles were emphasized. We completed an in-house study of ZrB2 plus 20 vol% SiC (abbreviated as ZS), ZrB2 plus 14 vol% SiC and 30 vol% C (ZSC), and SCS-9a SiC fiber-reinforced ZrB2 plus 20 vol% SiC (ZSS). HfB2-based compositions were not included in the study because of their high cost. The capability of UHTC for propulsion applications must be compared with that of mature, available, and commercially used ceramics such as silicon nitride (e.g., AS-800) to put things in proper perspective. In terms of mechanical properties, UHTCs fall short in terms of strength and fracture toughness. At about 1300 C, the creep resistance of ZS appears to be superior to the creep resistance reported for AS-800. However, the stress rupture life for Si3N4 under stress and temperature conditions similar to those used in this study is measured in hundreds of hours. Because of oxidation, ZS could not achieve such lives. In terms of oxidation resistance, acceptable amounts of material recession in 1 hour to thousands of hours, depending on the specific propulsion application, are on the order of 100 to 300 mm. This converts to an acceptable range of parabolic recession rate constants kp" of approximately less than or equal to 10(exp -1) to 10(exp -2) square millimeters per hour for a 1-hr application. For a 100-hr application, an acceptable range of kp" would be less than or equal to those values divided by 100. For the more oxidation resistant ZS material, measured parabolic recession rate constants were 4.7 x 10(exp -3) millimeters per hour at 1327 C, 7.8 x 10(exp -2) square millimeters per hour at 1627 C, and 1.3 millimeters per hour at 1927 C. Thus, recession rate constants for ZS are acceptable at 1327 C for a 1-hr application, but here silicon nitride is a superior material. At 1627 C, ZS oxidation is marginal for a 1-hr application, but dimensional growth would be an issue. In a 100-hr application, ZS cannot be considered at any temperature. Our cursory examination of thermal shock, both from a theoretical and experimental viewpoint, indicated that the ZS and ZSC UHTCs are inferior to AS-800 silicon nitride. On the basis of this limited study, UHTCs are not ready to be considered as aeropropulsion materials for any applications longer than a few minutes. Current materials suffer from aggressive oxidation and moisture attack (Quynhgiao Nguyen, NASA Glenn, and Raymond C. Robinson, QSS Group, Inc., Cleveland, OH, 2002, private communication), and they are susceptible to thermal shock. For long-term propulsion applications, major improvements in environmental durability are needed. Work is in progress to improve the oxidation resistance of UHTC materials.

Levine, Stanley R.↗
DOE OSTI2025

Grain boundary zirconia-modified garnet solid-state electrolyte

Here, we report a method for promoting electrochemical stability in garnet Li 6.4 La 3 Zr 1.4 Ta 0.6 O 12 solid-state electrolyte based on a composite two-phase oxide–oxide microstructure. Grain boundary precipitation of the controlled distribution of amorphous zirconium oxide microparticles is achieved through the addition of reactive tantalum carbide. During ambient-atmosphere sintering, the carbide decomposes through an in situ reaction, the ‘extra’ Ta substituting for Zr within the Li 6.4 La 3 Zr 1.4 Ta 0.6 O 12 lattice. Density functional theory (DFT) calculations identify a thermodynamically favourable reaction path and show how substituting Ta 5+ at Zr 4+ sites affects the crystal structure as well as bulk ionic and electronic conductivities. Quantitative stereology highlights that zirconia also acts as a sintering aid, reducing compact porosity. Cryogenic focused-ion-beam scanning electron microscopy and fractography analysis of cycled solid-state electrolytes illustrates that near-universally observed intergranular Li-metal dendrite propagation is suppressed by the two-phase microstructure, favouring transgranular dendrites instead. Importantly, DFT demonstrates that compared with the Li 6.4 La 3 Zr 1.4 Ta 0.6 O 12 surface, the zirconium oxide surface per se is less electronically conductive and does not trap excess electrons to reduce Li ions. This is a key reason for the substantial improvement in the electrochemical properties over the single-phase baseline.

