Freeze-cast alumina pore networks: Effects of freezing conditions and dispersion medium
Not Available
Miller, S. M.↗
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Microstructure-refinement–driven enhanced tensile properties of high-pressure die-cast A380 alloy through friction stir processing
This work employs friction stir processing (FSP), a well-known severe plastic deformation technique, to selectively modify the microstructure of thin-walled, high-pressure die-cast (HPDC) aluminum alloy A380, a major HPDC alloy fabricated in the die casting sector. FSP effectively breaks down Al dendrites and acicular Si particles, creating a homogenized distribution of equiaxed Si particles in the aluminum matrix. After FSP, the refined Si particles (~1.5 µm) are smaller than the eutectic Si particles (3–8 µm) in HPDC condition. In addition, interparticle distance has decreased almost 50% compared to dendritic arm spacing, and FSP has reduced the aspect ratio of Si particles to ~2. Furthermore, FSP eliminates porosity, and breaks down needle-like second-phase Fe-Mn and Cu-rich particles, yielding a refined, homogeneous distribution. The FSP-induced microstructural refinement and porosity reduction improve bulk yield strength and ductility by 23% and 66%. Tensile properties are enhanced beyond those of the die skin of the HPDC plate, and the alloy possesses lower defect density and a highly refined microstructure. This study establishes the viability of FSP as a tool for microstructure modification and mechanical property improvement for HPDC Al alloys for the light-weighting goal of the automotive industries.
36 MATERIALS SCIENCE↗
Microstructure analysis and machinability of additively manufactured A205 aluminum with heat treatments
A205 aluminum is one of the few high-strength aluminum alloys discovered for additive manufacturing (AM). The microstructure and heat treatment effects of AM A205 vary from those of its cast form, indicating possible differences in machinability during post-processing. Here, the objective of this research is to determine these variations of AM A205 and their effects on machinability using the cast condition as the baseline. As built and two heat-treated conditions (solution treatment and age hardening) were applied and compared in terms of their microstructures, micro-hardness, and machinability including vibration, specific cutting energy, chip morphology, and surface finish. The effects of print orientations on machinability were tested as well and found to be insignificant. AM as built had a higher hardness and slightly better machinability than its cast as built cast counterpart due to a smaller grain size (1 µm vs. 15 µm), eutectic θ-Al 2 Cu lattice and a fine, homogeneous TiB 2 distribution. The solution treatment dissolved this θ-Al 2 Cu lattice and increased the grain size (3 µm) which decreased the hardness, increased the ductility seen in the chips, and reduced the overall machinability. The presence of θ’-Al 2 Cu precipitates improved the machinability of aged AM A205. In addition, it was found that A205 exhibited both ductile and brittle behavior depending on the cutting speed, thus producing different machinability. The speed dependency is related to the material condition and heat treatment.
36 MATERIALS SCIENCE↗
A novel method combining additive manufacturing and alloy infiltration for NdFeB bonded magnet fabrication
In this paper, binder jetting additive manufacturing technique is employed to fabricate NdFeB isotropic bonded magnets, followed by an infiltration process with low-melting point eutectic alloys [i.e., Nd 3 Cu 0.25 Co 0.75 (NdCuCo) and Pr 3 Cu 0.25 Co 0.75 (PrCuCo)]. Densification and mechanical strength improvement are achieved for the as-printed porous part. Meanwhile, the intrinsic coercivity H ci is enhanced from 732 to 1345 kA/m and 1233 kA/m after diffusion of NdCuCo and PrCuCo, respectively. This study presents a novel method for fabricating complex-shaped bonded magnets with promising mechanical and magnetic properties.
