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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 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
NASA NTRS1989

The interface in tungsten fiber reinforced niobium metal-matrix composites

The creep resistance of tungsten fiber reinforced niobium metal-matrix composites was evaluated. The interface region between the fiber and matrix was characterized by microhardness and electron probe microanalysis measurements which indicated that its properties were between those of fiber and matrix. However, the measured properties of the composite exceeded those calculated by the rule of mixtures even when the interface zone was assumed to retain all the strength of the fiber. The composite structure appeared to enhance the strengths of both the fibers and the matrix above what they exhibited in stand-alone tests. The effect of fiber orientation and matrix alloy composition on the fiber/matrix interface were also evaluated. Small alloying additions of zirconium and tungsten to the niobium matrix affected the creep resistance of the composites only slightly. A decrease in the creep resistance of the composite with increasing zirconium content in the matrix was ascribed to an increase in the diffusion rate of the fiber/matrix interdiffusion reaction, and a slight increase in the creep resistance of the composite was observed with an addition of 9 w percent tungsten to the matrix. In addition, Kirkendall void formation was observed at the fiber/matrix interface; the void distribution differed depending on the fiber orientation relative to the stress axis.

Grobstein, Toni L.↗
DOE OSTI2021

Development of a Castable High Strength Secondary Aluminum Alloy from Recycled Wrought Aluminum Scrap

A new process for recycling scrap AA7075 aerospace alloys directly into a high strength castable secondary alloy is demonstrated. The process involves enhancing the castability of the utilized primary wrought aluminum scrap feedstock by scrap alloy chemistry optimization and zirconium additions in order to mitigate the inherent hot tearing tendency of 7075 alloys. By utilizing the new process, the hot tearing index of scrap AA7075 was reduced from 26 to 4. Defect-free castings were obtained in pilot-scale castings trials. The developed secondary 7075 alloy also satisfied preset target mechanical properties with strength levels exceeding 250 MPa and elongation levels higher than 3%. The validity of the new approach and its readiness for transition to commercial practice was demonstrated by casting cylinder heads directly from the modified 7075 scrap.

36 MATERIALS SCIENCE↗
NASA NTRS1989

TEM studies of oxidized NiAl and Ni3Al cross sections

Cross sections of oxide scale/(Ni-Al) intermetallics were prepared by a new method and studied using primarily TEM. The cross sections were prepared by encasing an oxidized metal specimen sandwich in a low-melting-temperature zinc alloy. Observations of oxidized zirconium-doped beta-NiAl cross sections revealed crystallographic voids beneath an adherent Al2O3 scale. The oxide-metal interface was incoherent, but a high dislocation density in the metal near the interface suggested that a large tensile stress was induced by the attached oxide scale. A duplex Al2O3-NiAl2O4 scale formed on zirconium-doped and zirconium/boron-doped gamma-prime-Ni3Al alloys. Additional results are presented involving oxidation mechanisms and oxide-metal interface structures.

Doychak, J.↗
DOE OSTI2024

Evaluation of Sb-Nd and Te-Nd phases within the U-Zr fuel matrix and their interactions with HT9 alloy

Antimony (Sb) and tellurium (Te) were investigated as potential additives for U-10Zr (wt.%) metallic fuel to limit the fuel-cladding chemical interaction (FCCI) with HT-9 alloy. Neodymium (Nd) was utilized to simulate the formation of lanthanide-based solid fission products which are known to play a detrimental role in FCCI. Fuel alloys of U-Zr-Sb-Nd and U-Zr-Te-Nd were evaluated in their annealed condition and compared against their as-cast conditions. Isothermal diffusion couple experiments were performed between U-Zr-Nd, U-Zr-Sb-Nd, and U-Zr-Te-Nd against the cladding alloy HT9 to evaluate the effectiveness of the additives to stabilize Nd within the fuel alloys, as well as investigate the interaction regions that form between the different fuel alloys and HT9. Further, SbNd and Sb 3 Nd 4 , and TeNd are found to be the primary neodymium-based phases formed in the U-Zr-Sb-Nd and U-Zr-Te-Nd alloys, respectively. The zirconium-based phase, Zr 2 Sb, is also found to form within the former alloy. All phases were found to remain stable through the diffusion experiments and exhibited no interaction with HT9 constituent elements. Preferential interaction between Nd with additivities Te and Sb compared to constituting elements in HT9 was further verified based on density functional theory (DFT) calculated enthalpy of mixing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1991

