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DOE OSTI2025

Mesoscale fractal whey protein particles derived from microscale linear-shaped protein assemblies (Part 1): Manufacturing method and particle characteristics

Whey protein isolates (WPI) are widely used in processed foods for their versatile functional properties. Modifying the structural properties of proteins by assembling them into mesoscale or microscale particles may improve their functionality and broaden their applications. This study aims to manufacture and characterize mesoscale whey protein particles (WPP) derived from WPI. Two types of WPP, WPP1 (0.05 mL/min) and WPP2 (0.25 mL/min), were prepared through a multistep approach involving liquid antisolvent (LAS) precipitation, heat treatment, and microfluidization. Liquid antisolvent precipitation was performed by injecting a 20% (wt/vol) WPI dispersion (pH 7) into an ethanol-glycerol mixture (75:25, vol/vol) under laminar flow, followed by heat treatment at 80°C for 20 min as a particle hardening step. This process produced stable fiber- and ribbon-shaped whey protein assemblies (WPA), which served as precursors to WPP. Subsequent microfluidization (150 MPa, 6 passages) reduced the size of WPA, yielding mesoscale WPP with irregular morphologies and a more uniform size distribution, as revealed by microscopy and dynamic light scattering. ζ-Potential and fluorescence labeling indicated higher surface charge and surface hydrophobicity of WPP compared with untreated WPI. The WPP showed internal mass fractal and surface fractal structures at larger length scales, analyzed using small-angle X-ray scattering. Fourier transform infrared spectroscopy demonstrated an increased fraction of intermolecular β-sheets in WPP, suggesting that hydrogen bonding contributed to their formation. Gel electrophoresis confirmed that disulfide bonds served as the primary cross-links stabilizing the WPP structure. Furthermore, turbidity measurements showed that WPP exhibited superior colloidal phase stability compared with untreated WPI and maintained high colloidal stability under both acidic and neutral pH conditions.

Antisolvent precipitation↗

DOE OSTI2026

Characterization of High-Entropy Alloys for Accelerator Beam Window Applications

Particle production targets and accelerator beam windows receive millions of high-energy/intensity beam pulses throughout their service lifetime. For next-generation multi-MW accelerators, it is necessary for beam windows to be able to withstand higher levels of thermal shock and radiation damage, while also having good high-temperature mechanical properties. Novel High-Entropy Alloys (HEAs) that form a single-phase microstructure, have nanoscale precipitates, and are made from low-Z elements are good candidates for beam window materials. This is due to their high-temperature strength and high resistance to radiation-induced swelling and hardening, as well as their beam transparency, allowing for maximized particle production and a longer service lifetime.

Kidwell, Tristan [U. Chicago (main)] (ORCID:000900↗

DOE OSTI2021

Emulation of neutron damage with proton irradiation and its effects on microstructure and microchemistry of Zircaloy-4

This work assesses the potential of proton irradiation to simulate the neutron damage to the matrix and laves phase Zr(Fe,Cr) 2 precipitates in Zircaloy-4. Isothermal proton irradiation has been performed on Zircaloy-4 samples at irradiation temperatures ranging from 250 to 350 °C. Two-step proton irradiation was also performed to enhance the amorphization of and iron loss from the laves phase Zr(Fe,Cr) 2 precipitates. The irradiated microstructures, including dislocation loops and rafts near SPPs, were observed in proton irradiated Zircaloy-4, which are consistent with neutron irradiated material at a similar damage level. The amount of irradiation-induced hardening after proton irradiation was similar to post neutron irradiated data. The significant amorphization of the SPPs and concurrent Fe redistribution observed on neutron irradiated materials can be effectively emulated using a two-step proton irradiation on Zircaloy-4. Hence, the neutron irradiation effect on Zircaloy-4 can be mostly captured using the two-step proton irradiation described in this study.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Utilizing a Dynamic Segmentation Convolutional Neural Network for Microstructure Analysis of Additively Manufactured Superalloy 718

Additive manufacturing (AM) is revolutionizing almost all industries through the production of intricate geometries previously prohibited by cost or machinability. Ni-based superalloys form a primary alloy class for high temperature applications in the petrochemical, aerospace, and nuclear industries because of their intrinsic resistance to creep and the ability to heat treat the superalloy for high strength. Despite these attractive properties, the extreme work hardening of Ni-based superalloys makes traditional manufacturing of complex shapes difficult and thus, these alloys are an attractive target for AM. Superalloy 718 was chosen as an example superalloy because of the wide variety of precipitates that can form within its composition space from the repetitive heating and cooling cycles of the AM process. The precipitates and other microstructure features, such as grain boundaries and dislocations, will dictate the mechanical properties and thus, there is an extensive challenge to characterize the size, number density, composition, and volume fraction of each microstructural feature from AM fabrication using analytical electron microscopy. This work focused on the application of a pixel-wise classification machine learning (ML) model called a dynamic segmentation convolutional neural network (DSCNN) to identify the microstructural features of an as-fabricated additively manufactured superalloy 718.

