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DOE OSTI2015

Quantitative Electron Nanodiffraction.

This Final report summarizes progress under this award for the final reporting period 2002 - 2013 in our development of quantitive electron nanodiffraction to materials problems, especially devoted to atomistic processes in semiconductors and electronic oxides such as the new artificial oxide multilayers, where our microdiffraction is complemented with energy-loss spectroscopy (ELNES) and aberration-corrected STEM imaging (9). The method has also been used to map out the chemical bonds in the important GaN semiconductor (1) used for solid state lighting, and to understand the effects of stacking sequence variations and interfaces in digital oxide superlattices (8). Other projects include the development of a laser-beam Zernike phase plate for cryo-electron microscopy (5) (based on the Kapitza-Dirac effect), work on reconstruction of molecular images using the scattering from many identical molecules lying in random orientations (4), a review article on space-group determination for the International Tables on Crystallography (10), the observation of energy-loss spectra with millivolt energy resolution and sub-nanometer spatial resolution from individual point defects in an alkali halide, a review article for the Centenary of X-ray Diffration (17) and the development of a new method of electron-beam lithography (12). We briefly summarize here the work on GaN, on oxide superlattice ELNES, and on lithography by STEM.

36 MATERIALS SCIENCE↗

DOE OSTI2015

DE-NE0000533 2015 final report

The objective of this work was to explore the development of advanced metal/ceramic composites with greatly improved radiation tolerance and stability above 500°C. Such composites combine the good properties of glasses (high strength and elastic limit, corrosion resistance) with those of crystals (high toughness, strain hardening). The ceramic component of the composite consisted of a high crystallization temperature amorphous material composed of Si-O-C while the metal component was Fe, chosen as a model material for steel. The potential impact of this work is the development of a new class of ceramic/metal composites that can be adapted for engineering applications, resulting in dramatically improved materials performance for advanced reactors. We observed that the combination of the composite constituents, as well as the interfaces between them, provided significantly enhanced radiation tolerance. Our research showed the SiOC materials to be extremely radiation tolerant up to 20 displacements per atom (dpa) at 600 °C. The composites of SiOC/Fe were also studies and found to be radiation tolerant up to 5 dpa at 300°C The approached used for the synthesis of these composites (physical vapor deposition) allowed for rapid alloy and composite prototype development and was therefore appropriate for the proposed exploratory study. It does not, however, limit the engineering applicability of the class of materials we have investigated because, in engineering practice, the amorphous ceramic can be made by a variety of chemical methods, including pyrolysis, and the amorphous-ceramic/steel composite can be manufactured by ball milling. The need to develop advanced cladding that does not react with hydrogen or other structural materials is urgent considering past accidents at Fukushima. Ceramic composites are therefore a strong option with stream reactions rates orders of magnitude lower than the Ziracloy currently in use. Different from its crystalline counterparts, the conventional concept of point defects does not apply to amorphous ceramics, so these materials are expect to have higher radiation tolerance. Furthermore, the interfaces created between amorphous-ceramics and metals represent a new type of defect sink that also improves the radiation resistance of the composite’s crystalline component. This project has shown that it is possible to develop high temperature irradiation resistant materials that are in critical need for nuclear applications under extreme conditions where in-core materials have to withstand neutron damage and high temperature.

36 MATERIALS SCIENCE↗

DOE OSTI2016

Characterization and Modeling of Grain Boundary Chemistry Evolution in Ferritic Steels under Irradiation

