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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 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Contaminant Migration Modeling to Support the In Situ Decommissioning of Hardened Nuclear Facilities - 20215

The decommissioning of hardened nuclear facilities provides a unique challenge in balancing current worker dose with exposure to future receptors and media. In situ decommissioning (ISD) is a cost-effective and safe option for the closure of such facilities but requires a detailed understanding of the threat that radioactive and hazardous constituents remaining in the facilities pose to groundwater and surrounding surface water. The Savannah River Site (SRS) extensively employs contaminant migration models within the decommissioning process to develop site-specific removal, grouting, and monitoring strategies in support of safe and effective final end states of nuclear facilities. This work discusses the use of contaminant migration modeling in the ISD process at SRS, including specific examples from the closure of P-Reactor and R-Reactor, Building 235-F, F-Canyon and F/H Laboratory Complex. Migration models developed during the closure of P-Reactor and R-Reactor were divided into four source areas: the reactor vessel, the process area, the disassembly basin, and the purification wing. Each source area was assigned a specific inventory and migration pathway and was then subject to varying hypothetical removal, capping, and grouting schemes. Modeling indicated that groundwater Maximum Concentration Limits (MCLs) might be exceeded if no action was taken for eleven and ten constituents of concern (COCs) at P-Reactor and R-Reactor, respectively. These exceedances could occur in as few as 200 to 500 years. Alternatively, the migration modeling demonstrated that the selected ISD actions reduced contaminant mobility which allowed for significant radioactive decay and resulted in fewer predicted exceedances of groundwater MCLs (five COCs for P-Reactor; eight COCs for R-Reactor). In response to the modeling results for P-Reactor and R-Reactor, effectiveness monitoring programs were developed to target contaminants, specific to each reactor, that may migrate to groundwater. Contaminant migration modeling also revealed that roof collapse was a large factor in the release of COCs to the environment, giving rise to roof improvements, and an inspection and vegetation control program to ensure roof stability over time. Contaminant migration modeling is also aiding in the closure planning for hardened facilities in F Area. Building 235-F housed the Actinide Billet Line, which produced Np-237 billets for irradiation in SRS reactors, and the Plutonium Fuel Form (PuFF) facility that produced Pu-238 heat sources for the space program. As a result of these missions, areas within Building 235-F contain considerable residual amounts of both Pu-238 and Np-237. Contaminant migration modeling of Building 235-F was originally performed in 2012 to identify the feasibility of ISD and the amount of radioactive material removal required to prevent the exceedance of groundwater MCLs. The original model indicated that a 60% reduction in the PuFF facility Pu-238 inventory could keep groundwater concentrations below standards, while Np-237 did not pose a threat to groundwater. However, updates to the model with an emphasis on source impact pathways revealed that, due to the orientation of the source areas relative to groundwater flow, no amount of reasonable removal of Pu-238 would keep groundwater concentrations below MCLs and that Np-237 could be a large contributor to localized MCL exceedances. With this insight, the refined 2019 model is being used to assist in the development of grouting plans specific to each facility source area, where bentonite may be utilized to slow the migration of Pu-238 and its daughter products from the PuFF facility and a reducing grout may decrease Np-237 transport by ensuring the nuclide remains in the less mobile +IV oxidation state. The beginning phases of contaminant migration modeling are underway for F-Canyon and associated facilities using lessons learned from the Reactors and Building 235-F. A contaminant migration pathway, similar to the pathway used for the reactor vessels, is being developed for the hot and warm canyons. The F-Canyon inventories are being spatially refined to identify specific and localized source zones, comparable to Building 235-F, that may impact groundwater. Contaminant migration modeling provides vital input within the ISD process, from facility investigation to post closure effectiveness monitoring, making it a valuable tool in the development of safe and effective end states for hardened nuclear facilities at SRS. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Response of 304L stainless steel and 6061-T651 aluminum alloy at -40 °C