Taller, Stephen↗

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 OSTI2025

Development and Evaluation of Embrittlement Resistant Alloys for Advanced LWR Cladding

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Development and Evaluation of Embrittlement Resistant Alloys for Advanced LWR Cladding

The primary challenge preventing the deployment of reactor designs that leverage wrought FeCrAl as an advanced light-water reactor (LWR) cladding is irradiation hardening and embrittlement. Wrought FeCrAl alloys experience a loss of post-irradiation ductility and fracture toughness under low-temperature neutron irradiation (< 350°C) resulting from the combined effects of dislocation loop formation and the irradiation-enhanced precipitation of Cr-rich alpha-prime precipitates throughout the microstructure. Over the past decade, significant improvements in our understanding regarding the effect of Cr and Al content have been enabled through neutron irradiations, but even optimized wrought alloys such as C26M remain vulnerable to brittle failure during storage, transportation, and handling following irradiation. This report summarizes multiple irradiation campaigns initiated over the past 7 years and provides recent insights into the effect of (1) minor alloying composition, (2) alloy processing strategy, (3) crystal structure, and (4) the use of advanced oxide dispersion strengthened (ODS) alloys. The first irradiation campaign irradiated body-centered cubic FeCrAl alloy C26M with and without elements such as Mo and Y. It also included face-centered cubic alumina-forming austenitic (AFA) alloys that have a steeper neutronic penalty but that have far superior high-temperature strength than wrought FeCrAl. Finally, this campaign compared conventional wrought C26M with the same material produced using modern powder metallurgical hotisostatic pressing (PM-HIP). The second irradiation campaign, initiated in 2018, irradiated wrought and FeCrAl-ODS materials up to 50 dpa to assess their irradiation resistance to high-doses. The results of this work indicate that wrought C26M suffers significant degradation following neutron irradiation at LWR-relevant temperatures regardless of Mo and Y content. The PM-HIP variant appears to be more resilient to irradiation-induced ductility loss in comparison with wrought variants but still loses almost all ductility by 8 dpa. Notably, the AFAs retained at least 5% total elongation after irradiation at the same dose and temperature condition, although additional deterioration may be expected at higher fluence levels. FeCrAl-ODS materials show the greatest promise as a transformative longer-term accident-tolerant fuel (LT-ATF) cladding material. The extruded 106ZY10C alloy retained 10% total elongation after 16 dpa irradiation and retained 8% total elongation after 50 dpa. Additional challenges with FeCrAl-ODS alloys also remain, including optimization of end cap joining methodologies, enhancement of fracture toughness, scaling production to prove economic viability, and optimizing post-pilger heat treatments to maximize ductility and irradiation resistance. This series of irradiations demonstrates the efficacy of the LWR testbed available within the United States for the rapid irradiation and down-selection of LT-ATF candidates. As the country rapidly accelerates its timeline for the deployment of advanced reactor concepts, the effective utilization of the High Flux Isotope Reactor for separate-effects style irradiations should continue to be prioritized to answer the final questions pertaining to LT-ATF candidates necessary for the deployment of advanced boiling water reactors and small modular reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Small-scale mechanical testing and characterization of fuel cladding chemical interaction between HT9 cladding and advanced U-based metallic fuel alloy

