Search NASA⌕ Search

SEARCH · Search NASA

Results for “ZIRCONIUM ALLOY”

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 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Analyzing the effect of pressure on the properties of point defects in $\gamma\mathrm{U–Mo}$ through atomistic simulations

Uranium–molybdenum (U–Mo) alloys in monolithic fuel foil are the primary candidate for the conversion of high-performance research reactors in the USA. Monolithic fuel is utilized in a plate-type design with a zirconium diffusion barrier and aluminum cladding. These fuel types are unique in that they contain no plenum for the release of fission gases, which, in conjunction with the aluminum cladding, can lead to large stress states within the fuel. The nature of how fundamental processes of radiation damage, including the evolution of point defects, under such stresses occur is unknown. In this work, we present molecular dynamics simulations of the formation energy of point defects under applied stress. We report this work will allow for the implementation of stress-dependent microstructural evolution models of nuclear fuels, including those for both fission gas bubble growth and creep, which are critical to ensure the stable and predictable behavior of research reactor fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Hydrogen Localization and Cluster Formation in α-Zr from First-Principles Investigations

We report alpha-zirconium (α-Zr) is a hydrogen atom host material used in nuclear energy and semiconducting applications due to Zr's excellent properties such as corrosion resistance, low thermal neutrons cross-section, and thermodynamically favorable interstitial sequestering of H atoms. Herein we use density functional theory simulations to systematically analyze the initial stages of H insertion in the Zr matrix, and determine optimal hydrogen localization, distribution, and clustering tendencies from a thermodynamics and kinetics perspective. Our results suggest that H occupation of tetrahedral sites is strongly favored, consistent with previous experimental and theoretical studies. Further, a cooperative effect in H localization is shown where H atoms favor clustering. The H pair is found to be the most stable by occupying tetrahedral sites with pairs at second nearest-neighbor positions (TTn2). Among the four H atom clusters, the most favorable configuration has two TTn2 pairs forming "box " or "zig-zag" shaped-cluster with the pairs separated by 3.03 or 3.25 Å, respectively. Nudged elastic band simulations suggest that the formation of TTn2 is facile and is limited by the small diffusion barrier of the monomer of less than 0.1 eV. Further, the formation routes for the 4H atom clusters are also presented. The current study can serve as a foundation to systematically model H atom clustering in metals and alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Thermodynamically informed priors for uncertainty propagation in first-principles statistical mechanics

Here, this work demonstrates how first-principles statistical mechanics approaches within a Bayesian framework can quantify and propagate uncertainties to downstream thermodynamic calculations. To address the issue of Bayesian prior selection, knowledge of 0 K ground states in the material system of interest is incorporated into the prior. The effectiveness of this framework is shown by creating a phase diagram for the fcc zirconium nitride system, including confidence intervals on order-disorder transition temperatures.

Bayesian methods↗
DOE OSTI2022

Thermal Neutron Scattering Law Evaluation for Zirconium Carbide and Critical Mass Calculations

Zirconium carbide (ZrC) is a candidate material for use in advanced high temperature reactors, including space nuclear thermal propulsion applications. Thermal neutron scattering laws (TSLs) are generated for carbon bound in ZrC, C(ZrC), and zirconium bound in ZrC, Zr(ZrC), using ab initio lattice dynamics methods. These evaluations are to be submitted for inclusion in ENDF/B-VIII.1 and use the incoherent approximation for inelastic scattering as well as the new mixed elastic scattering treatment. The application of disordered alloy theory is introduced to appropriately capture the isotopic composition of Zr and C in the elastic scattering cross section. Localized higher energy vibrations in the C(ZrC) phonon density of states that are separated from lower energy modes result in quantized oscillations in the inelastic contributions to the TSL with a significant likelihood of large energy down-scattering and up-scattering interactions, where the latter increases in probability with temperature. The quanta of energy transfer during neutron thermalization is substantially greater than classically expected within the thermal neutron energy range. MC21 critical mass calculations of ZrC mixtures with high-enriched uranium demonstrate an impact of the TSLs when compared to free-gas treatment for 235 U concentrations less than 0.2 g/cm 3 . Additional MC21 critical mass calculations with homogenous mixtures of ZrC and reactor-grade graphite also demonstrate sensitivity to the ZrC TSL for thermal spectrum driven fission systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Design of a 2 MW ZrC/W-based molten-salt-to-sCO 2 PCHE for concentrated solar power

