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NASA NTRS1976

Ternary boride product and process

A hard, tough, strong ceramic body is formed by hot pressing a mixture of a powdered metal and a powdered metal diboride. The metal employed is zirconium, titanium or hafnium and the diboride is the diboride of a different member of the same group of zirconium, titanium or hafnium to form a ternary composition. During hot pressing at temperatures above about 2,000.degree.F., a substantial proportion of acicular ternary monoboride is formed.

Clougherty, Edward V.↗

NASA NTRS2007

Ceramic Paste for Patching High-Temperature Insulation

A ceramic paste that can be applied relatively easily, either by itself or in combination with one or more layer(s) of high-temperature ceramic fabrics, such as silicon carbide or zirconia, has been invented as a means of patching cracks or holes in the reinforced carbon-carbon forward surfaces of a space shuttle in orbit before returning to Earth. The paste or the paste/fabric combination could also be used to repair rocket-motor combustion chambers, and could be used on Earth to patch similar high-temperature structures. The specified chemical composition of the paste admits of a number of variations, and the exact proportions of its constituents are proprietary. In general, the paste consists of (1) silicon carbide, possibly with addition of (2) hafnium carbide, zirconium carbide, zirconium boride, silicon tetraboride, silicon hexaboride, or other metal carbides or oxides blended with (3) a silazane-based polymer. Because the paste is viscous and sticky at normal terrestrial and outer-space ambient temperatures, high-temperature ceramic fabrics such as silicon carbide or zirconia fabric impregnated with the paste (or the paste alone) sticks to the damaged surface to which it is applied. Once the patch has been applied, it is smoothed to minimize edge steps as required [forward-facing edge steps must be < or equal to 0.030 in. (< or equal to 0.76 mm) in the original intended space-shuttle application]. The patch is then heated to a curing temperature thereby converting it from a flexible material to a hard, tough material. The curing temperature is 375 to 450 F (approx.190 to 230 C). In torch tests and arc-jet tests, the cured paste was found to be capable of withstanding a temperature of 3,500 F (approx. 1,900 C) for 15 minutes. As such, the material appears to satisfy the requirement, in the original space-shuttle application, to withstand re-entry temperatures of approx.3,000 F (approx. 1,600 C).

Adam, Steven J.↗

NASA NTRS2007

Ultrahigh-Temperature Ceramics

Ultrahigh temperature ceramics (UHTCs) are a class of materials that include the diborides of metals such as hafnium and zirconium. The materials are of interest to NASA for their potential utility as sharp leading edges for hypersonic vehicles. Such an application requires that the materials be capable of operating at temperatures, often in excess of 2,000 C. UHTCs are highly refractory and have high thermal conductivity, an advantage for this application. UHTCs are potentially applicable for other high-temperature processing applications, such as crucibles for molten-metal processing and high-temperature electrodes. UHTCs were first studied in the 1960 s by the U.S. Air Force. NASA s Ames Research Center concentrated on developing materials in the HfB2/SiC family for a leading-edge application. The work focused on developing a process to make uniform monolithic (2-phase) materials, and on the testing and design of these materials. Figure 1 shows arc-jet models made from UHTC materials fabricated at Ames. Figure 2 shows a cone being tested in the arc-jet. Other variations of these materials being investigated elsewhere include zirconium based materials and fiber-reinforced composites. Current UHTC work at Ames covers four broad topics: monoliths, coatings, composites, and processing. The goals include improving the fracture toughness, thermal conductivity and oxidation resistance of monolithic UHTCs and developing oxidation-resistant UHTC coatings for thermal-protection-system substrates through novel coating methods. As part of this effort, researchers are exploring compositions and processing changes that have yielded improvements in properties. Computational materials science and nanotechnology are being explored as approaches to reduce materials development time and improve and tailor properties.