Ferritic/martensitic (FM) steels such as HT-9, T-91 and NF12 with chromium concentrations in the range of 9-12 at.% Cr and high Cr ferritic steels (oxide dispersion strengthened steels with 12-18% Cr) are receiving increasing attention for advanced nuclear applications, e.g. cladding and duct materials for sodium fast reactors, pressure vessels in Generation IV reactors and first wall structures in fusion reactors, thanks to their advantages over austenitic alloys. Predicting the behavior of these alloys under radiation is an essential step towards the use of these alloys. Several radiation-induced phenomena need to be taken into account, including phase separation, solute clustering, and radiation-induced segregation or depletion (RIS) to point defect sinks. RIS at grain boundaries has raised significant interest because of its role in irradiation assisted stress corrosion cracking (IASCC) and corrosion of structural materials. Numerous observations of RIS have been reported on austenitic stainless steels where it is generally found that Cr depletes at grain boundaries, consistently with Cr atoms being oversized in the fcc Fe matrix. While FM and ferritic steels are also subject to RIS at grain boundaries, unlike austenitic steels, the behavior of Cr is less clear with significant scatter and no clear dependency on irradiation condition or alloy type. In addition to the lack of conclusive experimental evidence regarding RIS in F-M alloys, there have been relatively few efforts at modeling RIS behavior in these alloys. The need for predictability of materials behavior and mitigation routes for IASCC requires elucidating the origin of the variable Cr behavior. A systematic detailed high-resolution structural and chemical characterization approach was applied to ion-implanted and neutron-irradiated model Fe-Cr alloys containing from 3 to 18 at.% Cr. Atom probe tomography analyses of the microstructures revealed slight Cr clustering and segregation to dislocations and grain boundaries in the ion-irradiated alloys. More significant segregation was observed in the neutron irradiated alloys. For the more concentrated alloys, irradiation did not affect existing Cr segregation to grain boundaries and segregation to dislocation loops was not observed perhaps due to a change in the dislocation loop structure with increasing Cr concentration. Precipitation of α’ was observed in the neutron irradiated alloys containing over 9 at.% Cr. However ion irradiation appears to suppress the precipitation process. Initial low dose ion irradiation experiments strongly suggest a cascade recoil effect. The systematic analysis of grain boundary orientation on Cr segregation was significantly challenged by carbon contamination during ion irradiation or by existing carbon and therefore carbide formation at grain boundaries (sensitization). The combination of the proposed systematic experimental approach with atomistic modeling of diffusion processes directly addresses the programmatic need for developing and benchmarking predictive models for material degradation taking into account atomistic kinetics parameters

36 MATERIALS SCIENCE↗

DOE OSTI2016

Report on simulation of fission gas and fission product diffusion in UO 2

In UO 2 nuclear fuel, the retention and release of fission gas atoms such as xenon (Xe) are important for nuclear fuel performance by, for example, reducing the fuel thermal conductivity, causing fuel swelling that leads to mechanical interaction with the clad, increasing the plenum pressure and reducing the fuel–clad gap thermal conductivity. We use multi-scale simulations to determine fission gas diffusion mechanisms as well as the corresponding rates in UO 2 under both intrinsic and irradiation conditions. In addition to Xe and Kr, the fission products Zr, Ru, Ce, Y, La, Sr and Ba have been investigated. Density functional theory (DFT) calculations are used to study formation, binding and migration energies of small clusters of Xe atoms and vacancies. Empirical potential calculations enable us to determine the corresponding entropies and attempt frequencies for migration as well as investigate the properties of large clusters or small fission gas bubbles. A continuum reaction-diffusion model is developed for Xe and point defects based on the mechanisms and rates obtained from atomistic simulations. Effective fission gas diffusivities are then obtained by solving this set of equations for different chemical and irradiation conditions using the MARMOT phase field code. The predictions are compared to available experimental data. The importance of the large Xe U3O cluster (a Xe atom in a uranium + oxygen vacancy trap site with two bound uranium vacancies) is emphasized, which is a consequence of its high mobility and high binding energy. We find that the Xe U3O cluster gives Xe diffusion coefficients that are higher for intrinsic conditions than under irradiation over a wide range of temperatures. Under irradiation the fast-moving Xe U3O cluster recombines quickly with irradiation-induced interstitial U ions, while this mechanism is less important for intrinsic conditions. The net result is higher concentration of the Xe U3O cluster for intrinsic conditions than under irradiation. We speculate that differences in the irradiation conditions and their impact on the Xe U3O cluster can explain the wide range of diffusivities reported in experimental studies. However, all vacancy-mediated mechanisms underestimate the Xe diffusivity compared to the empirical radiation-enhanced rate used in most fission gas release models. We investigate the possibility that diffusion of small fission gas bubbles or extended Xe-vacancy clusters may give rise to the observed radiation-enhanced diffusion coefficient. These studies highlight the importance of U divacancies and an octahedron coordination of uranium vacancies encompassing a Xe fission gas atom. The latter cluster can migrate via a multistep mechanism with a rather low effective barrier, which together with irradiation-induced clusters of uranium vacancies, gives rise to the irradiation-enhanced diffusion coefficient observed in experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Generalized Rate Theory for Void and Bubble Swelling and its Application to Plutonium Metal Alloys