To increase the power cycle efficiency and lower the levelized cost of electricity (LCOE) of concentrated solar power (CSP) plants, printed circuit heat exchangers (PCHEs) capable of operating above 700 °C with molten chloride salt and a sCO 2 -based fluid are needed. In this paper, the design of a high-pressure, high-temperature, 2 MW PCHE comprised of a thermomechanically-robust, zirconium carbide/tungsten (ZrC/W) composite is conducted for CSP plants. The ZrC/W composite is a material with high thermal conductivity, high stiffness, and high failure strength at high temperatures, along with excellent resistance to thermal cycling and thermal shock. In this work, a thermomechanical design analysis was conducted to select appropriate material thicknesses of the ZrC/W plates, and to determine the geometrical dimensions and the thermal performance of the PCHE. The influences of the plate number and heat exchanger length on power density and pressure drop have also been systematically investigated. Economic analyses were conducted to compare the cost of ZrC/W-based PCHEs to those comprised of IN740H (a state-of-the-art, nickel-based superalloy) and 316 stainless steel (316SS). Finally, at a sufficiently high plate production rate, the manufacturing cost of ZrC/W-based PCHEs can be significantly lower, while achieving a much higher power density, compared with state-of-the-art, nickel alloy-based and stainless steel-based PCHEs.

14 SOLAR ENERGY↗
DOE OSTI2021

Effect of alloying with Sc, Nb and Zr on reduction-diffusion synthesis of magnetically hard Sm(Fe,Co,Ti) 12 -based monocrystalline powders

Powders of Sm(Fe,Co) 11.2 Ti 0.8 alloys modified with Sc, Nb and Zr, as well as with additional Ti were prepared by reducing mechanically activated raw oxides with Ca metal in the furnace preheated to 990–1250 °C. Expansion of the crystal lattice upon introduction of Nb or additional Ti implies that atoms of these elements replace the smaller Fe atoms in the tetragonal ThMn 12 -type structure. On the other hand, contraction of the lattice upon introduction of Sc or Zr was smaller than what was expected for replacement of the Sm atoms, which suggests that the Sc and Zr atoms replace both the Sm and Fe atoms. Washing away the reduction byproducts expands the crystal lattice of the 1:12 particles and increases their coercivity. The lattice expansion associated with the washing is believed to be caused by interstitial H atoms; more research, however, is needed to establish the mechanism(s) of the washing effect on the coercivity. The earlier reported development of a high coercivity in zirconium-modified monocrystalline particles achieved by increasing the reduction annealing temperature to ≈1200 °C was similarly characteristic of the particles modified with Sc (the coercivity reaches 11.5 kOe) and Nb (8.1 kOe), but not for the particles prepared with additional Ti where the maximum coercivity of 8.3 kOe develops for a lower annealing temperature. Furthermore, it is concluded that Sc, Nb and Zr modify the high-temperature phase equilibria of the Sm(Fe,Co) 11.2 Ti 0.8 alloys allowing for an effective high-temperature processing, whereas the alloy coercivity increases with the synthesis temperature through a different, still unknown mechanism which may involve suppression of the defects specific to the 1:12 crystals.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Designing an impedance-based technique for studying corrosion on cladding materials