Johnson, Sylvia M.↗

NASA NTRS2011

Lightweight, Ultra-High-Temperature, CMC-Lined Carbon/Carbon Structures

Carbon/carbon (C/C) is an established engineering material used extensively in aerospace. The beneficial properties of C/C include high strength, low density, and toughness. Its shortcoming is its limited usability at temperatures higher than the oxidation temperature of carbon . approximately 400 C. Ceramic matrix composites (CMCs) are used instead, but carry a weight penalty. Combining a thin laminate of CMC to a bulk structure of C/C retains all of the benefits of C/C with the high temperature oxidizing environment usability of CMCs. Ultramet demonstrated the feasibility of combining the light weight of C/C composites with the oxidation resistance of zirconium carbide (ZrC) and zirconium- silicon carbide (Zr-Si-C) CMCs in a unique system composed of a C/C primary structure with an integral CMC liner with temperature capability up to 4,200 F (.2,315 C). The system effectively bridged the gap in weight and performance between coated C/C and bulk CMCs. Fabrication was demonstrated through an innovative variant of Ultramet fs rapid, pressureless melt infiltration processing technology. The fully developed material system has strength that is comparable with that of C/C, lower density than Cf/SiC, and ultra-high-temperature oxidation stability. Application of the reinforced ceramic casing to a predominantly C/C structure creates a highly innovative material with the potential to achieve the long-sought goal of long-term, cyclic high-temperature use of C/C in an oxidizing environment. The C/C substructure provided most of the mechanical integrity, and the CMC strengths achieved appeared to be sufficient to allow the CMC to perform its primary function of protecting the C/C. Nozzle extension components were fabricated and successfully hot-fire tested. Test results showed good thermochemical and thermomechanical stability of the CMC, as well as excellent interfacial bonding between the CMC liner and the underlying C/C structure. In particular, hafnium-containing CMCs on C/C were shown to perform well at temperatures exceeding 3,500 F (.1,925 C). The melt-infiltrated CMC-lined C/C composites offered a lower density than Cf/SiC. The melt-infiltrated composites offer greater use temperature than Cf/SiC because of the more refractory ceramic matrices and the C/C substructure provides greater high-temperature strength. The progress made in this work will allow multiple high-temperature components used in oxidizing environments to take advantage of the low density and high strength of C/C combined with the high-temperature oxidation resistance of melt-infiltrated CMCs.

Wright, Matthew J.↗

NASA NTRS2012

Fischer-Tropsch Cobalt Catalyst Improvements with the Presence of TiO2, La2O3, and ZrO2 on an Alumina Support

The objective of this study was to evaluate the effect of titanium oxide, lanthanum oxide, and zirconium oxide on alumina supported cobalt catalysts. The hypothesis was that the presence of lanthanum oxide, titanium oxide, and zirconium oxide would reduce the interaction between cobalt and the alumina support. This was of interest because an optimized weakened interaction could lead to the most advantageous cobalt dispersion, particle size, and reducibility. The presence of these oxides on the support were investigated using a wide range of characterization techniques such as SEM, nitrogen adsorption, x-ray diffraction (XRD), temperature programmed reduction (TPR), temperature programmed reduction after reduction (TPR-AR), and hydrogen chemisorptions/pulse reoxidation. Results indicated that both La2O3 and TiO2 doped supports facilitated the reduction of cobalt oxide species in reference to pure alumina supported cobalt catalysts, however further investigation is needed to determine the effect of ZrO2 on the reduction profile. Results showed an increased corrected cluster size for all three doped supported catalysts in comparison to their reference catalysts. The increase in reduction and an increase in the cluster size led to the conclusion that the support-metal interaction weakened by the addition of TiO2 and La2O3. It is also likely that the interaction decreased upon presence of ZrO2 on the alumina, but further research is necessary. Preliminary results have indicated that the alumina-supported catalysts with titanium oxide and lanthanum oxide present are of interest because of the weakened cobalt support interaction. These catalysts showed an increased extent of reduction, therefore more metallic cobalt is present on the support. However, whether or not there is more cobalt available to participate in the Fischer-Tropsch synthesis reaction (cobalt surface atoms) depends also on the cluster size. On one hand, increasing cluster size alone tends to decrease the active site density; on the other hand, by increasing the size of the cobalt clusters, there is less likelihood of forming oxidized cobalt complexes (cobalt aluminate) during Fischer-Tropsch synthesis. Thus, from the standpoint of stability, improving the extent of reduction while increasing the particle size slightly may be beneficial for maintaining the sites, even if there is a slight decrease in overall initial active site density.

Klettlinger, Jennifer Lindsey Suder↗

NASA NTRS2016

Rapid Quench in an Electrostatic Levitator

The Electrostatic Levitation (ESL) Laboratory at the NASA Marshall Space Flight Center (MSFC) is a unique facility for investigators studying high-temperature materials. The ESL laboratory’s main chamber has been upgraded with the addition of a rapid quench system. This system allows samples to be dropped into a quench vessel that can be filled with a low melting point material, such as a gallium or indium alloy, as a quench medium. Thereby allowing rapid quenching of undercooled liquid metals. Up to eight quench vessels can be loaded into a wheel inside the chamber that is indexed with control software. The system has been tested successfully with samples of zirconium, iron-cobalt alloys, iron-chromium-nickel, titanium-zirconium-nickel alloys, and a silicon-cobalt alloy. This new rapid quench system will allow materials science studies of undercooled materials and new materials development. The system is described and some initial results are presented.