In the classical rate theory for void swelling, vacancies and self-interstitials are produced by radiation in equal numbers, and in addition, thermal vacancies are also generated at the sinks, primarily at edge dislocations, at voids, and at grain boundaries. In contrast, due to the high formation energy of self-interstitials for normal metals and alloys, their thermal generation is negligible, as pointed out by Bullough and Perrin. However, recent DFT calculations of the formation energy of self-interstitial atoms in bcc metals have revealed that the sum of formation and migration energies for self-interstitials atoms (SIA) is of the same order of magnitude as for vacancies. The ratio of the activation energies for thermal generation of SIA and vacancies is presented. For fcc metals, this ratio is around three, but for bcc metals it is around 1.5. Reviewing theoretical predictions of point defect properties in δ-Pu, this ratio could possibly be less than one. As a result, thermal generation of SIA in bcc metals and in plutonium must be taken into considerations when modeling the growth of voids and of helium bubbles, and the classical rate theory (CRT) for void and bubble swelling must be extended to a generalized rate theory (GRT).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Study of High Fluence Radiation-induced Swelling and Hardening under Light Water Reactor Conditions

This report documents a comprehensive model that has been developed to enable simulations of microstructural evolution under the irradiation conditions typical of light water reactor (LWR) internal components. The model, which accounts cascade production of point defects and vacancy, interstitial faulted dislocation loops, interstitial clusters migrating one-dimensionally and the evolution of the network dislocation structure, has been parameterized to account damage accumulation in austenitic stainless steels. Nucleation and growth of an ensemble of cavities is based on accounting the residual and produced by irradiation He atoms and existence of the dislocation and production biases. Additional applications and potential future developments for the model are also discussed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Multi-modal STEM-based tomography of HT-9 irradiated in FFTF

Under irradiation, point defects and defect clusters can agglomerate to form extended two and three dimensional (2D/3D) defects. The formation of defects can be synergistic in nature with one defect or defect-type influencing the formation and/or evolution of another. The resul is a need exists to perform advanced characterization where microstructures are accurately reproduced in 3D. Here, HT-9 neutron irradiated in the FFTF was used to evaluate the ability of multi-tilt STEM-based tomography to reproduce the fine-scale radiation-induced microstructure. High-efficiency STEM-EDS was used to provide both structural and chemical information during the 3D reconstruction. The results show similar results to a previous two-tilt tomography study on the same material; the α' phase is denuded around the Ni-Si-Mn rich G-phase and cavities. It is concluded both tomography reconstruction techniques are readily viable and could add significant value to the advanced characterization capabilities for irradiated materials.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Coupling between interfacial charge and mechanical deformation at high temperatures in ceramics. Final report

Basic research, under this project, has led to new fundamental discoveries in the effect of electric field (and temperature) on diffusion kinetics, electroluminescence and electronic conductivity in ceramic materials. For example, yttria stabilized zirconia sinters in just seconds well below 1000°C, while, at the same time, exhibiting a highly non-linear rise in conductivity, and intense electroluminescence. The phenomenon has been shown to occur in nearly all ceramics whether they are ionic or electronic conductors, or insulators. The rise in conductivity leads to Joule heating but in-situ measurements with a platinum standard at APS and Brookhaven (NSLS-II) have shown that the temperatures are much below what would have been needed to sinter in a few seconds. We have explored the hypothesis that huge concentrations of Frenkel pairs of vacancies and interstitials are produced. While these point defects accelerate sintering (by the interstitials being transported to the pores and the vacancies to the grain boundaries), they may also produce electron-hole pairs which can impart high electrical conductivity or recombine to emit photons. Many measurements, including in-situ measurements of lattice expansion are consistent with this hypothesis. However, the energy barrier for generating Frenkels in zirconia for example is much larger than can be expected at temperature and electrical fields in these experiments. Most recent work suggests that short wavelength phonons play a critical role in the genesis of the “flash” phenomenon. For example, the Debye temperature has been shown to be the lower bound for initiating flash. Molecular dynamics simulations have shown that large concentrations of Frenkel pairs can form under the proliferation of “optical” phonons at the edge of the Brillouin zone, that is with a wavelength at or below the lattice parameter. This project has led to the development of new experimental techniques as well as a software hardware interface for process and control and data analysis in real time, with a time scale of less than one second. This approach spells a new paradigm in materials science research. It potential impact on high rate manufacturing is self evident.