In nuclear cladding materials, changes in structure and chemistry over the lifetime of claddings can have a significant impact on fuel and reactor performance. Currently, chemical evolution over the lifetime of the fuel and cladding materials can only be inferred using post-irradiation examination. As an alternative to post-irradiation measurements, electrochemical impedance spectroscopy (EIS) can provide an in-pile measurement of changes in chemistry in oxides and hydrides. The objective of this study is to develop an impedance based sensing technology for measuring spatial- and time-resolved in-pile changes in cladding chemistry in order to understand the mechanisms of hydride formation, hydride dealloying and associated cladding corrosion so that a mitigation strategy and technology can be developed for reactor safe operation. Two prototype impedance-based sensors were designed and assembled using Zr-4 as working electrode and Pt as counter and reference electrode. One sensor is tubular type and the other one is plate-type. These sensors will be tested in static autoclave to investigate the effects of various environmental conditions (including temperature, pressure, and water chemistry) on cladding material corrosion mechanisms. These in-situ EIS sensors would provide corrosion-layer parameters in high-temperature water, such as (1) total layer thickness, (2) the evolution of oxide structures as uniform layer dielectric or sublayer(s), and (3) the possible electrochemistry of the volume corrosion process and the solid/liquid interface, such as the cyclic nature of corrosion of Zr alloys in high-temperature water. This work was focused on electrochemical sensor development, targeting cladding oxides/hydrides under thermodynamic equilibria; characterization of samples using Kelvin probe force microscopy (SKPFM), Raman spectroscopy, and scanning electron microscopy (SEM) with energy-dispersive X-ray spectroscopy (EDS) to provide high-resolution co-localized characterization of zirconium oxide and metal/oxide interface.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Radiation Effects on the Performance of Advanced Sulfur Monochloride Chlorination Processes

Advanced sulfur chloride-based chlorination technologies are being developed to enable efficient recycling of aluminum and zirconium-based materials used in the nuclear industry. However, the impacts of ionizing radiation on the performance of these sulfur chloride compounds are not well established, despite this being critical knowledge for assessing their feasibility and longevity under envisioned process conditions. Here, in the present article, we report on the effects of cobalt-60 gamma irradiation (≤ 5 MGy) on the aluminum alloy 6061 (AA6061-T6) chlorination yield in sulfur monochloride (S 2 Cl 2 ). Our findings indicate that, compared to nonirradiated solvent, radiation-induced changes in the chemical composition of S 2 Cl 2 —identified using Raman spectroscopy—afford an additional, dose-dependent exothermic process prior to the chlorination reaction’s typical thermodynamic behavior. We attribute this new process to reactions involving aluminum species (metal, oxide, or [oxy]hydroxides) and sulfur dichloride (SCl 2 ), an S 2 Cl 2 radiolysis product that accumulates with absorbed gamma dose, but is absent following an AA6061-T6 chlorination study. Despite the exothermicity of this new process, the overall yield of chlorination decreased with increasing preirradiation dose. Consequently, the chemical reactivity, specificity (aluminum metal vs aluminum passivation and corrosion layer constituents), and byproducts of SCl 2 must be more thoroughly evaluated to support the continued development of advanced S 2 Cl 2 chlorination technologies.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Exploring laser-material interactions of zirconium carbide under additive manufacturing conditions

Zirconium carbide (ZrC) is an ultra-high temperature ceramic with a melting temperature above 3000°C and a broad range of high temperature applications. Given the high melting and sintering temperatures of pure ZrC, producing near-net shape and fully dense parts remains challenging with conventional techniques. In this study, we investigate the fundamental laser-material interactions of ZrC under laser powder bed fusion (LPBF) additive manufacturing (AM) conditions. Normalized enthalpy, a scaling law term that is used in welding and AM literature for detailing laser-material interactions in metallic alloys, was calculated to determine the predictive capabilities of melt pool features in ZrC. Further, the melt pool quality of laser irradiated ZrC was used to compare LPBF relevant laser parameter combinations of laser power, scan speed, and beam diameter. Laser build parameters that resulted in desirable melt pool morphologies were applied to the fabrication of ZrC coupons using LPBF AM. A custom LPBF system was used to determine hatch spacing and layer height parameters that resulted in a fabricated sample with a density of 85% as measured by Archimedes and a Vicker's microhardness of 20.9 ± 1.9 GPa. This investigation reveals the laser-material interactions of ZrC under AM relevant conditions and is the first step towards LPBF fabrication of ZrC parts.