Fuel cladding chemical interactions (FCCI) occurred on the interface between the nuclear metal fuel and cladding is the primary cause of cladding wastage, weakening cladding mechanical integrity, and placing fuel and cladding at risk. Although the microstructural and phase information of FCCI has been fairly understood, mechanical properties remain less studied due to limited reaction volume. Here, through a combining of advanced electron microscopy characterizations and small-scale mechanical testing techniques, including indentation and micro-tensile testing, this study investigated the microscale mechanical properties of FCCI between the ferritic/martensitic (F/M) HT9 cladding and an advanced Uranium (U)-based metallic fuel irradiated at the Advanced Test Reactor to 2.2% FIMA with peak inner cladding temperature reached to 650 °C. Mechanical testing results show significant hardening and embrittlement in the FCCI region. The brittle fracture of FCCI specimen is mainly attributed to the formation of nano-crystallized intermetallic σ-FeCr phase. Whereas mechanical softening was revealed in the unreacted HT9 matrix due to irradiation-induced microstructural and microchemical evolution, specifically, the disappearance of martensitic lath structure and the formation of Fe 2 Mo Laves phase precipitation which consumed the solid solution strengthening Mo from the F/M HT9 matrix. Due to the achieved high cladding temperature, this fuel pin is of particular significance for revealing the high-temperature irradiation effect on the mechanical properties of HT9 cladding. Therefore, the outcomes of this study are expected to contribute to the development of multi-scale mechanical behavior modeling of HT9 cladding for Generation IV reactors which requires cladding to run at higher temperature (above 600 ?).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Review of HAYNES® 282® alloy for supercritical CO2 structural applications

In this paper, HAYNES® 282® alloy (282® alloy) properties, microstructure, high temperature oxidation resistance, weldability, and American Society of Mechanical Engineers (ASME) code case highlights are reviewed. The long term performance of high temperature, high strength, creep resistant alloys is critical to the success of supercritical carbon dioxide equipment. Several modern power generation technologies require alloys that can operate continuously at or above 700oC. 282® alloy is a precipitation strengthened nickel based super alloy that meets these requirements. The results of 282® alloy research reviewed in this paper include longterm cyclic oxidation behavior in air and sCO2, independent research and validation of 100k hour creep life, and successful welding of the alloy in the age hardened condition. The latter is critical for joining of hardened components (flanges, piping, heat exchanger internal, etc.) and field repair. Several highlights from the code case for this ASMEapproved material are also presented.

Tossey, Brett↗

DOE OSTI2022

High strain rate deformation of aged TRIP Ti-10V-2Fe-3Al (wt.%) examined by in-situ synchrotron X-ray diffraction

Transformation Induced Plasticity (TRIP) is a promising avenue for tailoring the work hardening response of metastable $β$ titanium (Ti) alloys. Here we show that aged TRIP Ti-10V-2Fe-3Al (wt.%) maintains higher elongations and flow stresses as strain rate increases, if phase stability and microstructural characteristics are tuned. Low temperature aging influences the matrix $β$ phase stability by $ω$ phase precipitation, which affords a promising way to impact the TRIP effect and obtain desirable mechanical properties, ranging from high damping capacity to good strength/ductility combinations. Although TRIP is active during quasi-static and dynamic testing up to 2000 s –1 , increasing aging time and/or strain rate reduces the overall propensity for the TRIP effect and extent of transformation, which occurs rapidly just at the onset of yielding. Importantly, TRIP with $ω$ phase precipitation provides interesting alloying, microstructure, and property design strategies for engineering applications like lightweight protective structures, where high strains and the need for energy absorption are encountered.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Effect of heat treatment on the tensile behavior of selective laser melted Ti-6Al-4V by in situ X-ray characterization

Selective Laser Melted Ti-6Al-4V (as-SLMed) exhibits decreased yield strength, increased work hardening, and increased ductility after heat treatment at 730 degrees C (HT-730) or 900 degrees C (HT-900) for 2 h. To understand the change of mechanical properties, in situ high energy X-ray diffraction (HEXRD) is used to examine the phase composition, load partitioning, slip system activity, and dislocation density evolution in all three specimens. The as-SLMed specimen is dominated by martensitic α'. After heat treatment, α' partly or fully decomposes into α+β, reducing the yield strength. In HT-730, beta precipitates with confined size show much higher lattice strain than the α'/α matrix during deformation; in HT-900, the lattice strain difference is mostly eliminated. This is a key reason for the increased ductility in HT-900. From the anisotropic lattice strain development, basal slip is identified as the easiest slip system in α'/α. Using an elasto-plastic self-consistent (EPSC) model, the critical resolved shear stress ratio between prismatic slip and basal slip (CRSS prism /CRSS basal ) is estimated to be 1.31 and 1.16 in the as -SLMed and the HT-900 specimens, respectively. Further, the alpha phase in HT-900 is able to activate multiple slip systems and accumulate more dislocations during plastic deformation. This explains why HT-900 has better ductility and higher work hardening rate than the other two specimens.