The objective of this work is to extend the thermal-mechanical, elastic-plastic calibrations for 304L stainless steel [1] and and 6061-T651 aluminum alloy [2] to the regime between room temperature and -40 °C. The basis to extend the calibration consisted of new uniaxial tension tests conducted at -40 °C using the same plate material stocks, circular cylindrical specimen geometries and testing apparatus as previously, followed by attempts to fit power-law hardening functions to replicate the response observed in the specimens and then extend the yield, hardening constant, hardening exponent and rate constant functions in the calibrations to cover the new temperature regime.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Development of coherent-precipitate-hardened high-entropy alloys with hierarchical NiAl/Ni 2 TiAl precipitates in CrMnFeCoNiAl x Ti y alloys

Coherent precipitates hardening is currently emerging strengthening mechanism of the various high entropy alloys (HEAs). Recently, CrMnFeCoNiAl x HEAs have been studied to show a phase transition from face-centered-cubic (FCC) to body-centered-cubic (BCC) and formation of coherent precipitates (B2-NiAl) within the BCC matrix. The coherent precipitates in the CrMnFeCoNiAl x alloys could contribute to increase the strength but lead to considerable reduction of the ductility. The present work systematically investigated a series of CrMnFeCoNiAl 0.5 Ti x alloys to further improve the strength and plasticity, as compared to the previously reported CrMnFeCoNiAl x HEAs. As a result, an increase of Ti addition leads to the phase transition from FCC to BCC and formation of lamellar structure and hierarchical precipitates reinforced by B2-NiAl and L2 1 -Ni 2 TiAl phases. Excellent mechanical properties were achieved from CrMnFeCoNiAl 0.5 Ti 0.1 and CrMnFeCoNiAl 0.5 Ti 0.2 alloys. Herein, the mechanical properties of the CrMnFeCoNiAl 0.5 Ti x alloys were discussed via theoretical strengthening mechanisms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Constitutive modeling of δ-phase zircaloy hydride based on strain rate dependent nanoindentation and nano-scale impact dataset

In this work, a constitutive model for zircaloy-4 embedded with δ-phase hydride rim/blister structures (δ-phase zircaloy hydride) is presented. Nanoindentation and nano-scale impact experiments were performed on the δ-hydride blisters as a function of temperature in order to obtain strain rate dependent location specific constitutive response. A power law viscoplastic constitutive model was chosen to represent the strain rate dependent behavior. The strength coefficient and the strain hardening exponent in the power law relation were found to decrease with increase in temperature indicating softening behavior at higher temperatures. The strain rate exponent remained constant with increase in temperature indicating that the rate dependent ductility does not change significantly at higher temperatures. A finite element method based model was created to analyze the failure behavior and the ductile to brittle transition in zircaloy hydride. The model predicted that fracture strength of δ-phase zircaloy hydride decreases linearly with increase in temperature.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Origin of ultralow phonon transport and strong anharmonicity in lead-free halide perovskites

All-inorganic lead-free halide double perovskites offer a promising avenue toward non-toxic, stable optoelectronic materials, properties that are missing in their prominent lead-containing counterparts. Their large thermopowers and high carrier mobilities also make them promising for thermoelectric applications. Here, in this work, we present a first-principles study of the lattice vibrations and thermal transport behaviors of Cs 2 SnI 6 and γ-CsSnI 3 , two prototypical compounds in this materials class. We show that conventional static zero temperature density functional theory (DFT) calculations severely underestimate the lattice thermal conductivities (κ l ) of these compounds, indicating the importance of dynamical effects. By calculating anharmonic renormalized phonon dispersions, we show that some optic phonons significantly harden with increasing temperature (T), which reduces the scattering of heat carrying phonons and enhances calculated κ l values when compared with standard zero temperature DFT. Furthermore, we demonstrate that coherence contributions to κ l , arising from wave like phonon tunneling, are important in both compounds. Overall, calculated κ l with temperature-dependent interatomic force constants, built from particle and coherence contributions, are in good agreement with available measured data, for both magnitude and temperature dependence. Large anharmonicity combined with low phonon group velocities yield ultralow values, with room temperature values of 0.26 W/m-K and 0.72 W/m-K predicted for Cs 2 SnI 6 and γ-CsSnI 3 , respectively. We further show that the lattice dynamics of these compounds are highly anharmonic, largely mediated by rotation of the SnI6 octahedra and localized modes originating from Cs rattling motion. These thermal characteristics combined with their previously computed excellent electronic properties make these perovskites promising candidates for optoelectronic and room temperature thermoelectric applications.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Post Irradiation Examination of Pressurized Water Reactor Stainless Steel Internal Components