36 MATERIALS SCIENCE↗
NASA NTRS1998

Experiments on Nucleation in Different Flow Regimes

The vast majority of metallic engineering materials are solidified from the liquid phase. Understanding the solidification process is essential to control microstructure, which in turn, determines the properties of materials. The genesis of solidification is nucleation, where the first stable solid forms from the liquid phase. Nucleation kinetics determine the degree of undercooling and phase selection. As such, it is important to understand nucleation phenomena in order to control solidification or glass formation in metals and alloys. Early experiments in nucleation kinetics were accomplished by droplet dispersion methods. Dilitometry was used by Turnbull and others, and more recently differential thermal analysis and differential scanning calorimetry have been used for kinetic studies. These techniques have enjoyed success; however, there are difficulties with these experiments. Since materials are dispersed in a medium, the character of the emulsion/metal interface affects the nucleation behavior. Statistics are derived from the large number of particles observed in a single experiment, but dispersions have a finite size distribution which adds to the uncertainty of the kinetic determinations. Even though temperature can be controlled quite well before the onset of nucleation, the release of the latent heat of fusion during nucleation of particles complicates the assumption of isothermality during these experiments. Containerless processing has enabled another approach to the study of nucleation kinetics. With levitation techniques it is possible to undercool one sample to nucleation repeatedly in a controlled manner, such that the statistics of the nucleation process can be derived from multiple experiments on a single sample. The authors have fully developed the analysis of nucleation experiments on single samples following the suggestions of Skripov. The advantage of these experiments is that the samples are directly observable. The nucleation temperature can be measured by noncontact optical pyrometry, the mass of the sample is known, and post-processing analysis can be conducted on the sample. The disadvantages are that temperature measurement must have exceptionally high precision, and it is not possible to isolate specific heterogeneous sites as in droplet dispersions. Levitation processing of refractory materials in ultra high vacuum provides an avenue to conduct these kinetic studies on single samples. Two experimental methods have been identified where ultra high vacuum experiments are possible; electrostatic levitation in ground-based experiments and electromagnetic processing in low earth orbit on TEMPUS. Such experiments, reported here, were conducted on zirconium. Liquid zirconium is an excellent solvent and has a high solubility for contaminants contained in the bulk material as well as those contaminants found in the vacuum environment. Oxides, nitrides, and carbides do not exist in the melt, and do not form on the surface of molten zirconium, for the materials and vacuum levels used in this study. Ground-based experiments with electrostatic levitation have shown that the statistical nucleation kinetic experiments are viable and yield results which are consistent with classical nucleation theory. The advantage of low earth orbit experiments is the ability to vary the flow conditions in the liquid prior to nucleation. The put-pose of nucleation experiments in TEMPUS was to examine.

Bayuzick, R. J.↗
DOE OSTI2025

Reactive Carbide‐Based Synthesis and Microstructure of NASICON Sodium Metal All Solid‐State Electrolyte

Reactive carbide precursor-based synthesis of NASICON-type NZSP (Na 1+x Zr 2 Si x P 3-x O 12 ) solid-state electrolyte (SSE) is demonstrated, in contrast to the established oxide-based approach. Exothermic decomposition of ZrC and SiC in air homogenizes microstructure, yielding 98% compact density after conventional sintering at 1200 °C. Quantitative stereology demonstrates that significant microstructural differences are present. Compacts of carbide-derived Carb-NZSP are 98% dense with a secondary zirconium oxide (ZrO 2 ) volume fraction of 0.2% ± 0.3%, versus 93% dense and 3% ± 1% for oxide-derived baseline. For Carb-NZSP, the secondary glassy phosphate phase is agglomerated, while for baseline, it is dispersed and percolated. Electrochemical testing combined with post-mortem analysis demonstrates how microstructural control of secondary phases is critical for dendrite suppression: Carb-NZSP critical current density (CCD) is 3.1 ± 0.8 mA cm −2 at 0.1 mAh cm −2 , versus 1.0 ± 0.7 mA cm −2 at 0.1 mAh cm −2 . Cryogenic focused ion beam (cryo-FIB) analysis demonstrates that in both materials, the porous 2D sheet-like sodium metal dendrites propagate around and subsume NZSP grains, likely following a path enriched with glassy phase and with porosity. Dendrites also flow around isolated zirconia particles. Phase field simulation reveals deflection of dendrites by mechanically tough zirconia, while brittle glassy phase accelerates dendrite growth, especially when finely distributed.