Investigation of welding and brazing of molybdenum and TZM alloy tubes

This effort involved investigating the welding and brazing techniques of molybdenum tubes to be used as cartridges in the crystal growth cartridge. Information is given in the form of charts and photomicrographs. It was found that the recrystallization temperature of molybdenum can be increased by alloying it with 0.5 percent titanium and 0.1 percent zirconium. Recrystallization temperatures for this alloy, known as TZM, become significant around 2500 F. A series of microhardness tests were run on samples of virgin and heat soaked TZM. The test results are given in tabular form. It was concluded that powder metallurgy TZM may be an acceptable cartridge material.

Lundblad, Wayne E.↗
NASA NTRS1991

The effect of copper, chromium, and zirconium on the microstructure and mechanical properties of Al-Zn-Mg-Cu alloys

The present study evaluates the effect of the systematic variation of copper, chromium, and zirconium contents on the microstructure and mechanical properties of a 7000-type aluminum alloy. Fracture toughness and tensile properties are evaluated for each alloy in both the peak aging, T8, and the overaging, T73, conditions. Results show that dimpled rupture essentially characterize the fracture process in these alloys. In the T8 condition, a significant loss of toughness is observed for alloys containing 2.5 pct Cu due to the increase in the quantity of Al-Cu-Mg-rich S-phase particles. An examination of T8 alloys at constant Cu levels shows that Zr-bearing alloys exhibit higher strength and toughness than the Cr-bearing alloys. In the T73 condition, Cr-bearing alloys are inherently tougher than Zr-bearing alloys. A void nucleation and growth mechanism accounts for the loss of toughness in these alloys with increasing copper content.

Wagner, John A.↗
NASA NTRS1985

Fatigue crack propagation of nickel-base superalloys at 650 deg C

The 650 C fatigue crack propagation behavior of two nickel-base superalloys, Rene 95 and Waspaloy, is studied with particular emphasis placed on understanding the roles of creep, environment, and two key grain boundary alloying additions, boron and zirconium. Comparison of air and vacuum data shows the air environment to be detrimental over a wide range of frequencies for both alloys. More in-depth analysis on Rene 95 shows at lower frequencies, such as 0.02 Hz, failure in air occurs by intergranular, environmentally-assisted creep crack growth, while at higher frequencies, up to 5.0 Hz, environmental interactions are still evident but creep effects are minimized. The effect of B and Zr in Waspaloy is found to be important where environmental and/or creep interactions are presented. In those instances, removal of B and Zr dramatically increases crack growth and it is therefore plausible that effective dilution of these elements may explain a previously observed trend in which crack growth rates increase with decreasing grain size.

Gayda, J.↗
NASA NTRS1986

Thermal aging effects in refractory metal alloys

The alloys of niobium and tantalum are attractive from a strength and compatibility viewpoint for high operating temperatures required in materials for fuel cladding, liquid metal transfer, and heat pipe applications in space power systems that will supply from 100 kWe to multi-megawatts for advanced space systems. To meet the system requirements, operating temperatures ranging from 1100 to 1600 K have been proposed. Expected lives of these space power systems are from 7 to 10 yr. A program is conducted at NASA Lewis to determine the effects of long-term, high-temperature exposure on the microstructural stability of several commercial tantalum and niobium alloys. Variables studied in the investigation include alloy composition, pre-age annealing temperature, aging time, temperature, and environment (lithium or vacuum), welding, and hydrogen doping. Alloys are investigated by means of cryogenic bend tests and tensile tests. Results show that the combination of tungsten and hafnium or zirconium found in commercial alloys such as T-111 and Cb-752 can lead to aging embrittlement and increased susceptibility to hydrogen embrittlement of ternary and more complex alloys. Modification of alloy composition helps to eliminate the embrittlement problem.