36 MATERIALS SCIENCE↗

DOE OSTI2020

HPC4Mfg with Samsung: Making semiconductor devices cool through HPC ab initio simulations

For decades, the semiconductor technology has followed the Moore’s law, making newer devices more powerful and energy efficient. Recently, however, it has reached a point where the performance and the energy efficiency of the device do not improve with the shrinking device size. One of the fundamental reasons of this deviation from the past trend is the interconnect resistance, which becomes larger with the shrinking size. The devices size is so small that the quantum mechanical effects can no longer be ignored and the traditional continuum simulation tools such as TCAD become inadequate. In this project, Samsung Semiconductor Inc. and Lawrence Berkeley National Laboratory has collaborated to perform first of kind device-scale ab initio simulations to optimize materials and interconnect morphology to minimize interconnect resistance. We have tested the use of LS3DF method and the PEtot_trans approach on top of the folded spectrum method (FSM) Escan code to calculate the scattering state, and to study various effects influence the interconnect conductivity. We found that, the LS3DF can be used to calculate such metallic system. On the other hand, the use of Escan code to solve the linear equation is not practical due to the slow convergence. We have implemented a Chebyshev filter technique to calculate a few hundred eigen states near the scattering state energy E, then use these eigen states as preconditioner to solve the linear equation. We have used this approach to study the different factors which affect the interconnect conductivity, including the shape, the point defect, the temperature, and the grain boundary.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2020

Irradiation enhanced diffusion and diffusional creep in U₃Si₂

U₃Si₂ is a candidate as an advanced fuel for light water reactors (LWRs). Compared to UO₂, there are significant data gaps for the thermophysical and thermomechanical properties of U₃Si₂. Atomic scale simulations can help fill in those gaps, while also providing a more mechanistic understanding of the material properties. Creep is an important property for fuel performance as it influences the pellet-cladding gap, which in turn affects the temperature of the fuel (a key parameter in many other aspects of fuel behavior). In this work, we employed a DFT-informed cluster dynamics framework to predict point defect concentrations in U₃Si₂. The concentrations, diffusivities, and volumes of these defects were used to develop a diffusional creep model based on bulk (Nabarro-Herring) and grain boundary (Coble) diffusion processes. It was found that irradiation enhanced Nabarro-Herring creep is dominant at low temperatures, while Coble creep dominates for high temperatures. The model compares well against available experimental data and has been implemented in BISON. A demonstration case using simple power profiles has been carried out, showing that negligible creep occurs due to the low temperatures experienced by U₃Si₂, a consequence of its high thermal conductivity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Summary of Previous Mechanical Test Data on ODS Alloys 14YWT and OFRAC up to 1000ºC

