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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 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Irradiation creep measurement and microstructural analysis of chromium nitride–coated zirconium alloy using pressurized tubes

Environmental barrier coatings for Zr-based materials are currently under development to reduce oxidation and embrittlement in light-water reactors. Chromium nitride is one such candidate for this application, particularly as accident-tolerant fuel cladding. However, quantifying the impact of coatings on the irradiation-induced creep of zircaloy (Zry) is critical as this mechanism often exceeds thermal creep rates under light-water reactor operating conditions and can be a limiting design characteristic. Additionally, examining irradiation effects in the microstructure at the coating interface is key to understanding the compatibility of the material system. Here, to accelerate the experimental measurement of irradiation creep and microstructure evolution in CrN-Zry, compact, pressurized creep tubes were fabricated and irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Miniature, thin-walled rodlets fabricated from annealed Zr-Sn barstock were coated with CrN using physical vapor deposition (PVD) to nominal thicknesses of 4 and 8 μm. Coated and uncoated rodlet specimens were internally pressurized and welded, generating nominal circumferential hoop stresses of 0, 90, or 180 MPa under 300°C irradiation conditions. Twelve specimens were measured diametrically prior to irradiation using a low-cost, automated, contactless laser profilometer developed for this work. Specimens were irradiated in sealed capsules for one 25-day HFIR cycle, accumulating approximately 1.8 $\times$ 10 21 n/cm 2 fast fluence (E n > 10 MeV). The irradiated samples were retrieved and remeasured using the same profilometry system in a shielded hot cell facility. Irradiation creep between specimens was compared using standard statistical tests and showed that both thicknesses of CrN coating had a negligible effect on the irradiation creep strain of the Zry material. Microstructure characterization of pre- and post-irradiated CrN-Zry specimens showed minimal changes due to irradiation but did show a substantial O-rich region at the Zry-CrN interface.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Development of Casting Techniques for d-phase Uranium-Zirconium Alloys: CRADA 524 [Abstract only]

Casting of d-phase HALEU UZr2 alloy is an early milestone in the fabrication process for Lightbridge Fuel™. The Radiochemical Processing Laboratory at PNNL has been identified as having both the capability and expertise to perform the necessary high temperature casting of samples of Lightbridge’s alloy. The necessary equipment, facility licensing, and shipping capabilities for special nuclear material coupled with its expertise in uranium casting techniques makes PNNL uniquely suited to this project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1966

Copper-acrylic enamel serves as lubricant for cold drawing of refractory metals

Acrylic enamel spray containing metallic copper pigment lubricates refractory metal tubing during cold drawing operations so that the tubing surface remains free from scratches and nicks and does not seize in the die. Zirconium alloys, zirconium, tantalum alloys, niobium alloys, vanandium alloys and titanium alloys have been drawn using this lubricant.

Beane, C.↗
DOE OSTI2023

Evaluation of Nickel Alloys, Manufactured through Powder Bed Fusion, for Application for Advanced Nuclear Reactors

In collaboration with other laboratories, Idaho National Laboratory has investigated additively manufactured alloys for applications in advanced nuclear reactors. The primary focus has been on nickel alloys, as well as “other” alloys, which are alloys not considered steel nor nickel-based alloys, manufactured by laser powder bed fusion. Suitability of investigated alloys were considered against a matrix of properties of interest, developed jointly with the collaborating laboratories. For the nickel alloys, Alloys 282 and 625 are considered promising due to their properties, existing data, and availability of commercial powder feedstock. Haynes 244 is also of interest for its molten salt corrosion resistance, however there is no current feedstock supply chain, and no experience with this alloy as an additively manufactured material. Iron-chrome-aluminum alloys, titanium alloys, zirconium alloys, ceramics, and ceramic inclusions were all considered within the other alloys category. The results and discussions within this report will be compiled with the investigations from the other collaborating laboratories and compiled into a final report later this year (2023).

36 MATERIALS SCIENCE↗
NASA NTRS1986

Nickel base coating alloy

Zirconium is added to a Ni-30 Al (beta) intermetallic alloy in the range of 0.05 w/o to 0.25 w/o. This addition is made during melting or by using metal powders. The addition of zirconium improves the cyclic oxidation resistance of the alloys at temperatures above 1100 C.

Barrett, C. A.↗
NASA NTRS1986

Characterization of precipitates in a niobium-zirconium-carbon alloy

A niobium alloy with 1 percent zirconium and 0.063 percent carbon by weight was investigated in the as-rolled and annealed conditions, and after high-temperature (1350 and 1400 K) exposure with and without an applied stress. In the as-rolled and annealed conditions, large metastable carbides were observed in addition to a regular distribution of small particles. During the high-temperature exposure, the majority of the large carbides were dissolved and a more stable carbide phase formed. This finely dispersed phase had a composition determined to be approximately 70 percent ZrC and approximately 30 percent NbC and showed some evidence of an orientation relationship with the matrix. The precipitates appeared to coarsen slightly after approximately 5000 hr exposure in the presence of an applied stress resulted in a decrease in the size and in the interparticle spacing of the stable precipitates. However, the composition of the precipitate phase and its ability to pin dislocations were not affected by the temperature or stress conditions.

