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A process for depositing elements by irradiating liquids is reported. Ultra pure elements are precipitated from aqueous solutions or suspensions of compounds. A solution of a salt of a metal to be prepared is irradiated, and the insoluble reaction product settles out. Some chemical compounds may also be prepared in this manner.
Grain growth studies of nanograined metals under both thermal and ion irradiation conditions has been studied extensively. Oxides, on the other hand, have not been studied as extensively, especially under irradiation. Some studies have focused on nanograined oxide grain growth on ZnO under thermal conditions only and on SnO 2 and CeO 2 using 250 keV Ar 2+ and 3 MeV Au + ions at 300 K, respectively. There does not appear to be any studies on grain growth of nanosized oxides which are commonly found to form under light water reactor conditions. Oxides such as magnetite (Fe 3 O 4 ), hematite (α-Fe 2 O 3 ), maghemite (γ-Fe 2 O 3 ), and wustite (FeO) are examples of common oxides which have not been studied. Here, this study focused on the grain growth of nanograined Fe 3 O 4 under both thermal and ion irradiation conditions in situ inside a Transmission Electron Microscope (TEM) to better understand the kinetics of grain growth of oxides.
Thermal conductivity is an important materials property related to heat transport, which is essential to many applications, ranging from thermoelectrics to nuclear reactor materials. High- quality thermal conductivity data is critical to these materials and their associated technologies, including the nuclear fuel materials like traditional oxide fuels and metallic uranium (U) fuels, e.g. U-Zr and U-Mo alloys. And thermal conductivity modeling is widely used to interpolate or extrapolate experimental data, to obtain high-quality thermal conductivity data over wide temperature and composition range, and to understand the impacts of different factors including defects and microstructures. However, a general thermal conductivity model for metal alloys which can work regrading different phase components is still missing. Also, for U alloys, a thermal conductivity model working with different phases is needed. Therefore, in this work, we developed thermal conductivity models for metal alloys, based on ab-initio calculations, semi-classical physics rules, and limited experimental data. The goals of these models are to help obtain high- quality thermal conductivity data within a little experimental input as possible, have models that are extendable to varying microstructures, including different types of irradiation effects, and provide mechanistic understanding of heat transfer in the modeled alloys. In this work, our model solves several challenges in the development. The DFT-BTE approach is applied to decrease the reliance on experimental data and make the model extendable to composition changes and defects. A practical and efficient way to combine the DFT inputs with physics rules is pointed out in this work too. A staged approach, which starts from simple cases of elemental metals, then extends to solid solutions, different compound phases, and multi-phase mixtures, is presented to work with the metal alloys in different phase components. Our models are demonstrated respectively on aU for elemental metal model, on U-Zr and U-Mo alloys in aU temperature range for multi-phase mixture model, on high-temperature U-Zr, U-Nb, and U-Mo alloys in the body-centered cubic phase for concentrated solid solution model, and on irradiated U-Mo alloys for irradiated metal alloy model. All models show great agreement with experimental data. In these demonstrations, our model shows its advantages compared to previous empirical fitting model. Our model requires fewer experimental data, due to the inputs from DFT. The quantitative insights into the different physical factors are provided in our model, as it incorporates the electron and phonon scattering mechanisms. Our model also can be extended to incorporate the contributions of point defects, grain boundaries, and noble gas bubbles, to integrate their effects on heat transfer. This model can serve as both a foundation for understanding the more complex thermal conductivity of realistic U alloy fuels and a useful tool to guide further modeling of thermal conductivity to aid materials and device design.
The Advanced Fuel Campaign’s Fission Accelerated Steady-state Testing (FAST) program uses metallic fuel pins with small diameters to reach a desired burnup more quickly. This enables accelerated testing of advanced fuel designs and decreases the time between idea conception and commercial usage. Post-irradiation examination is critical in this process, especially with respect to the fuel-cladding chemical interactions (FCCI). In this work, energy dispersive X-ray spectroscopy (EDS) is used to track how elements from the fuel and fission products have diffused through the cladding. Elemental redistribution along the fuel-cladding interface is mapped and FCCI region thicknesses are measured. The correlations between geometry, temperature, burnup, and FCCI thickness are presented.
Pulsed-laser irradiation causes the visible-near-infrared spectral slope of olivine (Fo90 and Fo99+) and SiO2 to increase (redden), while the olivine samples darken and the SiO2 samples brighten slightly. XPS analysis shows that irradiation of Fo90 produces metallic Fe. Analytical SEM and TEM measurements confirm that reddening in the Fo90 olivine samples correlates with the production of "nanophase" metallic Fe (npFe0) grains, 20-50 nm in size. The reddening observed in the SiO2 sample is consistent with the formation of SiO or other SiOx species that absorb in the visible. The weak spectral brightening induced by laser irradiation of SiO2 is consistent with a change in surface topography of the sample. The darkening observed in the olivine samples is likely caused by the formation of larger npFe0 particles, such as the 100-400 nm diameter npFe0 identified during our TEM analysis of Fo90 samples. The Fo90 reflectance spectra are qualitatively similar to those in previous experiments suggesting that in all cases formation of npFe0 is causing the spectral alteration. Finally, we find that the accumulation of successive laser pulses cause continued sample darkening in the Vis-NIR, which suggests that repeated surface impacts are an efficient way to darken airless body surfaces.