Internal structural components of pressurized water reactors (PWR), such as baffle former bolts, are subjected to significant neutron irradiation and mechanical stresses at elevated temperatures during plant operation. Over the long operation of the power plant, these conditions lead to potential degradation and reduced load-carrying capacity of these bolts. To understand property degradation more fully, to confirm the results of experimental irradiation programs, and to predict operational lifetime performance of structural materials in internal components, post irradiation examination of harvested materials from operating nuclear reactors is required. In this work, two high fluence 316 stainless steel baffle former bolts were retrieved from a commercial Westinghouse two-loop downflow type PWR and then sectioned for analysis via mechanical testing and microstructural characterization. The irradiation damage fluctuated along the bolt length with damage levels from 15 to 41 displacements per atom with the bolt head receiving approximately twice the neutron damage levels as that of the bolt thread section. Mechanical testing evaluations showed extensive irradiation hardening and a sharp decrease in fracture toughness in all parts of the bolt, though with limited variation along the bolt length. Microstructural characterization using analytical scanning transmission electron microscopy and atom probe tomography showed significant radiation induced precipitation, segregation, and dislocation loop and cavity formation. However, unlike the mechanical behavior, there was considerable variation along the bolt length but opposite of what may be expected, with more precipitation and cavity formation in the bolt thread where the neutron radiation dose is less than in the bolt head where the neutron radiation dose is higher. The cause of this variation is likely due to gradients in temperature, neutron energy spectra, and gamma irradiation.

Lach, Tim↗
DOE OSTI2023

Bridging microscale to macroscale mechanical property measurements of FeCrAl alloys by crystal plasticity modeling

FeCrAl alloys are candidates for accident tolerant fuel cladding of light water reactors. In this work, a microstructure- and temperature-dependent crystal plasticity model is employed to bridge microscale to macroscale mechanical property measurements of FeCrAl alloys. With the visco-plastic self-consistent (VPSC) polycrystal plasticity framework, a mechanism-based single crystal plasticity (MSCP) model adopts the Arrhenius type rate equation to describe the dependence of the critical resolved shear stress for dislocation slips on their temperature-dependent intrinsic frictional resistance and the microstructure-dependent irradiation hardening. The intrinsic frictional resistance associated with {110}<111> and {112}<111> slip systems were measured by in-situ micromechanical testing on unirradiated/irradiated samples at 25-500 °C. The irradiation hardening is estimated by the Bacon-Kocks-Scattergood (BKS) model with density and size of radiation-induced defects measured from microstructural characterization. Several features associated with thermo-mechanical behavior of unirradiated/irradiated polycrystalline FeCrAl alloys are captured. High density of deformation-induced dislocations and radiation-induced defects results in obvious hardening at room temperature, which is weakened at high temperature, and facilitates damage evolution during deformation. Moreover, both high temperature and radiation-induced defects, which facilitate dislocation multiplication, trigger large hardening rate. Finally, the proposed method together with application of accelerator-based ion irradiation technique is a surrogate approach to simulate neutron damage, improving the efficiency associated with evaluation of mechanical properties of FeCrAl alloys exposed to temperature, stress and radiation conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Ultra-high temperature testing and performance of L-PBF C103

Additive Manufacturing (AM) of refractory alloys is gaining traction as a materials processing route for components subject to extreme temperature environments. Due to the low oxidation resistance of refractory alloys, novel methods for evaluating their elevated temperature performance must be developed. In this work, a Gleeble® 3800 thermomechanical load frame was modified to evaluate the mechanical properties of laser powder bed fusion (L-PBF) consolidated niobium alloy C103 ranging from room temperature (RT) to 1400 °C. The fixturing and sample geometry were designed to accommodate Joule heating and prevent damage to the test chamber. Oxidation of the samples was minimized via testing in vacuum level of 1E-5 Torr. Ultimate tensile strength (UTS), yield strength (YS), elongation, and strain-hardening behavior were determined as a function of temperature. L-PBF C103 presented an average UTS of ∼650 MPa and over 25 % elongation at RT. Above RT, the UTS and YS dropped then leveled off from 500 °C to 1000 °C with values ranging from ∼400 MPa to ∼460 MPa, which is consistent with dynamic strain aging observed in this class of alloys. The strength rapidly declined after 1200 °C to ∼150 MPa at 1400 °C. Fractography indicated ductile fractures for the C103 at all test temperatures, and Electron Backscatter Diffraction (EBSD) analysis revealed a textured microstructure and the presence of dynamic recrystallization within the necked region of the sample tested at 1400 °C.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2020