SanSoucie, Michael P.↗

NASA NTRS2016

Rapid Quench in an Electrostatic Levitator

The Electrostatic Levitation (ESL) Laboratory at the NASA Marshall Space Flight Center (MSFC) is a unique facility for investigators studying high-temperature materials. The ESL laboratory's main chamber has been upgraded with the addition of a rapid quench system. This system allows samples to be dropped into a quench vessel that can be filled with a low melting point material, such as a gallium or indium alloy, as a quench medium. Thereby allowing rapid quenching of undercooled liquid metals. Up to eight quench vessels can be loaded into a wheel inside the chamber that is indexed with control software. The system has been tested successfully with samples of zirconium, iron-cobalt alloys, titanium-zirconium-nickel alloys, and a silicon-cobalt alloy. This new rapid quench system will allow materials science studies of undercooled materials and new materials development. In this presentation, the system is described and some initial results are presented.

SanSoucie, Michael P.↗

NASA NTRS2016

Submersion Quenching of Undercooled Liquid Metals in an Electrostatic Levitator

The NASA Marshall Space Flight Center (MSFC) electrostatic levitation (ESL) laboratory has a long history of providing materials research and thermophysical property data. The laboratory has recently added a new capability, a rapid quench system. This system allows samples to be dropped into a quench vessel that can be filled with a low melting point material, such as a gallium or indium alloy. Thereby allowing rapid quenching of undercooled liquid metals and alloys. This is the first submersion quench system inside an electrostatic levitator. The system has been tested successfully with samples of zirconium, iron-cobalt alloys, titanium-zirconium-nickel alloys, and silicon-cobalt alloys. This rapid quench system will allow materials science studies of undercooled materials and new materials development, including studies of metastable phases and transient microstructures. In this presentation, the system is described and some initial results are presented.

SanSoucie, Michael P.↗

DOE OSTI2020

Titanium-group nano-whiskers and method of production

Disclosed herein are structures comprising a titanium, zirconium, or hafnium powder particle with titanium carbide, zirconium carbide, or hafnium carbide (respectively) nano-whiskers grown directly from and anchored to the powder particle. Also disclosed are methods for fabrication of such structures, involving heating the powder particles and exposing the particles to an organic gas.

Seals, Roland D.↗

DOE OSTI2022

PLN-6268 Yttrium Hydride Post Irradiation Examination Plan

The overarching goal of the Department of Energy Office of Nuclear Energy (DOE-NE) microreactor program is to develop technologies for the deployment of civilian microreactors by stakeholders1. Microreactors are expressed as advanced transportable nuclear reactors operating at low power (<20MWth) but high temperatures (>600°C), as well as plug-and-play and inherently safe designs. One prerequisite of a microreactor is the compactness, so that a truck can transport the reactor under safe conditions with the current road infrastructure1,2. The compactness of these reactors likely can be attainable by use of solid components for the essentials of the nuclear core, such as fuel, core heat removal components, reflectors, and moderators. Among these essentials, where fuel enrichment must remain < 20% to meet High Assay Low Enrichment Uranium criteria (HALEU), the largest contribution to the compactness is offered by use of solid moderators which benefit from light atomic weight elements, such as hydrogen, carbon, and beryllium2. Among these, hydrogen-bearing materials, such as metal hydrides, are superior to other options from the lowest critical mass standpoint. Noting that, factors other than critical mass should be considered for a specific reactor design. Yttrium- or zirconium-based metal hydrides have been down-selected due to their neutronic performance. In addition to the neutronic perspective, maintaining hydrogen within the metal hydride is important at the high operating temperatures proposed by advanced reactors. Yttrium hydride (YHx) is, therefore, a proposed moderator material that offers better hydrogen retention at higher operating temperatures than zirconium hydrides due to higher retention and thermal stability of hydrogen in the metal3. The irradiated materials properties of metal hydrides, in this case YHx, must be assessed for the qualification of these moderators. Material testing and inspection processes must illustrate that the effects of dimensional and property changes on thermophysical and mechanical properties do not cause any significant changes on the microreactor safety, and the moderating power is maintained within design limits. Thus, the effect of irradiation on the thermophysical and mechanical properties must be determined. This post-irradiation examination (PIE) plan specifically aims to determine these properties for YHx following Advanced Test Reactor (ATR) irradiation.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2021