Grobstein, T. L.↗
DOE OSTI2024

Diffusion bonding of tungsten-vanadium-zirconium using vacuum hot pressing for the development of a low decay heat cladding solution for tungsten spallation targets

Tantalum has been used as cladding material for water-cooled solid tungsten targets at many leading spallation neutron production facilities thanks to its high neutron yield, manageable radiation damage behavior, and excellent corrosion/erosion resistance in radiation environments. However, from a safety hazard perspective, thermal neutron capture of tantalum in spallation environments causes a high specific decay heat in the target volume, which often becomes a limiting factor in increasing the beam power on the target. In this paper, we studied vacuum hot pressing (VHP) parameters to diffusion bond zirconium to tungsten to explore the feasibility of using zirconium alloys as an alternative cladding material to tantalum. Zirconium alloys have long been used as cladding material for early generation solid spallation targets, and nuclear fuel rods. In spallation environments zirconium has significantly lower decay heat with shorter decay time compared to tantalum. The hot isostatic pressing (HIP) of zirconium and tungsten is known to produce limited bonding quality due to the formation of the brittle ZrW2 intermetallic layer. To overcome this problem, placing a vanadium interlayer between tungsten and zirconium has been proposed by exploring parameter space in binary alloy phase diagrams. Under the VHP conditions, 860 ◦C at 70 MPa for 4 hours, Zr-V and V-W showed good diffusion bonding, which demonstrates the feasibility of a single step HIP process to make the zirconium alloy clad tungsten spallation volumes.

Lee, Yong Joong↗
NASA NTRS1969

Superconductivity in zirconium-rhodium alloys

Metallographic studies and transition temperature measurements were made with isothermally annealed and water-quenched zirconium-rhodium alloys. The results clarify both the solid-state phase relations at the Zr-rich end of the Zr-Rh alloy system and the influence upon the superconducting transition temperature of structure and composition.

Zegler, S. T.↗
DOE OSTI2020

Thermomechanical Processing of Uranium Alloys with 10 and 50 Weight Percent Zirconium

This work focuses on casting and thermomechanical processing for moderate sized (500-1000g) 10% by weight and 50% by weight uranium-zirconium alloys. Three castings of uranium – 10wt% zirconium and a fourth casting of uranium – 50wt% zirconium were completed using different castings processes. The first casting process relied on only induction mixing within the VIM system thereby eliminating the need for an arc-melting feedstock preparation step. The second casting process used the arc-melting process prior for feedstock to the same VIM system. In this study, the induction mixing process alone with a graphite crucible appears insufficient for the fabrication of UZr alloys. Graphite is suitable with the use of master alloy UZr feedstock produced by arc-melting. There was a moderate increase of approximately 100 ppm of carbon within the U10Zr alloys. However, there was a near 1000 ppm increase of carbon in the U50Zr alloy. This carbon content is considered too high for forming processes, and therefore, warrants the use of a ceramic crucible for UZr castings near 50wt% zirconium. Zirconia does not appear suitable for UZr castings near 50wt% zirconium. The use of Beryllia or Thoria may be a more suitable option. However, due to the significantly lower liquidus and lower zirconium content, Zirconia crucibles do appear viable for carbon reduction in the U10Zr system. Two thermomechanical processes are also studied in this work - hot extrusion of the U50Zr system and heat treatment in an argon atmosphere. This work demonstrated that extrusion of the U50Zr alloy near 680°C is a viable choice even with +1000 ppm carbon. Extrusion of the U50Zr system near 630°C was unsuccessful and does not appear viable unless significant force and tooling strength is available. There appears to be a moderate increase in extrudate hardness when compared to as-cast. Heat-treatment reduced the hardness of as-cast alloys in the U10Zr castings as well as resolve some variance in the microstructures. However, for the U50Zr billet that did not successfully extrude at 630°C, heat treatment at 900°C for 96 hours did not have a significant impact on hardness or microstructure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Progress Towards a Titanium-Zirconium-Molybdenum Alloy Coreblock Prototype Using Powder Bed Fusion

Microreactor designs make use of a coreblock for structural stability allowing for these devices to provide powder generation while being portable and inherently safe. In order to maximize energy output of the microreactors, materials must be able to withstand high temperature without melting and without substantial decreases in mechanical properties. As simultaneous projects looking at steels, graphite, and other materials are on-going, this work package will investigate titanium – zirconium – molybdenum alloy (TZM) as it allows for potentially higher microreactor operating temperatures than traditional reactor materials such as stainless steel. Notably, this alloy was developed in the 1960s for use in rockets and has a very high melting point (2623°C) but is difficult to form, machine and join. With limited needs in aerospace, literature is appropriately sparce in looking at the AM of TZM (examples include [1,2]). However, the properties appear to be an excellent match for the needs of the microreactor coreblock leading to the current project. In addition, small washers form the bulk of TZM parts needed in aerospace while the microreactor community was looking at larger, more complex parts. Therefore, AM of TZM is being explored to enable near – net shape of components with the needed complexity to impact the nuclear energy field.

36 MATERIALS SCIENCE↗