36 MATERIALS SCIENCE↗
Liquidus temperature and chemical durability of selected glasses to immobilize rare earth oxides waste
Not Available
Mohd Fadzil, Syazwani↗
Neutron-irradiation creep of silicon carbide materials beyond the initial transient
Irradiation creep beyond the transient regime was investigated for various silicon carbide (SiC) materials. Here, the materials examined included polycrystalline or monocrystalline high-purity SiC, nanopowder sintered SiC, highly crystalline and near-stoichiometric SiC fibers (including Hi-Nicalon Type S, Tyranno SA3, isotopically-controlled Sylramic and Sylramic-iBN fibers), and a Tyranno SA3 fiber–reinforced SiC matrix composite fabricated through a nano-infiltration transient eutectic phase process. Neutron irradiation experiments for bend stress relaxation tests were conducted at irradiation temperatures ranging from 430 to 1180 °C up to 30 dpa with initial bend stresses of up to ~1 GPa for the fibers and ~300 MPa for the other materials. Initial bend stress in the specimens continued to decrease from 1 to 30 dpa. Analysis revealed that (1) the stress exponent of irradiation creep above 1 dpa is approximately unity, (2) the stress normalized creep rate is ~1 × 10 –7 [dpa –1 MPa –1 ] at 430–750 °C for the range of 1–30 dpa for most polycrystalline SiC materials, and (3) the effects on irradiation creep of initial microstructures—such as grain boundary, crystal orientation, and secondary phases—increase with increasing irradiation temperature.
36 MATERIALS SCIENCE↗
Degradation of HT9 under simultaneous ion beam irradiation and liquid metal corrosion
A potentially promising coolant/structural material pair for a liquid-metal-cooled fast reactors is lead bismuth eutectic (LBE) coolant with the ferritic/martensitic steel HT9. The challenge of deploying LBE, however, is the corrosive environment it creates for structural materials. This corrosion can be mitigated with precise oxygen content control in the LBE to allow for the growth of passive protective oxide layers on the surface of the steel. In this paper, results are reported from the Irradiation Corrosion Experiment II (ICE-II), which allowed the simultaneous irradiation of a sample while in contact with LBE. It was found that a characteristic multilayer structure with an outer Fe 3 O 4 oxide and inner FeCr 2 O 4 spinel was grown and the oxidation was significantly larger in the irradiated region when compared to the region that was only exposed to LBE corrosion. As a result, a possible mechanisms are discussed to help understand this irradiation enhanced corrosion behavior.
36 MATERIALS SCIENCE↗
Evaluation of Pb-17Li compatibility of ODS Fe-12Cr-5Al alloys
The Dual Coolant Lead Lithium (DCLL: eutectic Pb–17Li and He) blanket concept requires improved Pb–17Li compatibility with ferritic steels in order to demonstrate acceptable performance in fusion reactors. As an initial step, static Pb-17at.%Li (Pb-17Li) capsule experiments were conducted on new oxide dispersion strengthened (ODS) FeCrAl alloys ((1) Y 2 O 3 (125Y), (2) Y 2 O 3 +ZrO 2 (125YZ), (3) Y 2 O 3 +HfO 2 (125YH), and (4) Y 2 O 3 +TiO 2 (125YT)) produced at ORNL via mechanical alloying (MA). Tests were conducted in static Pb–17Li for 1000 h at 700°C. Alloys showed promising compatibility with Pb–17Li with small mass change after testing for 125YZ, 125YH and 125YT, while the 125Y alloy experienced the highest mass loss associated with some oxide spallation and subsequent alloy dissolution. X-ray diffraction methods identified the surface reaction product as LiAlO 2 on all four alloys. A small decrease (~1 at.%) in Al content beneath the oxide scale was observed in all 4 ODS alloys, which extended through 60 μm beneath the oxide/metal interface. This indicates improvements in alloy dissolution by decreasing the amount of Al loss from the alloy. Scales formed on 125YZ, 125YH and 125YT were examined via scanning transmission electron microscopy (S/TEM) and revealed incorporation of Zr-, Hf-, and Ti-rich precipitates within the LiAlO2 product, respectively. This indicates an inward scale growth mechanism. Future work in flowing Pb–17Li is needed to further evaluate the effectiveness of this strategy in a test blanket module.
