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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 1,621 records · Page 90
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Investigation of thermal aging effects on the tensile properties of Alloy 617 by in-situ synchrotron wide-angle X-ray scattering

The nickel-base Alloy 617 has been considered as the lead candidate structural material for the intermediate heat exchanger (IHX) of the Very-High-Temperature Reactor (VHTR). In order to assess the long-term performance of Alloy 617, thermal aging experiments up to 10,000 h in duration were performed at 1000 degrees C. Subsequently, in-situ synchrotron wide-angle X-ray scattering (WAXS) tensile tests were carried out at ambient temperature. M23C6 carbides were identified as the primary precipitates, while a smaller amount of M6C was also observed. The aging effects were quantified in several aspects: (1) macroscopic tensile properties, (2) volume fraction of the M23C6 Phase, (3) the lattice strain evolution of both the matrix and the M23C6 precipitates, and (4) the dislocation density evolution during plastic deformation. The property-microstructure relationship is described with a focus on the evolution of the M23C6 phase. For aging up to 3000 h, the yield strength (YS) and ultimate tensile strength (UTS) showed little variation, with average values being 454 MPa and 787 MPa, respectively. At 10,000 h, the YS and UTS reduced to 380 MPa and 720 MPa, respectively. The reduction in YS and UTS is mainly due to the coarsening of the M23C6 Precipitates. After long term aging, the volume fraction of the M23C6 phase reached a plateau and its maximum internal stress was reduced, implying that under large internal stresses the carbides were more susceptible to fracture or decohesion from the matrix. Finally, the calculated dislocation densities were in good agreement with transmission electron microscopy (TEM) measurements. The square roots of the dislocation densities and the true stresses displayed typical linear behavior and no significant change was observed in the alloys in different aging conditions.

Alloy 617↗
DOE OSTI2026

Process-driven roadmap for depositing super duplex stainless steel via wire Arc additive manufacturing

Here, this study systematically investigates the effects of shielding gas, bead spacing, weld mode, and travel speed on the phase balance, porosity, and hardness of wire arc additively manufactured (WAAM) super duplex stainless steel (ER2594). Robotic WAAM was employed to fabricate multilayer walls under systematically varied process conditions, followed by phase transformation simulations, X-ray computed tomography (XCT), electron backscatter diffraction (EBSD), and microhardness evaluation. Thermodynamic simulations predicted rapid cooling of AM process can suppress the potential formation of deleterious precipitates which was later validated via cross-sectional microstructure analyses of printed samples. XCT revealed porosity levels below 0.2% for all deposits, with reduced travel speed significantly lowering defect volume. Microstructural analyses revealed the evolution of various austenite precipitates, including grain boundary austenite (GBA), Widmanstätten austenite (WA), and intergranular austenite (IGA), sequentially upon cooling of the ferrite phase. Among all process parameters, weld transfer mode exhibited the strongest influence on phase balance; pulsed mode promoted higher ferrite retention (~ 36%) compared to RapidX mode. No consistent relationship between stabilized phase fraction and captured microhardness was observed. This work provides critical insights for optimizing WAAM parameters to control phase balance and mechanical performance, which is essential for producing high-integrity super duplex stainless-steel components for nuclear and marine applications.

Grain orientation↗
DOE OSTI2020

Effects of irradiation spectrum on the microstructural and mechanical properties of bulk metallic glasses

Zr 52.5 Cu 17.9 Ni 14.6 Al 10 Ti 5 (BAM-11) and Cu60Zr 20 Hf 10 Ti 10 bulk metallic glass (BMG) specimens were ion irradiated to characterize and compare their irradiation induced microstructural and mechanical property evolution. For the ion irradiations, samples of each BMG were exposed to 9 MeV Ni 3+ and 5.5 MeV C + ions to a midrange (~1.5 μm depth) dose of 0.5 displacements per atom (dpa) at temperatures ranging from room temperature to 360 °C in order to examine potential effects associated with low (5.5 MeV C) and high (9 MeV Ni) average primary knock on energies. BAM-11 BMG samples were also irradiated by neutrons (E > 0.1 MeV) at ~70 °C to a fluence of 1.4 × 10 20 n/cm 2 (dose of 0.1 dpa). Importantly, no pronounced irradiation spectrum effects were observed for the nanoindentation mechanical properties of the two BMGs following ion irradiation. However, significant softening was observed in the BAM-11 BMG sample irradiated by neutrons, while annealing of the unirradiated samples led to a marked increase in hardening. Finally, results of the nanoindentation experiments indicate that softening is caused by irradiation induced creation of free volume defects, while the hardening is caused by their annihilation.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Microstructure and Mechanical Properties of Spray-Formed 7075 Alloy During Retrogression