The United States Department of Energy is seeking to evaluate the feasibility of extended dry storage of aluminum-clad spent nuclear fuel (ASNF) in helium-backfilled cannisters. A significant research effort has been devoted to determining the amount of radiation-induced molecular hydrogen generation from corrosion layers that would be present on the fuel assembly surfaces. However, to date, no evidence of radiation-induced changes in the microstructure of the corroded aluminum surfaces have been reported. This report provides surface characterization, in terms of corrosion layer composition and morphology, for aluminum alloy 6061-T6 “plume” samples subjected to one of three potential ASNF drying techniques—“vacuum only”, “vacuum + 4 hours at 100 oC”, and “vacuum + 4 hours at 220 oC”—and gamma irradiated (= 53 MGy) in the presence of helium gas with ~0% added relative humidity. At these high absorbed gamma doses, for the first time ever, radiation-induced circular defects were found on the sample surface, regardless of the drying regime employed. Additionally, large cracks that penetrated through the corrosion layer to expose bare aluminum metal were observed in irradiated “plume” specimens subjected to drying conditions of “vacuum + 4 hours at 220 oC”. The nature and implications of these defects on the extended dry storage of ASNF is unclear, requiring further study.
Multi-principal component transition metal (TM) diborides represent a class of high-entropy ceramics (HECs) that have received considerable interest in recent years owing to their promising properties for extreme environment applications that include thermal/ environmental barriers, hypersonic vehicles, turbine engines, and next-generation nuclear reactors. While the addition of chemical disorder through the random distribution of TM elements on the cation sublattice has offered opportunities to tailor elastic stiffness and hardness, the effects of irradiationinduced structural damage on the physical properties of these complex materials have remained largely unexplored. To this end, changes in the hardness and elastic moduli of a high-entropy TM diboride (Hf0.2Nb0.2Ta0.2Ti0.2Zr0.2)B2 and three of its quaternary subsets following irradiation with 10 MeV gold (Au) ions to fluences of up to 6 × 10^15 Au cm-2 are investigated at the micrometer and sub-micrometer length-scales via the dispersion of laser-generated surface acoustic waves (SAW) and nanoindentation, respectively. The nanoindentation measurements show that the TM diborides exhibit an initial increase in hardness following irradiation with energetic Au ions, with a subsequent decrease in hardness following further irradiation. One quaternary composition, (Hf1/3Ta1/3Ti1/3)B2, exhibits a notable exception to the trend and continues to exhibit an increase in hardness with ion irradiation fluence. Although differences in the absolute values of the effective elastic moduli obtained from the measured SAW dispersion and nanoindentation are observed (and attributed to microstructural variations at the measurement length-scale), both techniques yield similar trends in the form of an initial reduction and subsequent saturation in the elastic modulus with increasing ion irradiation fluence. The quaternary TM diboride (Hf1/3Ta1/3Ti1/3)B2 again exhibits a departure from this trend. The high-entropy TM diboride (Hf0.2Nb0.2Ta0.2Ti0.2Zr0.2)B2 exhibits the greatest recovery in hardness and modulus when irradiated to high ion fluences following initial changes at low fluence, indicating superior resistance to radiation-induced damage over its quaternary counterparts. Opportunities for designing HECs with superior hardness and modulus for enhanced radiation resistance (compared to their single constituent counterparts) by tailoring chemical disorder and bond character in the lattice are discussed.