Stephens, J. R.↗
NASA NTRS1987

Thermal aging effects in refractory metal alloys

The alloys of niobium and tantalum are attractive from a strength and compatibility viewpoint for high operating temperatures required in materials for fuel cladding, liquid metal transfer, and heat pipe applications in space power systems that will supply from 100 kWe to multi-megawatts for advanced space systems. To meet the system requirements, operating temperatures ranging from 1100 to 1600 K have been proposed. Expected lives of these space power systems are from 7 to 10 yr. A program is conducted at NASA Lewis to determine the effects of long-term, high-temperature exposure on the microstructural stability of several commercial tantalum and niobium alloys. Variables studied in the investigation include alloy composition, pre-age annealing temperature, aging time, temperature, and environment (lithium or vacuum), welding, and hydrogen doping. Alloys are investigated by means of cryogenic bend tests and tensile tests. Results show that the combination of tungsten and hafnium or zirconium found in commercial alloys such as T-111 and Cb-752 can lead to aging embrittlement and increased susceptibility to hydrogen embrittlement of ternary and more complex alloys. Modification of alloy composition helps to eliminate the embrittlement problem.

Stephens, Joseph R.↗
NASA NTRS1988

Fatigue crack propagation of nickel-base superalloys at 650 deg C

The 650 C fatigue crack propagation behavior of two nickel-base superalloys, Rene 95 and Waspaloy, is studied with particular emphasis placed on understanding the roles of creep, environment, and two key grain boundary alloying additions, boron and zirconium. Comparison of air and vacuum data shows the air environment to be detrimental over a wide range of frequencies for both alloys. More in-depth analysis on Rene 95 shows at lower frequencies, such as 0.02 Hz, failure in air occurs by intergranular, environmentally-assisted creep crack growth, while at higher frequencies, up to 5.0 Hz, environmental interaction are still evident but creep effects are minimized. The effect of B and Zr in Waspaloy is found to be important where environmental and/or creep interactions are presented. In those instances, removal of B and Zr dramatically increases crack growth and it is therefore plausible that effective dilution of these elements may explain a previously observed trend in which crack growth rates increase with decreasing grain size.

Gayda, J.↗
DOE OSTI2023

Casting and Characterization of U-50Zr

Uranium alloyed with 50 wt% zirconium (U-50Zr) is a proposed light water reactor (LWR) nuclear fuel by Lightbridge Corporation (LTBR). The proposed method for making the U-50Zr alloy is to arc-melt master alloys and then remelt in a vacuum induction melter (VIM) to consolidate the material and cast into an intermediate shape. After casting, the material will eventually need to be formed into a desired fuel shape. The work discussed in this report resulted from a joint effort between Ltbr and Pacific Northwest National Laboratory (PNN) to investigate a 500g – 1kg scale casting process to produce the U-50Zr alloy in the desired δ-UZr 2 phase and characterize the impurities and microstructure that result from the casting process. Master alloys were fabricated in an arc melter, then five castings were carried out in a VIM with multiple inert coating and crucible materials to find an appropriate combination. Both ZrO 2 and graphite crucibles were used and different combinations of Y 2 O 3 , CaZrO 3 , and TiC to identify which would contain the molten metal with the least reaction. On each casting, the C, O, N, H impurities were analyzed as well as the phase by x-ray diffraction and microstructure. Of the five castings, two resulted in majority of δ-UZr 2 phase-pure material and had impurity levels within acceptable ranges. The two most successful castings utilized a graphite crucible with a TiC undercoating and a Y 2 O 3 overcoat. The O and N levels were below 1000 ppm and the C content was variable but did not result in measurable carbide formation. The highest success casting resulted in an average of 282 ppm C, 567 ppm O, 217 ppm N and 79 ppm H. This casting's crucible and inert coating material was repeated with slightly different casting parameters and resulted in higher C numbers but similar phase identification. The differences between each casting are discussed and recommendations are made for future experiments to better decide on a casting process to go forward with.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1958

Boron and Zirconium from Crucible Refractories in a Complex Heat-Resistant Alloy

In a laboratory study of the factors involved in the influence of induction vacuum melting on 55ni-20cr-15co-4mo-3ti-3al heat resistant alloy, it was found that the major factor was the type of ceramic used as the crucible. The study concluded that trace amounts of boron or zirconium derived from reaction of the melt with the crucible refactories improved creep-rupture properties at 1,600 degrees F. Boron was most effective and, in addition, markedly improved hot-workability.