36 MATERIALS SCIENCE↗
DOE OSTI2022

On the efficacy of post-build thermomechanical treatments to improve properties of Zirconium fabricated using ultrasonic additive manufacturing

Here, hot-isostatic pressing has been applied to Zirconium plate fabricated using the ultrasonic additive manufacturing (UAM) technique to enhance interfacial bond quality. Specimens heated to 800 C for 1 h at 100 MPa pressure showed grain growth across many prior foil-to-foil interfaces, thereby increasing foil adhesion. In addition to the material softening induced by a loss of Hall-Petch strengthening, premature failure of specimens loaded parallel to the build direction was observed. Premature specimen failure was attributed to the local delamination at prior foil-to-foil boundaries where grain growth was pinned by Ti impurities introduced during the UAM process via the interaction between a Ti-alloy buffer foil and the welded Zr-foils underneath. In addition, the presence of Ti along select foil interfaces resulted in the nucleation and growth of secondary (Zr,Ti)(Fe,Cr) laves phases during higher-temperature thermomechanical processing. Using a combination of micro-scale X-ray computed tomography, fractography, and in-situ digital image correlation, the effect of defects along prior foil-to-foil boundaries was revealed as delamination-assisted plasticity accelerated specimen failure in preferred tensile orientations. These findings underscore the importance of impurity control when optimizing weld quality of higher-strength material systems using ultrasonic welding.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Benchmark Gap Assessment for the Manufacturing of High-Assay Low-Enriched Uranium Fuels

This document develops basic critical conditions for spheres—moderated and unmoderated, as well as reflected and unreflected—in consideration of nuclear criticality safety of a potential fuel production facility producing high-assay low-enriched uranium (HALEU) fuel of several different types like tristructural-isotropic (TRISO), uranium metal and alloys, oxide and non-metallic forms. In addition to spherical arrangements, TRISO particle manufacturing process–specific equipment is modeled as it would be for the criticality safety analysis. The objective is to develop representative systems that can then be used for comparison with existing benchmarks. SCALE/TSUNAMI is used to assess the similarity index between these systems to assess validation gaps for possible fuel production applications of proposed advanced reactors. Several different fuel types were evaluated, including TRISO, uranium metal, uranium molybdenum, uranium zirconium, uranium dioxide, uranium nitride, uranium hydride, U-ZrH, and uranium chloride. This selection of fuel types covers a breadth of proposed reactor types, as well as intermediate steps in the production and fabrication of the fuel

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Probing the Radial Chemistry of Getter Components in Light Water Reactors via Controlled Electrochemical Dissolution

Getters are among the key functional components in the Tritium Producing Burnable Absorber Rods (TPBARs) of Light Water Reactors, and are used to capture the released tritium gas. They are cylindrical tubes comprised of the alloy zircaloy-4 which is coated with nickel plating both internally and externally. Their exposure to tritium as well as the irradiation environments in the light water reactors changes their structure, chemical composition and chemistry. Understanding the radial tritium distribution is key to gaining insight into the evolution of new chemistry upon irradiation in order to predict getter performance. Consequently, efforts are underway to devise a method that can allow selective probing into the getter core radially to probe the chemistry akin to carefully and selectively peeling of the layers of an onion. We have recently devised an electrochemical dissolution technique for controlled radial dissolution of zirconium based liners; this work describes the first proof-of-concept demonstration of the applicability of this technique on the controlled radial dissolution of nickel plated getters. The technique is complemented with periodic analysis of the electrolyte solution to measure the rates of release of the various elemental components and therefore glean information on their radial distribution across the getter; further complementary microscopy on the exposed surfaces allows us insight into the chemistry at the getter core.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

A critical analysis of U-Pu-Zr phase transitions using calorimetric, microstructural, and phase equilibria data