The mechanical properties, conductivity, microstructure, precipitation behavior and evolution process of elements and phases of spray-formed 7075 alloy during retrogression are studied by the tensile test, electric test, hardness test, transmission electron microscope (TEM), x-ray diffraction (XRD) analysis, electron-dispersive spectroscopy (EDS) and lattice parameter measurements. The results show that the mechanical properties and conductivity of the alloy can be effectively improved when treated at a suitable retrogression temperature and time. According to the TEM observations, XRD analysis, EDS and an accurate measurement of the lattice parameter, the redissolution of elements and precipitated phases can be readily observed. From the optimization selection of the RRA treatment for the 7075 alloy, the regularity of redissolution and precipitation of the elements and phases are evaluated. The ultimate tensile strength, yield strength, hardness, elongation and conductivity of the spray-formed 7075 alloy after preaging at 120 °C for 16 h, retrogression at 200 °C for 8 min and reaging at 120 °C for 24 h are 791, 736 MPa, 181 HBW, 8.5% and 22.6 MS/m, respectively.

36 MATERIALS SCIENCE↗
DOE OSTI2021

A coupled fluid-mechanical workflow to simulate the directed energy deposition additive manufacturing process

Simulation of additive manufacturing processes can provide essential insight into material behavior, residual stress, and ultimately, the performance of additively manufactured parts. In this work, we describe a new simulation based workflow utilizing both solid mechanics and fluid mechanics based formulations within the finite element software package SIERRA (Sierra Solid Mechanics Team in Sierra/Solid Mechanics 4.52 User’s Guide SAND2019-2715. Technical report, Sandia National Laboratories, 2011) to enable integrated simulations of directed energy deposition (DED) additive manufacturing processes. In this methodology, a high-fidelity fluid mechanics based model of additive manufacturing is employed as the first step in a simulation workflow. This fluid model uses a level set field to track the location of the boundary between the solid material and background gas and precisely predicts temperatures and material deposition shapes from additive manufacturing process parameters. Next, the resulting deposition shape and temperature field from the fluid model are then mapped into a solid mechanics formulation to provide a more accurate surface topology for radiation and convection boundary conditions and a prescribed temperature field. Solid mechanics simulations are then conducted to predict the evolution of material stresses and microstructure within a part. By combining thermal history and deposition shape from fluid mechanics with residual stress and material property evolutions from solid mechanics, additional fidelity and precision are incorporated into additive manufacturing process simulations providing new insight into complex DED builds.

36 MATERIALS SCIENCE↗
DOE OSTI2020

In situ X-ray measurements over large Q-space to study the evolution of oxide thin films prepared by RF sputter deposition

Sputter deposition is a versatile and industrially important deposition technique for thin films, with increasing demand for matching the characteristics of thin film materials to specific requirements. Additionally, the actual film properties are largely determined by sputtering parameters such as pressure conditions, temperature and power settings. By means of various X-ray diffraction and scattering techniques, it is shown that the characterization of film formation and growth is feasible in real time at synchrotron sources, thus adding an important dimension to the fundamental understanding of the evolution of thin film microstructure. In particular, grazing incidence small-angle X-ray scattering, grazing incidence X-ray powder diffraction and X-ray reflectometry are used in a complementary manner to study the influence of deposition temperature and substrate choice on the crystallization kinetics and growth of polycrystalline BaTiO 3 films.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Role of microstructure on twin nucleation and growth in HCP titanium: A statistical study

In this study, a detailed statistical analysis is performed using Electron Back Scatter Diffraction (EBSD) to establish the effect of microstructure on twin nucleation and growth in deformed commercial purity hexagonal close packed (HCP) titanium. Rolled titanium samples are compressed along rolling, transverse and normal directions to establish statistical correlations for {10–12}, {11–21}, and {11–22} twins. A recently developed automated EBSD-twinning analysis software is employed for the statistical analysis. Finally, the analysis provides the following key findings: (I) grain size and strain dependence is different for twin nucleation and growth; (II) twinning statistics can be generalized for the HCP metals magnesium, zirconium and titanium; and (III) complex microstructure, where grain shape and size distribution is heterogeneous, requires multi-point statistical correlations.