The Nanostructured Ferritic Alloys (NFA) 14YWT and OFRAC were developed for future fission and fusion nuclear energy reactors requiring high-temperature mechanical properties that are tolerant to extreme neutron irradiation environments. The NFA contain a high concentration of Ti-, Y- and O-enriched nanoclusters (NC) and ultra-fine grains to achieve high temperature strength and creep properties and high sink strength for trapping irradiation induced point defects to minimize hardening and swelling and transmutated He atoms to form intragranular nano-size bubbles that prevent formation of coarse bubbles on grain boundaries that cause embrittlement. The mechanical properties of 14YWT and OFRAC have been acquired from tensile and creep tests conducted in the past at Oak Ridge National Laboratory. The development of 14YWT started in 2000, resulting in the production of numerous heats. Tensile properties were obtained from eleven heats of 14YWT from room temperature to 800ºC. Tensile data for the SM10 heat of 14YWT was extended to 1,000ºC. Initial development of OFRAC occurred in 2016. Tensile tests conducted from room temperature to 800ºC revealed similar properties of OFRAC with those obtained from the newer generation of 14YWT heats. The creep properties of 14YWT-SM10 evaluated using constant stress tensile and load time-to-failure tests at 800ºC showed low minimum creep rates with stresses of 200 and 100 MPa. The single time-to-failure test of 14YWT-SM10 at 800ºC and 100 MPa was terminated after 20,357 hours with no specimen failure and a low creep strain of ~0.24 %. The creep properties of OFRAC determined from strain-rate jump tests at were similar those of 14YWT-SM10. The stress exponent for 14YWT and OFRAC at 800ºC are similar and are consistent with threshold stress behavior. Since both 14YWT and OFRAC are candidates for fuel cladding in future fast reactors, several fabrication studies were recently conducted and have successfully demonstrated that thin wall tubes can be fabricated from 14YWT and OFRAC by cold pilger rolling and high precision tube rolling. The high temperature mechanical properties and feasibility of fabricating thin wall tubes make 14YWT and OFRAC candidates for application as heat pipes in advanced micro-reactors. The purpose of this report is to summarize the previously obtained tensile and creep data at temperatures up to 1000ºC for NFA 14YWT and OFRAC that have been acquired over the past 20 years at ORNL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

Sensing and arresting metal corrosion in molten chloride salts at 800 degrees Celsius

Originally this was planned as a 5 task, 3 year project, starting in May 2019. Task 1 Detecting oxidants and metal corrosion in molten chloride salt at 800°C. Task 2 Effectiveness of Zr getter using head gas sensors of inlet and outlet gases. Task 3 Cathodic protection of metal in molten salt at reducing potential. Task 4 Find reduction potential of metals in molten chloride salts & quantum modeling. Task 5 Develop corrosion test sections integrated into molten salt flow loop. After Q4, the project was scaled back and the reorganized SOPO was reduced to 3 tasks to be completed at the end of Q7 on 01/05/2021. The 3 SOPO tasks are: Task 1 Sense oxidants and the metal corrosion in molten Na-K-Mg-Cl salt at 800°C. Task 3 Cathodically protect H230 alloy in molten Na-K-Mg-Cl salt at 800C by making H230 alloy the cathode of a power supply with a chromium-rich-metal as the anode. Task 4 Find the reduction potential of Cr metal in molten Na-K-Mg-Cl salt at 800°C. The total timeline was reduced from 12 to 7 quarters with the end date being, January 5, 2021. This amounted to task 1,3 and 4 being done in Q5,6 and 7 of an extended year 1, ending on January 5, 2021. Task 1, 3 and 4 were completed. For task1) the increase of oxidants in molten salt is detected by two things: i) a positive shift of the electrode potential for a Haynes 230 alloy (H230) metal coupon versus a silver/silver-chloride reference electrode (SSE) and ii) an increase in the corrosion rate of a H230 electrode at 800C. For task 3) applying a cathodic potential (negative overpotential) to a H230 electrode versus SSE arrested the corrosion of the cathodically protected H230 compared to an unprotected “spectator H230 electrode” in the same molten chloride salt. Lastly, for task 4) the reduction potential of chrome and chrome rich, and nickel metal electrodes were measured. The most valuable outcome of this work is that a power supply can arrest metal corrosion for metal in molten chloride salt when the power supply applies a cathodic potential to bare metal, like H230, in molten chloride salts, like ternary Mg-K-Na-chloride salt. H230 is a high strength metal for containing molten chloride salt and ternary Mg-K-Na-chloride salt is used as heat transfer fluids in concentrating solar power plants. When to use cathodic protection for protecting metal in molten chloride salts? Two scenarios are envisioned: 1) for protecting high-valued components, like solar collectors and heat exchangers and 2) for protecting the bare surface in holidays (point defects) of corrosion inhibitors that are to intended to cover the entire surface of metal.