Development of Cable Aging Acceptance Criteria for Nuclear Facilities

The aging of nuclear power plant electrical cables has been the subject of substantial research and development (R&D) projects in the nuclear industry, national and international laboratories, universities, and vendor organizations for several years. More specifically, these R&D efforts were conducted to develop equipment and techniques for assessing age-related degradation of in-service cables in nuclear facilities including power plants, research reactors, waste facilities and fuel fabrication plants. In recent years, cable R&D efforts in the nuclear arena have also included work to develop new cables for the next generation of reactors. Over time, exposure to harsh environmental conditions such as elevated temperatures, radiation, and humidity in nuclear installations can result in age-related degradation and failure of cables. In the current fleet of nuclear reactors, there are thousands of miles of cabling installed in each plant and many of these cables are exposed to these harsh environmental conditions. For these cables, the jacket and insulation polymers harden and become brittle over time, making them more susceptible to crack formation and growth, moisture intrusion, and other mechanisms that can lead to cable failure. Moreover, the existing U.S. fleet of 98 nuclear reactors has an average operating age of 38 years. Many of these nuclear power plants have applied for and almost all have been granted regulatory approval for license renewals to operate for 60 years, 20 years beyond their original 40-year life. Further, subsequent license renewals (SLRs) are underway with a few commercial nuclear power sites already approved to operate up to 80 years. As these reactors pursue operating life extensions, utilities must find a way to address issues associated with age-related degradation of cables. Designers of small modular reactors are also in need of technologies to evaluate the performance of cables that will be installed in harsher environments (e.g. higher temperatures, radiation doses, dose rates, etc.) than those present in current-generation reactors. Today, a variety of cable condition monitoring (CM) techniques have been developed and successfully used in nuclear facilities. These techniques are used to identify age-related degradation and assess the condition of cables to determine if their performance characteristics have changed with age. However, objective criteria must be developed for these CM tests to help quantify cable condition and thereby develop repair and replacement schedules. With operating life extending to 80 years and more SLRs on the horizon, both nuclear facilities and regulators need an objective means to determine the aged condition of cables.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Development of Cable Aging Acceptance Criteria for Nuclear Facilities

The aging of nuclear power plant electrical cables has been the subject of substantial research and development (R&D) projects in the nuclear industry, national and international laboratories, universities, and vendor organizations for several years. More specifically, these R&D efforts were conducted to develop equipment and techniques for assessing age-related degradation of in-service cables in nuclear facilities including power plants, research reactors, waste facilities and fuel fabrication plants. In recent years, cable R&D efforts in the nuclear arena have also included work to develop new cables for the next generation of reactors. Over time, exposure to harsh environmental conditions such as elevated temperatures, radiation, and humidity in nuclear installations can result in age-related degradation and failure of cables. In the current fleet of nuclear reactors, there are thousands of miles of cabling installed in each plant and many of these cables are exposed to these harsh environmental conditions. For these cables, the jacket and insulation polymers harden and become brittle over time, making them more susceptible to crack formation and growth, moisture intrusion, and other mechanisms that can lead to cable failure. Moreover, the existing U.S. fleet of 98 nuclear reactors has an average operating age of 38 years. Many of these nuclear power plants have applied for and almost all have been granted regulatory approval for license renewals to operate for 60 years, 20 years beyond their original 40-year life. Further, subsequent license renewals (SLRs) are underway with a few commercial nuclear power sites already approved to operate up to 80 years. As these reactors pursue operating life extensions, utilities must find a way to address issues associated with age-related degradation of cables. Designers of small modular reactors are also in need of technologies to evaluate the performance of cables that will be installed in harsher environments (e.g. higher temperatures, radiation doses, dose rates, etc.) than those present in current-generation reactors. Today, a variety of cable condition monitoring (CM) techniques have been developed and successfully used in nuclear facilities. These techniques are used to identify age-related degradation and assess the condition of cables to determine if their performance characteristics have changed with age. However, objective criteria must be developed for these CM tests to help quantify cable condition and thereby develop repair and replacement schedules. With operating life extending to 80 years and more SLRs on the horizon, both nuclear facilities and regulators need an objective means to determine the aged condition of cables.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Defect And Damage Characterization Of Additively Manufactured Titanium Alloy Ti-5553 Using Traditional Computed Tomography Volume Segmentation And Machine Learning Algorithms