36 MATERIALS SCIENCE↗

DOE OSTI2025

Microstructure and mechanical behavior of a TiC nanoprecipitate strengthened V Alloy

The V-4Cr-4Ti (V44) alloys have been proposed as the prime candidate structural material for self-cooled liquid Li blanket and other designs for fusion energy applications. However, the applications of the V44 alloy are limited to a narrow operation temperature window, due to reduction in creep strength at or above 700 °C and susceptibility to irradiation hardening and embrittlement when irradiated below 400 °C. Here, in this work, we explore the feasibility of designing a novel V alloy to form a high number density of TiC nanoprecipitates, in order to simultaneously improve creep strength and provide defect sinks to mitigate irradiation hardening. Computational thermodynamics was used to design a new alloy (V44C) to achieve our goal of high TiC nanoprecipitates density within the alloy V44 matrix. To ensure scalability, the new alloy was made through arc-melting and ingot-casting followed by hot forging, cold rolling and heat treatments of homogenization and precipitation aging. The microstructure was characterized by SEM, TEM, XRD and APT, confirming the existence of nanoprecipitates predicted in the thermodynamic calculations. In addition to microstructural evaluation tensile properties at room temperature and 700 °C, and Charpy impact energy at room temperature were measured. The microstructure and mechanical properties were then compared with those from a historic reference V44 alloy. The tensile strength improvement in V44C was rationalized based on particle and solid solution strengthening mechanism. The fracture behavior was discussed based on the fractography results and necking deformation behavior.

Alloys development↗

DOE OSTI2021

Deformation Mechanism of Age Hardened Alloy 709 Materials

This report provides an in-depth study of the microstructures of creep-deformed Alloy 790 materials. Five specimens, all creep-tested at 600°C under 330 MPa to failure, were investigated. The creep specimens were machined from plates fabricated by different melt practices, Argon Oxygen Decarburization (AOD), or AOD followed by Electroslag remelting (ESR), and different heat treatments: 1100 or 1150°C Solution Anneal (SA), and some with additional Precipitation Treatment (PT) of 775°C for 10 hours. TEM samples were made from the gauge and the grip sections of each creep-deformed specimen. Bright-field (BF) imaging, high-angle annular dark field (HAADF) imaging, and energy dispersive x-ray spectroscopy (EDS) using a scanning transmission electron microscope (STEM) were carried out on all the samples.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2022

Quantification of room temperature strengthening of laser shock peened Ni-based superalloy using synchrotron microdiffraction

Laser shock peening (LSP), a surface modification technique, is promising to enhance the strength and wear resistance for Ni-based superalloys. To understand the strengthening mechanism in a laser shock peened Ni-based superalloy DZ417G, we utilize synchrotron poly- and monochromatic X-ray microdiffraction, as well as electron microscopy and microhardness to quantify the local microstructures and mechanical properties at various depths. In the 1.2-mm-deep hardened layer, the microhardness increases monotonically by ~50% from the unaffected interior to the surface. Quantitative microdiffraction analysis shows that large amounts of dislocations are introduced by LSP. High densities of 7.1 × 10 15 m -2 and 11.8 × 10 15 m -2 are seen close to the peened surface for the γ- and γ'-phases, respectively, which are 5 and 20 times of those in the unaffected region. Different gradients of dislocation density are observed for the two phases from interior to surface, and their combined effect accounts well for the hardness increment. Due to the unaltered γ'-precipitates and chemical composition in the LSP affected zone, the large density of dislocations dominates the observed strengthening. Combined poly- and monochromatic X-ray microdiffraction allows quantifying the local microstructures of plastic deformation over a large sampling scale that can hardly be achieved using other materials characterization techniques.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Deconvoluting the Effect of Chromium and Aluminum on the Radiation Response of Wrought FeCrAl Alloys After Low-Dose Neutron Irradiation

FeCrAl alloys have been extensively investigated over the past decade as a candidate material for accident-tolerant fuel cladding in light water reactors. Here, we have completed a first of its kind study where Al and Cr concentrations are varied systematically under neutron irradiation to elucidate the post irradiation microstructural and mechanical response as a direct result of alloying content. Neutron irradiations were performed on alloys with composition Fe-(10-13)Cr-(5-7)Al in wt.% at temperatures of 214, 357, and 557°C to a dose of ~1.8 dpa (displacements per atom). Dislocation loop sizes and dispersion characteristics did not show strong compositional dependence. However, there were noticeable effects of composition on the dispersion of Cr-rich α'-precipitates, particularly a decrease in number density with increasing Al content or a decrease in Cr content. The present results confirm previous studies indicating the Cr concentration in these precipitates is lower than that expected in binary FeCr alloys and that Al can act as a destabilizing alloying element for the deleterious α'-phase.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Multimodal and multiscale strengthening mechanisms in Al-Ni-Zr-Ti-Mn alloy processed by laser powder bed fusion additive manufacturing