36 MATERIALS SCIENCE↗
DOE OSTI2024

An integrated modeling framework with open architecture for phase field simulation of multi-component alloys

An integrated modeling framework (PanPhaseField) has been developed, which enables a direct and fast coupling between CALPHAD calculations and large-scale phase field simulations for multi-component alloys. Further, it adopts an open architecture allowing for integration of user-defined phase field models in a plug-and-play manner by taking full advantage of the user-friendly graphical interface of Pandat software. The developed modeling platform becomes an enabling tool that can be used to simulate the evolution of spatially varying microstructures of industrial complex alloys for various engineering applications.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Mechanism-based modeling of solute strengthening: Application to thermal creep in Zr alloy

In this paper, a crystallographic thermal creep model is proposed for Zr alloys that accounts for the hardening contribution of solutes via their time-dependent pinning effect on dislocations. The core-diffusion model proposed by Soare and Curtin (2008a) is coupled with a recently proposed constitutive modeling framework (Wang et al., 2017, 2016) accounting for the heterogeneous distribution of internal stresses within grains. The Coble creep mechanism is also included. This model is, in turn, embedded in the effective medium crystallographic VPSC framework and used to predict creep strain evolution of polycrystals under different temperature and stress conditions. The simulation results reproduce the experimental creep data for Zircaloy-4 and the transition between the low (n~1), intermediate (n~4) and high (n~9) power law creep regimes. This is achieved through the dependence on local aging time of the solute-dislocation binding energy. The anomalies in strain rate sensitivity (SRS) are discussed in terms of core-diffusion effects on dislocation junction strength. The mechanism-based model captures the primary and secondary creep regimes results reported by Kombaiah and Murty (2015a, 2015b) for a comprehensive set of testing conditions covering the 500–600 °C interval, stresses spanning 14–156 MPa, and steady state creep rates varying between 1.5·10 -9 s -1 to 2·10 -3 s -1 . There are two major advantages to this model with respect to more empirical ones used as constitutive laws for describing thermal creep of cladding: 1) specific dependences on the nature of solutes and their concentrations are explicitly accounted for; 2) accident conditions in reactors, such as RIA and LOCA, usually take place in short times, and deformation takes place in the primary, not the steady-state creep stage. Finally, as a consequence, a model that accounts for the evolution with time of microstructure is more reliable for this kind of simulation.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Radiation tolerance of neutron-irradiated model Fe-Cr-Al alloys

The Fe Cr Al alloy system has the potential to form an important class of enhanced accident-tolerant cladding materials in the nuclear power industry owing to the alloy system's higher oxidation resistance in high-temperature steam environments compared with traditional zirconium-based alloys. However, radiation tolerance of Fe Cr Al alloys has not been fully established. In this study, a series of Fe Cr Al alloys with 10 18 wt % Cr and 2.9 4.9 wt % Al were neutron irradiated at 382 C to 1.8 dpa to investigate the irradiation-induced microstructural and mechanical property evolution as a function of alloy composition. Dislocation loops with Burgers vector of a/2 111 and a 100 were detected and quantified. Results indicate precipitation of Cr-rich is primarily dependent on the bulk chromium composition. Mechanical testing of sub-size-irradiated tensile specimens indicates the hardening response seen after irradiation is dependent on the bulk chromium composition. Furthermore, a structure property relationship was developed; it indicated that the change in yield strength after irradiation is caused by the formation of these radiation-induced defects and is dominated by the large number density of Cr-rich α' precipitates at sufficiently high chromium contents after irradiation.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Determining the effects of U/Pu ratio on subsolidus phase transitions in U-Pu-Zr metallic fuel alloys