Decker, R F↗
NASA NTRS2003

Filler wire for aluminum alloys and method of welding

A weld filler wire chemistry has been developed for fusion welding 2195 aluminum-lithium. The weld filler wire chemistry is an aluminum-copper based alloy containing high additions of titanium and zirconium. The additions of titanium and zirconium reduce the crack susceptibility of aluminum alloy welds while producing good weld mechanical properties. The addition of silver further improves the weld properties of the weld filler wire. The reduced weld crack susceptibility enhances the repair weldability, including when planishing is required.

Bjorkman, Jr., Gerald W. O.↗
DOE OSTI2023

Atom Probe characterization of neutron irradiated commercial ZIRLO® and AXIOM X2® alloys

As part of the Mechanistic Understanding of ZIrconium Corrosion (MUZIC) program, this team has been instrumental to obtain EPRI and Westinghouse sponsorships to prepare the neutron irradiated ZIRLO® and X2® from Vogtle nuclear power plant and ship them from Studsvik’s (Sweden) hot cells to Idaho National Laboratory (INL) to be part of the NSUF library. The uniqueness of these sets of neutron irradiated alloys is that they are at the same two extremes of the fuel cycles, allowing us to study the microchemistry evolution as a function of irradiation doses, exposure time and solute content. Advanced commercial Zirconium-Niobium (Zr-Nb) alloys, such as ZIRLO® and AXIOM®, have been developed by Westinghouse to enhance the corrosion resistance of fuel cladding material for longer service time. This study specifically aims at investigating the neutron irradiation induced Nb redistribution in AXIOM® X2® to understand the effect of neutron irradiation induced microchemistry changes and the irradiation enhanced corrosion resistance observed in ZrNb alloys. The detailed sample information on chemical composition and irradiation is given in shown in Table 1a and 1b respectively. To study the neutron IIP/nanoclusters and Nb concentration in the solid solution, atom probe tomography (APT) is the primary tool to obtain reliable chemical information. APT specimens were prepared in shielded focused ion beam (FIB) at the Irradiated Materials Characterization Laboratory (IMCL) at INL followed by APT analysis at CAES facility using LEAP 4000XR

36 MATERIALS SCIENCE↗
NASA NTRS1987

The effect of sulfur and zirconium Co-doping on the oxidation of NiCrAl

The adhesion behavior of Al2O3 scales formed on NiCrAl+Zr alloys was examined as a function of both sulfur and zirconium doping levels. In general, very high levels of zirconium were required to counteract the detrimental effects of sulfur. A sulfur-zirconium adherence map was constructed, as determined from the oxidation and spalling behavior in 1100 C cyclic tests. For low sulfur alloys, the amount of zirconium required for adherence at any given sulfur level can be described by Zr greater than 600 S sup 0.2 (in ppma). These results underscore the importance of sulfur to adhesion mechanisms and suggests that sulfur gettering is a first order effect of reactive element additions to MCrAl alloys.

Smialek, James L.↗
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↗
DOE OSTI2021

Fundamental Studies of Tritium Formation and Diffusivity in Pure and Defective Zircaloy-4 Getters

Zirconium (Zr) and its alloys are used as fuel cladding in nuclear reactors. These materials have their low absorption cross-section for thermal neutrons and have good mechanical and thermal properties. In addition, they are used as tritium getters in tritium-producing burnable absorber rods (TPBARs), in the form of nickel-plated Zircaloy-4 tubes. To produce 3 H, the breading blankets such as lithium aluminate are irradiated in pressurized water. 3 H produced in this process is captured by the getter and 3 H reacts chemically with Zr to form metal hydride (ZrTx). The precipitation of metal hydride results in not only the volume increase and hydrogen embrittlement of zircaloy but also changes in getter’s chemical properties. This degrades the getter’s performance. Therefore, understanding the behavior of 3 H (such as diffusion and solubility) in zircaloy is important to understand getter performance.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