The unique thermokinetics of laser-powder bed fusion additive manufacturing (L-PBFAM) has been exploited for development of a novel high-strength Al-Ni-Ti-Zr-Mn alloy. The addition of 0.5 wt% Mn leads to extraordinary improvement in ultimate tensile strength (502 MPa) and work hardening due to the activation of two Mn-induced strengthening mechanisms. First, by a bimodal particle strengthening effect due to Al 31 Mn 6 Ni 12 nano-quasi-crystal particles rejected in inter-dendritic spaces and fibrous Al 3 Ni eutectic dendritic channels, which predominately contributes to the strength improvement, and second by solid solution strengthening from remaining Mn entrapped in Al. These mechanisms supplement the particle strengthening effect imparted by coherent and incoherent Al 3 (Ti,Zr) co-precipitates present at melt pool boundaries, dislocation strengthening due to solidification induced strain, and Hall-Petch and backstress strengthening effect due to heterogenous grain size distribution occurring at various length scales. In conclusion, the solidification dynamics and hierarchical heat distribution that are associated with L-PBFAM resulted in complex spatial variations in these strengthening phenomena and were investigated via a high-throughput multiscale structure–property correlation that involved thermokinetic simulation, X-ray diffraction, high-resolution nanoindentation mapping, and site-specific transmission electron microscopy of the alloy.

36 MATERIALS SCIENCE↗

DOE OSTI2025

Improved creep performance of CoCrFeNi high entropy alloys by Mo addition

The CoCrFeNi family of high entropy alloys has good mechanical and corrosion properties, performing like stainless steels with low yield stresses. They could, however, benefit from hardening methods for structural and high temperature applications. Here, this work investigates the potential for the CoCrFeNi alloy to be hardened with the addition of Mo. A series of CoCrFeNiMo alloys and their CoCrFeNi counterparts were subsequently manufactured at NETL and then mechanically creep tested at 650°C. The alloys were designed to possess a range of stacking fault energies which influence the mechanical properties of the alloys via the motion of their dislocations under stress. Results show that Mo-containing variants have creep rupture lives ranging from 4000 to 8000 hours, i.e., about two orders of magnitude longer than those for Mo-free variants. The significantly improved creep performance of the Mo-containing alloys is due to a combination of solid solution strengthening and precipitation strengthening from µ-phase and carbides.

36 MATERIALS SCIENCE↗

DOE OSTI2025

Conceptual Model Testing Related to SDU 6 Drainwell Observations

From its inception in the early 1950s through the end of the Cold War in the early 1990s, the Savannah River Site (SRS) produced nuclear materials for national defense in five reactors. Additionally, irradiated reactor fuel and target tubes were dissolved in nitric acid to recover plutonium and uranium using the PUREX (Plutonium Uranium Reduction EXtraction) process. Liquid waste from these chemical separations processes was then stored onsite in 51 underground tanks. Eight waste storage tanks have been operationally closed (i.e. cleaned and grouted) and the remaining tanks hold a mixture of liquids, insoluble solids, and precipitated salts (SRMC-LWP-2022-00001), the latter generated by evaporating water from the liquid waste. Waste is currently being retrieved from tanks and separated into 1) high-radioactivity, low-volume, and 2) low-radioactivity, high-volume components, principally through the Salt Waste Processing Facility (SWPF) (SRMC-LWP-2023-00001). The former waste stream is vitrified in the Defense Waste Processing Facility (DWPF), stored onsite, and destined for offsite disposal in a deep geologic repository. The latter stream is mixed with dry cementitious materials in the Saltstone Production Facility (SPF) and the wet slurry placed in onsite Saltstone Disposal Units (SDUs) within the Saltstone Disposal Facility (SDF), where it hardens into a cement waste form termed saltstone. A low-infiltration surface cover system will be placed over the SDF at closure, where SDUs will then be in the subsurface post-closure (SRR-CWDA-2019-00001).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