14 SOLAR ENERGY↗

DOE OSTI2024

STEM Imaging of Materials: Final Report for DE-FG02-08ER46547

This project supported research advancing the state of the art in scanning transmission electron microscopy (STEM) imaging and characterization of materials, especially electronic materials. The project began with studies of point defects in materials using high-resolution Z-contrast STEM, then evolved to including methods in four-dimensional (4D) STEM and machine learning. Highlights include: (1) the discovery of stable, p-type ZnO via Sb doping, (2) demonstration that the high performance of InGaN LEDs is not due to In composition fluctuations as previously proposed, (3) the demonstration of the first sub-pm precision high-resolution STEM images using non-rigid registration of a series of STEM images, (4) the extension of high-precision STEM to low dose imaging and spectrum imaging, (5) discovery of a synthesis method for twisted spiral growth of transition-metal dichalcogenide materials via chemical vapor deposition, and (6) development of machine-learning methods for analysis and denoising of enormous 4D STEM data sets.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Aging Mechanisms of Broad Area High-Power ~800nm Laser Diodes

Improved lifetime of high-power broad area laser diodes will enhance their utility for manufacturing, scientific, and medical laser systems. However, factors contributing to gradual degradation of diode laser performance are not fully understood. Here we study the aging mechanisms of ~800 nm aluminum-free laser diodes grown by solid-source MBE. Oxygen was intentionally introduced at three different concentration levels into the quantum well (QW) and waveguide (WG) at concentrations <5×10 15 cm -3 . Diode power, slope efficiency (SE th ), threshold current (I th ), and device parameters such as QW and WG lifetimes were measured at discrete time steps throughout the aging run. The data show that lasing characteristics prior to aging are degraded by oxygen introduction, but the gradual power degradation rate of ~5% over 500 hours at current density 1.74 kA/cm 2 does not change with oxygen concentration. Aging-induced changes in nonradiative lifetimes and threshold were observed for shorter aging times, but stabilized after ~100 hrs. and cannot explain the observed power degradation. This suggests that gradual degradation in these devices is not due to nonradiative recombination at low levels of point defects. The data shows that this degradation results primarily from changes in slope efficiency, and simulation of these results suggests that the power degradation arises from increased intracavity absorption (ICA). Efforts to slow ICA growth, instead of efforts to control low levels of nonradiative recombination centers, may be the most fruitful path forward to improve laser diode lifetime.

42 ENGINEERING↗

DOE OSTI2024

Exploring New Diamond Surfaces with Precision Chemistry and Quantum Spectroscopy

Point defects in diamond known as color centers are a promising platform for quantum sensing. As atom-like systems, they can exhibit excellent spin coherence and can be manipulated with light. As solid-state defects, they can be produced at high densities and be incorporated into scalable devices. Diamond is a uniquely excellent host; it has a large band gap, can be synthesized with sub-ppb impurity concentrations, and can be isotopically purified to eliminate magnetic noise from nuclear spins. Specifically, the nitrogen vacancy (NV) center has been demonstrated to be a highly sensitive, non-invasive magnetic probe capable of resolving the magnetic field of a single electron spin with nanometer spatial resolution. However, the development of NV centers and other color centers in diamond as a quantum platform crucially relies on developing methods to functionalize and process diamond surfaces. Diamond is the hardest material, making polishing and processing challenging, and diamond surfaces are inert and sterically hindered, making them notoriously difficult to chemically terminate and functionalize. Our approach is to discover new surface chemistries for diamond by combining multimodal surface spectroscopy with NV-based quantum spectroscopy to carefully characterize the diamond surface. We have developed a suite of techniques for characterizing the diamond surface using rapid tools for reaction screening and detailed surface spectroscopy tools for definitive identification of the surface bonding.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Preliminary results addressing material qualification using combined ion irradiation and modeling data