The mechanical response of a component is affected by defects, such as porosity, arising from the laser powder bed fusion (LPBF) fabrication process. Thus, it is important to develop accurate and efficient inspection methods for identifying porosity. In this work, porosity identified in an X-ray computed tomography (XCT) volume of a Ti-5553 coupon was compared to pores identified in a serial sectioned volume that represented the ground truth. The porosity of the XCT scan was identified using contrast-based, ISO-based, and machine learning (ML) methods for segmentation. Large inherent porosity was easy to identify, but the ISO thresholding still struggled due to the intensity gradient resulting from both the beam hardening in XCT and the uneven lighting of the serial sectioning panels. Further, the results show that ML-based methods were better suited for identifying small pores and reducing the amount of false positives. Additionally, high strain-rate impact testing was done on some of the XCT samples as well as post-mortem XCT inspection, and the same suite of segmentation and quantification tools were used to identify the large spallation cavities. The comparison of porosity pre- and post-mortem provides insight on the influence of the LPBF porosity on the formation of spall cavities.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Development, validation, and verification of multi-pass thermo-mechanical welding simulations using the open-source MOOSE framework: NeT TG4 benchmark weldment

This study develops and validates a sequentially coupled thermo-mechanical welding simulation for the three-pass 316L stainless steel NeT TG4 benchmark weldment using the open-source Multiphysics Object-Oriented Simulation Environment (MOOSE) and the Nuclear Engineering Material model Library (NEML). A diffused ellipsoidal heat source was calibrated against thermocouple data and weld macrographs to accurately model the fusion zone geometry and transient thermal fields. Material hardening is represented using the Lemaitre-Chaboche mixed isotropic-kinematic hardening model, while four annealing models - no annealing, single-stage at 1050 °C and 1300 °C, and two-stage at 800 °C/1300 °C - were implemented to assess the impact of annealing models on the accuracy of the predicted welding-induced plasticity, distortions, and residual stresses. The predictions were validated against experimental measurements and benchmarked against results from commercial software, demonstrating that thermo-mechanical MOOSE welding simulations achieve comparable accuracy with enhanced computational efficiency. This work highlights the potential of using open-source finite element frameworks like MOOSE for advanced manufacturing simulations.

Ji, Wendy [Australian Nuclear Science and Technolo↗
DOE OSTI2020

Annual LDRD Report (FY2019)

The performance of Nuclear Deterrence (ND) systems could be improved by adopting advanced commercial CMOS technologies. However, the radiation hardness of these ND systems must be assured. This project quantified dose rate upset thresholds, allowing us to evaluate whether advanced commercial technologies can be used in rad hard applications. This project also evaluated the susceptibility of advanced commercial technologies to neutron displacement damage and single event effects, and developed hardness assurance methods. As a result of this work, Sandia is now collaborating more closely with DoD agencies and their contractors to understand and improve the radiation hardness of advanced commercial technologies. Sandia Capabilities (SPHINX, lon Beam Lab, FPGA test capabilities) were developed and are being utilized by other programs. Staff were trained to do radiation survivability testing and developed proposals that were funded. Internal and external collaborations and partnerships were developed, including with Georgia Tech. This project received the Best Paper award at the 2019 Hardened Electronics And Radiation Technology (HEART) Conference.