Here, ternary alloys consisting primarily of uranium, plutonium, and zirconium (U-Pu-Zr) are among the leading candidate fuel systems considered for fast spectrum nuclear reactors. Despite historical operation data from the testing of U-Pu-Zr rods in the Experimental Breeder Reactor-II, considerable uncertainty about the evolution of phases and microstructure across the ternary composition space exists. Due to sluggish kinetics and other difficulties in handling metal actinide specimens, quantitative measurements of phase-transitions in U-Pu-Zr alloys remain sparse in scientific literature, with most investigators reporting either phase-transition temperatures or phase identification data, but not both from the same specimens. The purpose of this paper is to critically compare experimental and calculated phase transition data and correlate with the microstructure and phase characterization data of as-cast and annealed U-Pu-Zr alloys. Phase transition peaks were measured using differential scanning calorimetry in the subsolidus regions (723-948 K) of three ternary U-Pu-Zr alloys with the same zirconium concentration but various U/Pu ratios. Overlapping peaks were deconvoluted using a Frazier-Suzuki peak fitting algorithm, and the critical peak temperatures and enthalpies were calculated. In general, increasing concentrations of Pu were associated with enhanced thermal stability of the body-centered cubic γ phase upon both heating and cooling. Experimental phase transition temperatures in this study tended to agree well with the predictions of the established ternary phase diagrams and other reported phase transition temperatures in literature. Additionally, the TAF-ID thermodynamic database was used to calculate a U-Pu-40 at.% Zr pseudobinary diagram as well as ternary diagrams from 773 to 973 K. The equilibrium phase transition temperatures tended to be considerably lower than measured peak temperatures upon both heating and cooling. Recommendations for improving the quality of data in future U-Pu-Zr characterization studies are also discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Physics-based model of irradiation creep for ferritic materials under fusion energy operation conditions

Irradiation creep is known to be an important process for structural materials in nuclear environments, potentially leading to creep failure at temperatures where thermal creep is generally negligible. While there is a great deal of data for irradiation creep in steels and zirconium alloys in light water reactor conditions, much less is known for first wall materials under fusion energy conditions. Lacking suitable fusion neutron sources for detailed experimentation, modeling, and simulation can help bridge the dose-rate and spectral-effects gap and produce quantifiable expectations for creep deformation of first wall materials under standard fusion conditions. In this paper, we develop a comprehensive model for irradiation creep created from merging a crystal plasticity representation of the dislocation microstructure and a defect evolution simulator that accounts for the entire cluster dimensionality space. Both approaches are linked by way of a climb velocity that captures dislocation-biased defect absorption and a dislocation strengthening term that reflects the accumulation of defect clusters in the system. Here, we carry out our study in Fe under first wall fusion reactor conditions, characterized by a fusion neutron spectrum with average recoil energies of 20 keV and a damage dose rate of ≈3×10 –7 dpa/s at temperatures between 300 and 800 K.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Consideration of Thermal Embrittlement in Alloy 316H for Advanced Non-Light Water Reactor Applications

To construct advanced non-light water reactors (ANLWRs) operating in the temperature range above that for the traditional light water reactors (LWRs), Alloy 316H is one of the candidate materials because of its inexpensiveness, significant service experience, and qualification for nuclear applications by the American Society of Mechanical Engineers (ASME). However, during the life span at temperatures expected for the ANLWRs, the alloy is likely to experience thermal embrittlement that has not been a concern for the traditional LWRs. To prepare for the development, the possibility of adverse thermal embrittlement effects on Alloy 316H performance in the ANLWRs must be evaluated and a technical basis regarding thermal embrittlement, if necessary, must be established for structural integrity analysis to provide reasonable assurance of adequate nuclear safety protection.In this paper, current technical basis for nuclear applications of Alloy 316H deterioration from thermal aging is briefly introduced. The likelihood of adverse thermal embrittlement effects on Alloy 316H performance is evaluated through historical data on microstructural and mechanical property evolution. Characterization of thermal embrittlement is then discussed, followed by a review of predictive models and trend curves for alloy embrittlement. Based on the review and evaluation, technical gaps for addressing thermal embrittlement issues are identified and gap-filling actions are recommended for establishing a technical basis to enable adequate consideration of thermal embrittlement in Alloy 316H applications to the ANLWRs.