Transition metal-based materials for use in high temperature and corrosive environments

A material (e.g., an alloy) comprises molybdenum, rhenium, and at least one element selected from the group consisting of tellurium, iodine, selenium, chromium, nickel, copper, titanium, zirconium, tungsten, vanadium, and niobium. Methods of forming the material (e.g., the alloy) comprise mixing molybdenum powder, rhenium powder, and a powder comprising at least one element selected from the group consisting of tellurium, iodine, selenium, chromium, nickel, copper, titanium, zirconium, tungsten, vanadium, and niobium. The mixed powders may be coalesced to form the material (e.g., the alloy).

Tripathy, Prabhat K.↗

DOE OSTI2022

Control rods for light water reactors

A control rod for a nuclear fuel assembly is described herein that includes a neutron absorbing material having a melting point greater than 1500° C. that does not form a eutectic with a melting point less than 1500° C., and may further include a cladding material having a melting point greater than 1500° C. The cladding material is selected from the group consisting of silicon carbide, zirconium, a zirconium alloy, tungsten, and molybdenum. The absorbing material is selected from the group consisting of Gd2O3, Ir, B4C, Re, and Hf. The metal cladding or the absorbing material may be coated with an anti-oxidation coating of Cr with or without a Nb intermediate layer.

Lahoda, Edward J.↗

DOE OSTI2022

A review of cladding failure thresholds in RIA conditions based on transient reactor test data and the need for continued testing

Transient reactor experiments on light water reactor (LWR) fuel pins had been conducted since the beginning of the nuclear era to help determine core coolability and cladding failure thresholds. During one such test in November of 1993 at the CABRI transient test reactor on a test involving a high burnup fuel rod with a corroded Zircaloy-4 cladding it was first observed that cladding failures could occur prior to a departure from nucleate boiling (pre-DNB) at lower-than-expected peak radial average enthalpies. Thirteen additional tests would be performed in the CABRI reactor over the next decade on fuel rods with higher burnups [1]. Additionally, a larger testing program at the NSRR reactor in Japan with high burnup fuel would uncover a similar trend of pre-DNB ruptures in high burnup test rods at lower-than-expected peak enthalpies [2]. Numerous out of pile testing programs involving a variety of innovative mechanical testing techniques have been employed in an attempt to better quantify the failure thresholds of corroded zirconium alloy cladding in these rapid heating and loading conditions [3][4][5]. While previously interim guidance had been issued, in June of 2020 the NRC officially published updated regulatory guidance to account of these pre-DNB failures in Regulatory Guide 1.236 [6]. This paper will present an independent review of the publicly available transient reactor test database on higher burnup LWR pins conducted at the CABRI and NSRR reactors as well as review of a selection of published out of pile mechanical testing methods. The purpose of the review is to determine how well the new regulatory limits are supported by experimental data. The review will identify if additional transient reactor tests could provide additional support for the NRC guidance or identify the need for revisions. The evaluation will consider how far the existing database can be extrapolated when considering low hydrogen zirconium alloy claddings (with and without protective coatings) containing very high burnup (> 70 MWd/kgU) UO2 fuel pellets. Finally, the authors will suggest how out of pile mechanical tests can be used in conjunction with a limited number of transient reactor tests to develop cladding specific failure thresholds in RIA type transients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Titanium-group nano-whiskers and method of production