36 MATERIALS SCIENCE↗
Irradiation stability and thermo-mechanical properties of NITE-SiC irradiated to 10 dpa
In this study, five variants of nano-infiltration transient eutectic (NITE) SiC were prepared using nanopowder feedstock and sintering additive contents of <10 wt%. The dense monolithic materials were subsequently irradiated to 2 and 10 dpa in a mixed spectrum fission reactor at nominally 400 and 700°C. The evolution in swelling, strength, and thermal conductivity of these materials were examined after irradiation, where in all cases properties saturated at < 2dpa, without appreciable change for further irradiation to 10 dpa. Swelling behavior appeared similar to high-purity chemical vapor deposition (CVD) SiC within measurement uncertainty. The strength roughly doubled after irradiation. Finally, thermal resistivity increase as a result of irradiation was ~20% higher when compared to CVD-SiC.
36 MATERIALS SCIENCE↗
Analysis and modeling of the equilibrium behaviors of U and Pu in molten LiCl-KCl/Cd system at 500 °C
In the field of pyroprocessing of used nuclear fuels, liquid cadmium is applied in different ways to recover transuranics from molten salt, while effecting some degree of separations between transuranics and lanthanides. In this work, a model of the separation performance between uranium and plutonium across the phase boundary between molten salt and liquid cadmium is presented. The salt is the LiCl-KCl eutectic with solute concentrations of UCl 3 and PuCl 3 . The cadmium is the liquid cadmium phase field from the Cd-U-Pu ternary alloy system. Lastly, the model used to describe the behavior of uranium and plutonium in the salt/cadmium system is applied to describe the behavior of plutonium and lanthanide in the salt/cadmium system.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Uranium exchange kinetics in a molten LiCl-KCl/Cd system at 500 °C
A dissolution experiment is performed in order to examine the kinetics of uranium isotope equilibration in a molten salt and cadmium system. The salt contains a solute concentration of UCl 3 in LiCl-KCl eutectic and the cadmium contained dissolved uranium at 500 °C. A small piece of high enriched uranium is introduced to the cadmium, which perturbs the equilibrium with respect to the uranium enrichment of the salt and cadmium phases. The series of salt and cadmium samples trace how the enrichment of the two phases reach equilibrium (in the absence of a chemical driving force) over the course of approximately 24 h. Here, a model linking dissolution and isotope exchange kinetics is developed to track the evolution of uranium isotope composition. With the assumption of unaccounted salt mass incurred from the dated sample reporting practice, the model explains the trend of the measured uranium isotope composition well.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Thermal analysis of projected molten salt compositions during FFTF and EBR-II used nuclear fuel processing
This work studies the change in liquidus temperature of the salt used in the Mk-IV electrorefiner at the Idaho National Laboratory during processing campaigns of sodium-bonded driver fuels from Experimental Breeder Reactor II and the Fast Flux Test Facility reactor. Modeling and Simulation Tool for Electrochemical Recycling Systems (MASTERS), an INL proprietary pyroprocessing flowsheet simulation tool, was used to simulate the processing campaigns and determine the resulting composition changes of the Mark-IV electrorefiner salt. Surrogate salt samples simulating the Mk-IV electrorefiner salt during the fuel processing campaigns were prepared, and the thermal properties were measured via differential scanning calorimetry. Results from this study indicate that after processing approximately 2,150 kg of uranium metal from used fuel, the liquidus temperature of the molten salt exceeds 520°C, which is a significant increase from 352°C (the melting temperature of eutectic LiCl-KCl). It was also discovered that the liquidus temperature is strongly dependent on the amount of NaCl accumulating in the Mk-IV electrorefiner salt. Thus, the assessment of the liquidus temperature during the subsequent fuel processing campaigns can be simplified by considering the ternary LiCl-KCl-NaCl system only.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Compatibility of FeCrAlMo with flowing PbLi at 500°-650°C
To explore the upper temperature limit of a fusion blanket utilizing eutectic Pb-17at.%Li for tritium breeding, a series of three mono-metallic thermal convection loops have been fabricated from alloy APMT (Fe–21Cr–5Al–3Mo) and operated with commercial PbLi and increasing peak temperatures of 550°, 600° and 650 °C, flow rates of 0.4–0.7 cm/s and temperature gradients of 85–115 °C. Chains of APMT specimens including 25 mm long tensile specimens were hung in the hot and cold legs of the loops to evaluate post-exposure room temperature tensile properties after 1000 h exposures in each experiment. For specimens exposed to flowing PbLi above 500°C, only minor changes in tensile properties were observed after exposure. Most APMT specimens were pre-oxidized to form alumina prior to exposure and exhibited small mass losses as the alumina was partially transformed to α-LiAlO 2 that tended to spall after exposure. Specimens without pre-oxidation exhibited higher mass losses and formed γ-LiAlO 2 but with less spallation. While promising for increasing blanket temperatures, the small mass losses suggest that the surface oxide is not completely stable. In conclusion, even at the highest temperatures, there was no indication of classic dissolution/precipitation behavior suggesting that higher temperature compatibility may be possible.