36 MATERIALS SCIENCE↗
DOE OSTI2021

High Level Gap Analysis for Accident Tolerant and Advanced Fuels for Storage and Transportation

This initial gap analysis considers proposed accident tolerant fuel (ATF) options currently being irradiated in commercial reactors, since these are most likely for future batch implementation. Also, advanced fuel (AF) options that may be likely for use in advanced reactors are considered. The cladding technologies considered were chromium-coated zirconium-based alloys, FeCrAl, and both monolithic and matrix composite Silicide carbide (SiC). The fuel technologies considered were chromium-doped uranium dioxide fuel, uranium alloys, uranium nitride, and uranium silicide. Numerous national labs, industry, and countries are performing significant testing and modeling on these proposed technologies to establish performance, but at this time none of the prototypes being irradiated have achieved end-of-life (EOL) burnup. There are some testing results after one burnup cycle to verify in-reactor performance, but little data beyond that. As the ATF prototypes acquire more burnup, data will be produced that is relevant to storage and transportation. The DOE:NE Spent Fuel and Waste Science and Technology (SWFST) Storage and Transportation (ST) Control Account will evaluate the performance data as it becomes available for application to the identified gaps for ST.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Alternate Methods for Cleaning Zirconium Plate

The United States High Performance Research Reactor Conversion Program identified the need to transition from highly enriched uranium fuel to lowly enriched uranium fuel for high power research and civil reactors. One of the fuel configurations selected for these reactors was a uranium (U) foil alloyed with 10% molybdenum (Mo). The U-10Mo fuel foil was then covered with a zirconium (Zr) interlayer and pressed into an aluminum cladding. Idaho National Laboratory prescribed a need to remove surface oxidation on the Zr prior to bonding it with U-10Mo fuel foil. To determine the best method of oxidation removal, a variety of polishing methods, including hand polishing with a diamond paste or Scotch-Brite™ pads, mechanically wet-polishing with abrasive belts, and chemical etching with hydrofluoric/nitric acid mixtures, were investigated on pre-polished Zircaloy 702 sheets purchased directly from the manufacturer. No change in surface roughness was detected using any of these methods relative to the as-received material. In fact, scanning electron microscopy detected grit from hand polishing with Amplex Grade 30 waste soluble diamond paste and Scotch-Brite™ pads. Visible staining was found, and fluorine detected via X-ray photoelectron spectroscopy on the flash etching samples. These contaminants were not seen on the as-received material. In addition to the need to remove the oxide layer on the Zr prior to co-rolling, there are concerns that the Neolube debonding agent on the co-rolling can may contaminate the Zr surface after co-rolling. The experimental results found that zirconium in contact with Neolube coated steel plates was easily cleaning with ethanol wiping after heat treated and no carbide inclusions remained on the surface. Overall, all cleaning methods tested did not significantly change the surface roughness of the plates or reduce the oxide layer present on the zirconium. Carbon was found on all the samples, before and after cleaning, except for the flash etch samples which detected fluorine instead. The insignificant changes seen in the cleaned plates relative to the as received plates question the need for cleaning prior to rolling. Further investigation in the requirements for cleaning at this step is recommended. Current methods for cleaning the surface after hot rolling with alcohol appear sufficient. After hot rolling, if Neolube does peel off the can and onto the zirconium coated foil surface, a dry abrasive or diamond paste polish is not recommended due to the chance of abrasive being embedded into the surface. However, the ability of an etchant to lift the lubricant off the zirconium surface indicates that a spot etch method could work to clean specific spots where Neolube is present, if the need arises.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Development of Refractory Alloys and Refractory Coatings for Advanced Nuclear Reactors