Metallic fuels consisting primarily of uranium, plutonium, and zirconium (U-Pu-Zr) are a leading material candidate for fast-spectrum nuclear reactors. Early demonstration programs proved the principle of safe and efficient fast reactor operation, however there is still considerable uncertainty regarding the phase equilibria and microstructural evolution across the ternary composition space. Quantitative phase formation and identification measurements are scarce and often incomplete, with studies reporting either phase transition temperatures or phase identification data, but not both from the same specimens. In this study, we critically compared experimental and calculated phase transition data and correlated with the microstructure and phase characterization data of as-cast and annealed U-Pu-Zr alloys. Differential scanning calorimetry (DSC) was used to measure phase transitions in the subsolidus regions (723−948 K) of three ternary U-Pu-Zr alloys with similar plutonium concentrations but various U/Zr ratios. Due to sluggish kinetics and narrow ranges of phase stability, complex peaks required the use of a Frazier-Suzuki peak fitting algorithm to deconvolute and calculate transition peak temperatures and enthalpies. We also identified trends of phase transition behavior by critically comparing our DSC data with previous phase transition measurements as well as historical and calculated phase equilibrium diagrams. In conclusion, this provides a critical approach for benchmarking and assessing the quality of new U-Pu-Zr phase equilibria data prior to its incorporation into nuclear material databases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Development of Chemical Procedures for Isotope Harvesting: Separation of Trace Hafnium from Tungsten

The separation of trace hafnium from bulk tungsten alloys is of interest for isotope harvesting at the National Superconducting Cyclotron Laboratory and the Facility for Rare Ion Beams because 172 Hf, the parent of 172 Lu, accumulates in tungsten alloy beam blockers at these facilities. In this work, a procedure for the separation of trace hafnium from a bulk tungsten alloy (454 g) was established using tracer isotopes ( 175 Hf, 88 Zr, 173 Lu, and 88 Y). Here, the procedure employed dissolution in an HF–HNO 3 solution followed by a calcium fluoride precipitation, and then extraction chromatography was used for more selective separation steps. Two stages of column separations using LN resin (HDEHP based) and TRU resin (CMPO based) were performed. Gamma-ray spectroscopy and mass spectrometry were used to analyze the final hafnium sample and follow the steps in the chemical processing. Furthermore, the final recovery of hafnium was high (90 ± 8)%, and the mass of tungsten and other transition metals was reduced to near background levels (as determined by ICP-MS of the blank acid solutions). Zirconium follows hafnium quantitatively in this procedure; there was no detectable 173 Lu in the final hafnium sample.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

Graph-theoretical KMC Framework for Calculating Effective Diffusivity in TPBAR Components: Effective Diffusivity of Tritium in α-Zr Grain Boundaries

We report the development of a computer simulation tool based on the Graph-theoretical kinetic Monte Carlo (GT-KMC) approach that can be used to simulate diffusion and derive effective diffusivity of species in complex structures of TPBAR components. The GT-KMC framework was implemented in AKSOME, an on-lattice self-learning kinetic Monte Carlo tool developed at PNNL and was previously used to study solute diffusion in metal alloys. This newly developed tool will be helpful in simulating atom diffusion and extracting their effective diffusivities in grain boundaries and interfaces in various TPBAR components. Once benchmarked, the tool will be used to simulate the diffusion of tritium (hydrogen) along grain boundaries in α-Zr using the activation energy barrier data previously obtained from Density Functional Theory calculations.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Status Report on Fast Flux Test Facility Mechanistic Fuel Failure Experiment Analysis with BISON for Post Irradiation Examination Support

The renewed interest in metallic U-Zr nuclear fuel alloy has led to a drive for deeper understanding of the mechanisms driving the phenomena observed under irradiation conditions. The Department of Energy Advanced Fuel Campaign has developed infrastructure to support metallic fuel development, including Post Irradiation Examination (PIE) of legacy Fast Flux Test Facility (FFTF) Mechanistic Fuel Failure (MFF) experiments. The PIE performed on legacy FFTF MFF experiments gives insight on metallic fuel performance and can address the lack of knowledge and scarcity of reliable data identified in several studies over recent years. Unfortunately, PIE efforts can cost significant time and resources which can impede the progress of metallic U-Zr fuel development. Metallic U-Zr fuel performance modeling can be used to inform PIE efforts on regions of interest for relevant investigations and can help understand phenomena observed in PIE. This report demonstrates the current progress of FFTF MFF fuel performance simulations using the BISON fuel performance code and discusses the support provided by simulation to PIE efforts. Progress in temperature, profilometry, fission gas release, plenum pressure, and zirconium redistribution simulation results have been demonstrated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