36 MATERIALS SCIENCE↗
Effect of reduced dehalogenation on the performance of Y zeolite-based sintered waste forms
Not Available
Gardner, Levi (ORCID:0000000273277620)↗
Microstructural and crystallographic effects of sol-gel synthesized Ti-doped UO 2 sintered under reducing conditions
Titanium (Ti)-doped UO 2 microspheres of three different Ti concentrations (1000, 2000, and 4000 wppm) were synthesized using an internal gelation process. The microspheres were pressed into pellets, and a two-step heat treatment was applied to form monolithic cylindrical pellets with high densities (≥95%TD). Microstructure of these samples consisted of equiaxed grains with >300% increase in average grain size compared to the undoped UO 2 pellets. Secondary Ti-rich chemical phases corresponding to a liquid eutectic formed during sintering were observed at grain boundaries of UO 2 for samples doped with 4000 wppm Ti. Furthermore, these Ti-rich chemical phases were not observed in 1000 or 2000 wppm Ti samples at microscale using electron microscopy investigations. The 0.02–0.04% lower lattice parameter values for the Ti-doped UO 2 samples compared to the undoped UO 2 confirms the incorporation of Ti into the UO 2 lattice.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Effect of Li metal addition on corrosion control of Hastelloy N and stainless steel 316H in molten LiF-NaF-KF
Molten fluoride salts are used in various energy harvesting applications such as solar collectors and molten salt nuclear reactors (MSRs). Thermodynamically driven selective dissolution of alloying elements due to impurities present within the fluoride salts have been identified to be the main corrosion mechanism for structural alloys in molten fluoride salts. Impurities in these salts, during salt preparation as well as during operation, can be difficult to control. One way to prevent the selective dissolution of active alloying elements is to make the molten salt reducing by either purifying it to remove impurities or to add some reactive metal that will react with the impurities and make the salt reducing. Here this study is focused on the corrosion behaviour of Hastelloy N and 316H stainless steel in molten LiF-NaF-KF eutectic mixture (FLiNaK) at 700°C and how this behaviour changes with the addition of Li metal as a reducing impurity. Our results indicate that the active metal addition like Li can be very effective in controlling the redox behaviour of salt, which in-turn can mitigate corrosion. Even the addition of small quantities of Li metal can mitigate the selective dissolution of active alloying elements for the tested structural materials in FLiNaK. Our results also suggest that the presence of excess amounts of Li, above the minimum concentration required to react with the impurities in FLiNaK, does not have any detrimental effect in terms of formation of new intermetallic phases at the surface of the selected alloys under tested conditions.
36 MATERIALS SCIENCE↗
Characterization of the molten salt FMgNaK through ab initio molecular dynamics and experimental density measurements
The selection of a salt species for use in a molten salt reactor MSR is a key part of any MSR design. However, many salts have sparse or no thermophysical property data sets, especially those with higher melting points. One such salt is the eutectic mixture of (NaF)0.345(KF)0.59(MgF 2 )0.065, or FMgNaK, and is explored here through ab initio molecular dynamics (AIMD) simulation (1023-1273K) as well as experimental measurement of the liquid density using the Archimedean method (973-1223K). Predicted densities from AIMD are similar to experimentally-measured densities, though with more noise, suggesting that the small scale necessitated by AIMD simulations may be problematic for simulating FMgNaK in particular. Here the coefficient of thermal expansion is predicted from simulation, and salt structure is characterized. Mg-F-Mg chaining is observed in the salt network, though the low concentration of Mg inhibits chaining on the scale that is observed in (LiF) 0.67 (BeF 2 ) 0.33 , or FLiBe.
36 MATERIALS SCIENCE↗