Additively-manufactured (AM) materials have attracted increasing attention in recent years as a new method to make novel and customized components. While AM and conventionally produced materials are compositionally similar, they do possess different microstructures, necessitating assessment of materials produced via AM for their behavior in reactor environments. Some microstructures unique to AM materials, such as compositional micro-inhomogeneity and dislocation cell structures, are of particular importance since they may lead to different radiation performance. The performance of AM materials for advanced nuclear reactor applications is of interest to the Advanced Materials and Manufacturing Technologies (AMMT) program under the Department of Energy Office of Nuclear Energy. The AMMT program aims to demonstrate its new accelerated development and qualification methods via laser powder bed fusion (LPBF) 316 stainless steel (SS). Focusing on material bearing both 316L and 316H specifications, we integrate ion irradiation and modeling. This year, we focus on answering foundational questions related to process variability, alloy chemistry variation, and microchemical segregation. Experimental results provide information and motivate questions to the modeling effort, which aims to develop the ability to model radiation-driven microstructural evolution in additively-manufactured 316 stainless steel under a variety of advanced reactor conditions, including different temperatures, neutron spectra, and fluxes, in a sort of "virtual experiment". We perform in-situ and ex-situ ion irradiations and microstructural characterizations to support the development of AM materials for reactor applications, develop a phase field model of radiation-induced segregation in additively manufactured material with high angle grain boundaries and dislocation cells, and investigate the effect of carbon and chromium content on point defect behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Nanostructured Alumina Forming Austenitic Alloy (NAFA) Production Using Mechanical Alloying and High-Temperature Consolidation

Alumina-forming austenitic (AFA) alloys are a promising class of nuclear materials because of their high-temperature oxidation/corrosion resistance and mechanical properties. Unfortunately, these alloys are limited in use for core material applications, and they are specifically limited for use as nuclear fuel cladding because of their high Ni-transmutation and helium generation rate in-service. Improving these alloys through a fine dispersion of oxide precipitates and thus increasing the effective irradiation sink strength of the alloy system may mitigate many of the degradation phenomena expected during alloy deployment. These phenomena include high-temperature helium embrittlement and cavity swelling for lead-cooled fast reactor applications. This work effort uses combination of conventional and advanced manufacturing approaches are being used to fabricate nanostructured AFA (NAFA) materials. As the first objective of this initiative, a conventional AFA was modified using mechanical alloying and extrusion to alter the precipitation characteristics to include a fine dispersion of nanoscale oxides intended to serve as traps for irradiation-induced point defects and transmuted He within the lattice. This analysis compared the efficacy of the conventional mechanical alloying and extrusion approach with the unalloyed AFA consolidated approach using hot isostatic pressing (HIP). It was found that, although the mechanical alloying approach is successful in producing a fine distribution of oxides within the first nanostructured AFA (NAFA-1), the distribution is heterogeneous because of the mild milling parameters used to prevent cold welding of powder to the spherical milling media. The additional dispersion of oxides, coupled with a higher volume fraction of other secondary phases in the NAFA-1, produces higher alloy strengths that range up to 600°C in comparison to the unalloyed HIP AFA, thus exceeding the operating temperature of lead-cooled fast reactors. However, the strength of the NAFA-1 is lower than that of the HIP AFA at 800°C, which is presumed to be a function of increased secondary phases from the nonoptimized AFA chemistry. Future work is planned on a newly procured NAFA-2 chemistry that is more suitable for advanced reactor applications exploring new manufacturing routes.

36 MATERIALS SCIENCE↗

DOE OSTI2024

Effects of irradiation temperature on the microstructure and deuterium retention in γ-LiAlO 2 pellets

This report presents the experimental findings obtained from November 2023 to September 2024. The study aims to investigate the effects of temperature on the microstructure, deuterium (D) retention, and lithium (Li) loss in γ-LiAlO 2 pellets subjected to sequential He + and D + ion irradiation to a high dose. Sequential ion irradiation has been a key method in our previous studies to simulate the behavior of γ-LiAlO 2 under neutron irradiation. In addition to dose and dose rate, irradiation temperature is a critical factor influencing microstructural and compositional changes. LiAl 5 O 8 precipitates have been observed in γ-LiAlO 2 pellets irradiated with reactor neutrons at ~300°C. These precipitates also form during ion irradiation at an elevated temperature of 500°C, but not at 300°C. In our ion irradiation experiments, the dose rate is typically three orders of magnitude higher than that of neutron irradiation, leading to a more rapid damage production. To better emulate the microstructural features in neutron-irradiated pellets using ion irradiation, a higher irradiation temperature is needed to accelerate the diffusion of point defects and enhance defect recovery rates, thereby compensating for the effects of the higher dose rate. The microstructural changes observed are the result of competing processes occurring during ion irradiation.

36 MATERIALS SCIENCE↗