The next generation of nuclear reactors will benefit from materials that enable operation at higher temperatures (>500°C), higher irradiation doses (up to 200 displacements per atom (dpa)), and the use of more corrosive and reactive coolants. This work package represents the first experimental steps towards a longer-term effort to develop refractory materials for nuclear energy applications which will enable operation under these conditions. Specifically, this work package focuses on additive manufacturing of refractories as both a refractory liner coating deposited onto the interior surface of a metallic tubular backbone and as bulk refractory alloys. During fiscal year 23 (FY23), several refractory metal coating systems and bulk alloys were examined and selected using a decision criteria matrix. The refractory metal coating systems included molybdenum, tungsten, and zirconium as refractory coatings on backbones of either carbon-carbon (C/C) or silicon carbide-silicon carbide (SiC/SiC) ceramic matrix composites. The bulk refractory alloys included C-103, WTa, and WNiFe as bulk alloys. During FY24 additional bulk refractory alloys and metallic backbones were evaluated using the decision criteria matrix based on input from the AMMT leadership team. These included 316 SS and 316H SS for the metallic backbones and Mo-La, Ta, and Nb1Zr as bulk refractory alloys. The primary focus for the FY24 effort was placed on establishing the capabilities to deposit refractory coatings based on the results of the scoring in the decision criteria matrix and finalizing the additively manufactured TZM studies which were incorporated into the AMMT program from the microreactor program. Further efforts were dedicated to establishing the capabilities to additively manufacture down-selected bulk refractory alloys.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1971

NASA research on refractory compounds.

The behavior and properties of the refractory carbides, nitrides, and borides are being investigated by NASA as part of its research aimed at developing superior heat resistant materials for aerospace applications. Fundamental studies on the electronic and defect structures of the carbides indicate that there is promise for improving the strength and ductility of these materials. Studies of the zirconium-carbon-oxygen system show that zirconium oxycarbides of different compositions and lattice parameters can be formed between 1500 and 1900 C and are stable below 1500 C. More applied studies show that hot working generally improves the microstructure and therefore the strength of TiC and NbC. Sintering studies on UN indicate that very high densities can be achieved. Hot pressing of cermets of HfN and HfC produces good mechanical properties for high temperature bearing applications.

Gangler, J. J.↗
DOE OSTI2023

Shock-Wave Proprieties of High-Strength Ceramics

The shock and release response of a suite of high-strength ceramics has been investigated with controlled planar impact methods and velocity interferometry diagnostics. The study provides Hugoniot, yield and post-yield compressive strength, phase transformation and tensile spall material properties data for these ceramics.

36 MATERIALS SCIENCE↗
NASA NTRS1981

Adhesion and friction of transition metals in contact with nonmetallic hard materials

Sliding friction experiments were conducted with the metals yttrium, titanium, tantalum, zirconium, vanadium, neodymium, iron, cobalt, nickel, tungsten, platinum, rhenium, ruthenium, and rhodium in sliding contact with single crystal diamond, silicon carbide, pyrolytic boron nitride, and ferrite. Auger electron spectroscopy analysis was conducted with the metals and nonmetals to determine the surface chemistry and the degree of surface cleanliness. The results of the investigation indicate the adhesion and friction of the transition metals in contact with diamond, silicon carbide, boron nitride, and ferrite are related to the relative chemical activity of the metals. The more chemically active the metal, the higher the coefficient of friction and the greater amount of transfer to the nonmetals.

Miyoshi, K.↗