Menchhofer, Paul A.↗

DOE OSTI2024

UZrCN Formation via Arc Melting – A Novel Synthesis Study

The next generation of nuclear reactors for both power production and space nuclear propulsion require fuel that is more durable, thermally stable, and more thermally conductive to support rapid heat transfer. High temperature gas reactors (HTGR), advanced gas reactors (AGR), and space-based nuclear thermal propulsion (NTP) are advanced reactor concepts that require a fuel type that can withstand high temperatures (1000-2900K) and flow of corrosive gas coolants such as helium, hydrogen, and carbon dioxide. One fuel with the potential to meet these demanding requirements is uranium-zirconium-carbonitride (UZrCN). UZrCN has many favorable fuel qualities compared to other eligible fuel forms such as uranium dioxide (UO2) and uranium mononitride (UN) that could support the aforementioned reactor concepts. UZrCN has an exceptionally high operating temperature and thermal conductivity which are highly desirable to improve reactor economics and safety. It far exceeds the properties of UO2 which is the most common fuel form in the United States. UZrCN also surpasses UN in terms of thermal conductivity and operating temperature by eliminating the dissociation problem UN has at 1700K. UZrCN could improve gas reactor performance and enable NTP technologies; however, it is an under-researched fuel that lacks rigorous scientific study. In recent efforts by the Idaho National Laboratory, a variety of novel methods to produce this fuel composition have been explored. One such method is via arc melting of uranium, zirconium, and carbon under a nitrogen atmosphere. Alloy fabrication using arc melting has been utilized for close to 150 years now and is well-understood as a method for rapid alloy prototyping. This process will be used to perform in-situ nitriding to form UZrCN.

36 MATERIALS SCIENCE↗

DOE OSTI2025

DISSOLUTION OF SURROGATE U-Zr FUEL USING ALNIFLEX CONDITIONS

Non-aluminum clad spent nuclear fuel (NASNF) stored in L Basin at the Savannah River Site (SRS) is widely varied in fuel composition, design, packaging, and physical condition. The complexity of the NASNF inventory presents significant challenges, and technology development is necessary for successful disposition. One such fuel in the inventory is metallic uranium-zirconium (U-Zr) alloy fuel, the focus of this study. Electrolytic or nitric acid only dissolution of metallic U-Zr alloy can form insoluble zirconium oxide, which results in up to 52% loss of U to insoluble solids, and can be subject to potentially uncontrolled oxidation reactions [1, 2]. The AlNiflex process was determined to be a viable dissolution flowsheet for the U-Zr alloy fuel. Under a narrow set of solution concentrations, a combination of hydrofluoric acid (HF), nitric acid (HNO3), aluminum nitrate (Al(NO3)3), and hexavalent chromium can safely dissolve U-Zr intermetallic alloys, keep Zr soluble, and not significantly corrode stainless steel (SS) vessels [3, 4.

Gogolski, Jarrod M. [Savannah River National Labor↗

DOE OSTI2025

Zr Nuclear Data Campaign: Measurement of (n,γ) cross section of 90,91 Zr [Abstract]

The isotopes of Zr with mass number A = [90; 91; 92; 94] make up more than 97% of naturally occurring Zr and are important to many nuclear applications due to their desirable mechanical and nuclear properties. For example, the zirconium alloys are used as cladding material in nuclear reactors. In addition to the mechanical properties, Zr-90 has a closed neutron shell, hence the capture cross sections of the zirconium isotopes are relatively small, such that they improve the neutron economy in reactors by preferentially scattering neutrons rather than absorbing them. This same quality also presents a challenge to measuring the capture cross section, σ γ (E), of Zr isotopes. For incident neutron energies < 0.1 MeV, the ENDF library reports a relative uncertainty of approximately 20% and uncertainty greater than 20% above this energy. This motivates the Nuclear Criticality Safety Program (NCSP) to perform new neutron induced cross section measurements to accurately evaluate these four isotopes of Zr.

ENDF↗

DOE OSTI2022

Leveraging computational thermodynamics to guide SiC-ZrC chemical vapor deposition process development

Here using the CALPHAD approach to understand zirconium carbide deposition, a series of phase equilibria were calculated from a custom thermodynamic database based on a literature source, and the equilibria were used to explore the potential chemical vapor deposition (CVD) processing space in the ZrCl 4 -CH 3 SiCl 3 -CH 4 -H 2 system as a function of pressure, temperature, and gas composition. Several gas ratios were considered. At a given ZrCl 4 :CH 3 SiCl 3 ratio within the range studied, the most important factor was found to be the ratios of CH 4 :ZrCl 4 , wherein the nature of the composition – carbide vs. silicide – could be controlled. A pure binary composition of ZrC and SiC is expected to form by increasing the initial amount of methane and decreasing the amount of hydrogen from values predicted purely based on thermodynamic equilibrium. Rietveld analysis of the x-ray diffractograms from corresponding experimental depositions confirmed that increasing the CH 4 :ZrCl 4 ratio increased the fraction of carbon-containing species (SiC, ZrC) and decreased the fraction of non-carbides (ZrSi, ZrSi 2 , etc.), as predicted from the CALPHAD results.

36 MATERIALS SCIENCE↗