36 MATERIALS SCIENCE↗
DOE OSTI2021

A predictive discrete-continuum multiscale model of plasticity with quantified uncertainty

Multiscale models of materials, consisting of upscaling discrete simulations to continuum models, are unique in their capability to simulate complex materials behavior. The fundamental limitation in multiscale models is the presence of uncertainty in the computational predictions delivered by them. In this work, a sequential multiscale model has been developed, incorporating discrete dislocation dynamics (DDD) simulations and a strain gradient plasticity (SGP) model to predict the size effect in plastic deformations of metallic micro-pillars. The DDD simulations include uniaxial compression of micro-pillars with different sizes and over a wide range of initial dislocation densities and spatial distributions of dislocations. An SGP model is employed at the continuum level that accounts for the size-dependency of flow stress and hardening rate. Sequences of uncertainty analyses have been performed to assess the predictive capability of the multiscale model. The variance-based global sensitivity analysis determines the effect of parameter uncertainty on the SGP model prediction. The multiscale model is then constructed by calibrating the continuum model using the data furnished by the DDD simulations. A Bayesian calibration method is implemented to quantify the uncertainty due to microstructural randomness in discrete dislocation simulations (density and spatial distribution of dislocations) on the macroscopic continuum model prediction (size effect in plastic deformation). Here, the outcomes of this study indicate that the discrete-continuum multiscale model can accurately simulate the plastic deformation of micro-pillars, despite the significant uncertainty in the DDD results. Additionally, depending on the macroscopic features represented by the DDD simulations, the SGP model can reliably predict the size effect in plasticity responses of the micropillars with below 10% of error.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Development of Corrosion Resistant Coatings for Structural Materials for Liquid Fueled Molten Salts Reactors Applications (Final Report)

Current structural alloys, code certified for the MSRs temperature ranges, contain high levels of chromium, making them highly susceptible to molten halide salt corrosion. One potential solution to circumvent the need for code certification of novel alloys, which is expensive and time consuming, is to design claddings that will protect the underlying code certified materials from corrosion damage during operation. In this work, we examined the corrosion of Ni and Cu electroplated, Ni and CuNi weld overlay, Mo-laser clad, and carburized claddings on SS316H for use in molten salt reactor environments. After characterization of the cladded materials (Task 1), static corrosion tests were used to assess corrosion resistance of the claddings in typical molten fluoride salts (Task 2). The corrosion tests were performed in molten FLiNaK at 700°C up to 1000 hours. Pre- and post-corrosion Scanning Electron Microscopy (SEM), energy-dispersive-spectroscopy (EDS), Transmission Electron Microscopy (TEM), X-ray Diffraction (XRD) and glow-discharge-optical-emission-spectroscopy (GDOES) were performed on cladding cross-sections and surfaces to evaluate degradation. The Cu and Ni electroplated samples, as well as the carburized samples, showed excellent corrosion resistance relative to the bare SS316H. To assess high temperature cladding stability, ageing experiments were performed at temperatures up to 900 C in inert atmosphere for the electroplated and carburized samples and a diffusion model was developed to predict long term cladding behavior (Task 3). It was found that the Cu cladding was basically insensitive to the high temperature ageing, except for small secondary phases forming at the interface. On the other hand, the Ni electroplated cladding experienced significant interdiffusion. Nevertheless, the gain in corrosion resistance for a 100μm Ni electroplated cladding is phenomenal, with more than 50% reduction in chromium dissolution from the substrate material for the first 15 years of salt exposure at 700°C. To assess radiation resistance and phase stability, high-temperature 4MeV Ni heavy ion irradiation was performed across the cladding/substrate interface up to 50 displacement-per-atom (DPA) at 500°C and 700°C to assess the phase stability and irradiation behaviors of the electroplated systems (Task 4). The Ni and Cu electroplated systems did not experience void swelling at the contrary to the SS316H substrate due to their nanocrystalline nature. Some level of recrystallization was observed in the cladding, as well as radiation induced segregation and enhanced diffusion. The interface acted as a potent sink for point defects with the presence of a void denuded zone. The mechanical properties of the claddings were assessed using thermal shock, micro-indentation, nano-indentation, and four-point bend testing experiments (Task 5). These experiments were performed on the claddings in as-received, corroded, irradiated, and thermally aged states to determine the effects of typical molten salt reactor environments on cladding integrity. No significant mass loss was observed after repeated thermal shocks. While the electroplated samples softened after high-temperature ageing, irradiation hardening compensate this effect, such that there is little different with the as-received materials. The results of these experiments suggest that the electroplated (copper/nickel) show the greatest promise for application in molten salt reactor development. These claddings prevented any chromium dissolution from occurring during the static corrosion experiments. Additionally, they demonstrated favorable properties for high temperature diffusion, phase stability, interfacial mechanical properties, and irradiation resistance. Weld-overlay cladding are also of great interest and could reach properties similar to the electroplated samples upon optimization (especially with multiple weld passes). The carburized SS316H showed excellent behavior as well, but more work needs to be done to assess their long-term stability. Finally, the Mo-laser clad system was not pursued further due to the manufacturing challenges. It is also worth noting that the Ni cladding systems should behave relatively well in terms of weldability since the Ni-weld overlay microstructure and associated corrosion rate are sound. While the NiCu weld-overlay has even lower corrosion rates than the Ni-weld overlay, more studies on welding of Cu-electroplated systems are necessary since Cu clusters are known to embrittle steels.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Effects of chemical segregation on ductility-anisotropy in high strength Fe-Mn-Al-C lightweight austenitic steels