Ren, Weiju↗
DOE OSTI2017

Mechanism-based modeling of solute strengthening: application to thermal creep in Zr alloy

This report focuses on the development of a physics-based thermal creep model aiming to predict the behavior of Zr alloy under reactor accident condition. The current models used for this kind of simulations are mostly empirical in nature, based generally on fits to the experimental steady-state creep rates under different temperature and stress conditions, which has the following limitations. First, reactor accident conditions, such as RIA and LOCA, usually take place in short times and involve only the primary, not the steady-state creep behavior stage. Moreover, the empirical models cannot cover the conditions from normal operation to accident environments. For example, Kombaiah and Murty [1,2] recently reported a transition between the low (n~4) and high (n~9) power law creep regimes in Zr alloys depending on the applied stress. Capturing such a behavior requires an accurate description of the mechanisms involved in the process. Therefore, a mechanism-based model that accounts for the evolution with time of microstructure is more appropriate and reliable for this kind of simulation.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Enhancements to Engineering-scale Reactor Pressure Vessel Fracture Capabilities in Grizzly

The Grizzly code is being developed to model the effect of aging in nuclear power plant systems, components, and structures. A significant part of this effort has been to develop capabilites to model the effects of embrittlement in reactor pressure vessels on their integrity. This includes both modeling of microstructure and engineering property evolution, and engineering-scale probabilistic fracture mechanics analysis. This report documents recent advances to the engineering-scale fracture mechanics capability for evaluation of RPV integrity under transient loading in Grizzly. These developments are in three areas: probabilistic fracture mechanics, general reduced order models for fracture a flaw locations, and improvments to the XFEM capability used in Grizzly for fracture mechanics analysis. The combination of these developments brings Grizzly closer to a state where it can be applied in production as a general tool for engineering-scale fracture mechanics analysis. The probabilistic fracture mechanics capabilities are included in the 1.5 testing version of Grizzly, and will be further refined in preparation for production use in the 2.0 released planned for fiscal year 2018.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Post-Irradiation Examination of Low Neutron Dose Metal Fuel Samples

The FCT irradiation series at the High Flux Isotope Reactor (HFIR) using the Hydraulic Tube Facility (HT) was initiated in FY-2012 with the goal of performing irradiation testing on advanced fuels and materials. Data from the FCT irradiation series was to be used to further the fundamental scientific understanding of phenomena involved in the evolution of metal fuels microstructure under irradiation based on separate effects testing of factors including elemental composition, temperature, and neutron fluence. This data would then be used to inform theoretical models for in-service performance of metallic fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Post-Irradiation Examination of Low Neutron Dose Metal Fuel Parallelepipeds

Metal fuel samples were subjected to series of low-dose irradiations at the High Flux Isotope Reactor (HFIR) Hydraulic Tube (HT) Facility in FY 2012. The objective of these irradiations was to further fundamental scientific understanding of phenomena involved in the evolution of metal fuels microstructures under irradiation based on separate effects testing of various factors including elemental composition, temperature, and neutron fluence. These data would then be used to inform theoretical models for in-service performance of metallic fuels. Because metallic fuels are generally considered for fast spectrum reactor designs, specialized irradiation capsules, referred to as “rabbits,” were designed with gadolinium thermal neutron shielding surrounding the samples. This design resulted in a fast-neutron spectrum comparable with the nominal in-reactor conditions for metal fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Irradiation creep model for α-Fe: Dislocation climb in the presence of stress fields (Final Report)

Irradiation creep is known to be an important process for structural materials in nuclear environments, potentially leading to creep failure at temperatures where thermal creep is generally negligible. While there is a great deal of data for irradiation creep in steels and zirconium alloys in light water reactor conditions, much less is known for first wall materials under fusion energy conditions. Lacking suitable fusion neutron sources for detailed experimentation, modeling, and simulation can help bridge the dose-rate and spectral-effects gap and produce quantifiable expectations for creep deformation of first wall materials under standard fusion conditions. In this work, we present a comprehensive model for irradiation creep created from merging a crystal plasticity representation of the dislocation microstructure and a defect evolution simulator that accounts for the entire cluster dimensionality space. Both approaches are linked by way of a climb velocity that captures dislocation-biased defect absorption and a dislocation strengthening term that reflects the accumulation of defect clusters in the system. We carry out our study in Fe under first wall fusion reactor conditions, characterized by a fusion neutron spectrum at temperatures between 300 and 800°K.

36 MATERIALS SCIENCE↗