Ductile Fe-Mn-Al-C lightweight steels offer great potential as high specific strength materials for weight critical applications. In the present work, very high yield (~1 GPa) and ultimate tensile strength levels (~1.5 GPa) were obtained along the rolling and transverse directions of Fe-30Mn-8.5Al-0.9Si-0.9C-0.5Mo (wt.%), with high tensile elongation to failure of ~35%. This excellent mechanical behavior was attributed to the nanoscale, ordered κ-carbide precipitates formed during aging after rolling. However, tensile ductility was not isotropic through all directions of the rolled plate. In particular, the ductility was consistently lower than 5% true strain along the normal direction. Upon detailed microstructural investigations, micro-scale lamellar chemical micro-segregation bands were detected, which caused the observed anisotropic embrittlement along certain plate directions. These bands resulted in low failure strains along the normal direction since it is oriented perpendicular to these chemical segregation bands, resulting in a failure type reminiscent of delamination-induced cleavage fracture. Chemical segregation causes an inhomogeneous κ-carbide precipitate distribution upon aging, producing more pronounced mechanical anisotropy after peak-aging as compared to the solution heat treated samples. Surprisingly, deformation twinning was observed after tensile deformation of solution-heat treated specimens via electron-backscatter diffraction experiments, which is anomalous for a high stacking faulty energy (SFE) material. Further, it was proposed that chemical micro-segregation promotes locally different SFE regions, leading to deformation twinning in this nominally high SFE material. Overall, this work demonstrates the effects of chemical segregation on mechanical properties and microstructural evolution in a high performance, lightweight Fe-Mn-Al-C steel.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Strain engineering of the charge and spin-orbital interactions in Sr2IrO4

In the high spin–orbit-coupled Sr 2 IrO 4 , the high sensitivity of the ground state to the details of the local lattice structure shows a large potential for the manipulation of the functional properties by inducing local lattice distortions. We use epitaxial strain to modify the Ir–O bond geometry in Sr 2 IrO 4 and perform momentum-dependent resonant inelastic X-ray scattering (RIXS) at the metal and at the ligand sites to unveil the response of the low-energy elementary excitations. Furthermore, we observe that the pseudospin-wave dispersion for tensile-strained Sr 2 IrO 4 films displays large softening along the [h,0] direction, while along the [h,h] direction it shows hardening. This evolution reveals a renormalization of the magnetic interactions caused by a strain-driven cross-over from anisotropic to isotropic interactions between the magnetic moments. Moreover, we detect dispersive electron–hole pair excitations which shift to lower (higher) energies upon compressive (tensile) strain, manifesting a reduction (increase) in the size of the charge gap. This behavior shows an intimate coupling between charge excitations and lattice distortions in Sr 2 IrO 4 , originating from the modified hopping elements between the t 2g orbitals. Our work highlights the central role played by the lattice degrees of freedom in determining both the pseudospin and charge excitations of Sr 2 IrO 4 and provides valuable information toward the control of the ground state of complex oxides in the presence of high spin–orbit coupling.

36 MATERIALS SCIENCE↗