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As a result of years of operations of the Advanced Test Reactor (ATR) at the Idaho National Lab (INL), the ATR canal has become congested due to the storage of materials including waste consisting of highly irradiated remote handled (RH) metals. In order to support its continuing nuclear mission the ATR requires canal space to support fuel storage and future core internal change out (CIC) components. INL staff determined that the canal must be cleaned out and waste removed to efficiently utilize the valuable canal floor space. It was determined that a methodology for processing, packaging, characterizing, and removing current and future irradiated hardware from the ATR canal is needed. Orano Federal Services (OFS) and its subcontractor Babcock Services Inc. (BSI) were selected to provide remote tooling in order to safely and efficiently size reduce the material currently stored in the ATR canal. INL developed detailed requirements that established the design and performance requirements for the tooling. Orano TN and BSI had recently successfully completed similar tasks in clearing highly activated materials from spent fuel pools and canals at commercial nuclear plants. That experience directly translated into a proven approach that could be used at the ATR thereby saving time and money and lowering risks. Based on BSI's demonstrated proficiency in the design and deployment of remote size reduction specialty tooling at comparable commercial facilities, Orano Federal Services was confident that a solution to INL's requirements could be delivered. The tooling had to be designed to be compatible with the operational requirements of the ATR canal as well as the functionality needed by the facility operators that will use the equipment. Some of the major design requirements and functions of the equipment included remote operations, capability to shear highly irradiated sections of aluminum and steel tubing, pipe, and components ranging in size from 0.6 cm to 16.5 cm diameter into 1.5 m -1.8 m lengths, utilizing buoyancy compensation to assist with underwater operations, and use commercially available existing technology and components as much as practical. The design team evaluated and selected commercially available equipment that could be modified to meet the design and performance specifications. In addition, a waste sizing table (WST) was designed and fabricated to further ensure the size reduction equipment would perform as expected and give the facility operators a stable and usable remote, underwater work platform. After successful development of the tooling, a mock up test and operator training was conducted to ensure the tooling performed as expected and met the requirements of INL. (authors)
Conditions for aluminum surface blister formation from hydrogen introduced by proton irradiation
The Miniature Fuel Molybdenum-99 (Mini99) experiment is a first-of-a-kind experiment in the High Flux Isotope Reactor. Mini99 is the first MiniFuel irradiation experiment to use liquid metal inside the MiniFuel subcapsule for additional temperature control. In this work, the initial post-irradiation examination is reported. As-built simulations were conducted and compared with designs, and physical post-irradiation examinations were conducted to determine actual test temperatures via SiC thermometry samples, to evaluate the extent of fission gas release from the samples, and to remove the fuel from the subcapsules. As-built simulations compared well with experiment designs and were in reasonable agreement with the measured temperatures inferred from the passive SiC thermometers. Additionally, fission gas release values were relatively small (on the order of 3%); one sample reached approximately 8%. Although some of this behavior is attributable to recoil, most of the gas release was likely caused by the physics of the fuel behavior during the test. Finally, disassembly showed that at least four of the fuel samples degraded into small components and granules at some point during the test, with the state of the other two samples unproven. Additional work is planned for further examination of these samples via various microscopy techniques.
Metallic nuclear fuels have been proposed for use in several emerging nuclear reactor designs, and a number of codes have been developed to model these fuels and assess their performance. Qualification of these metallic nuclear fuels will ultimately require monitored irradiation of lead test assemblies, but the use of fuel performance codes can reduce the uncertainty associated with these efforts by quantifying uncertainties and estimating margins to failure ahead of time. In this work, metallic fuel performance code requirements are defined for the Versatile Test Reactor (VTR) project using input from the Experimental Breeder Reactor II functional requirements, operational requirements, and design criteria. This work focuses on the thermomechanical responses and irradiation behaviors of metallic fuel elements that are representative of the proposed VTR driver fuel design concepts. As such, many of the code requirements and physics/modeling discussions in this work are expected to be broadly applicable to metallic fuel applications outside of the VTR project.
A hybrid atomistic-continuum method can model the microstructure evolution of metals subjected to laser irradiation. This method combines classical molecular dynamics (MD) simulations with the two-temperature model (TTM) to account for the laser energy absorption and heat diffusion behavior. Accurate prediction of the temperature evolution in the combined MD-TTM method requires reliable accuracy in electron heat capacity, electron thermal conductivity, and electron–phonon coupling factor across the temperatures generated. This study uses the electronic density of states (DOS) obtained from first-principle calculations. The calculated electron temperature-dependent parameters are used in MD-TTM simulations to study the laser metal interactions in FCC and BCC metals and the phenomenon of laser shock loading and melting. This study uses FCC Al and BCC Ta as model systems to demonstrate this capability. When subjected to short pulsed laser shocks, the dynamic failure behavior predicted using temperature-dependent parameters is compared with the experimentally reported single-crystal and nanocrystalline Al and Ta systems. The MD-TTM simulations also investigate laser ablation and melting behavior of Ta to compare with the ablation threshold reported experimentally. In conclusion, this manuscript demonstrates that integrating the temperature-dependent parameters into MD-TTM simulations leads to the accurate modeling of the laser–metal interaction and allows the prediction of the kinetics of the solid–liquid interface.
The world is currently suffering socially, economically, and politically from the recent pandemic outbreak due to the Coronavirus Disease 2019 (COVID-19), and those in hospitals, schools, and elderly nursing homes face enhanced threats. Healthcare textiles, such as masks and medical staff gowns, are susceptible to contamination of various pathogenic microorganisms, including bacteria and viruses. Here, metal–organic frameworks (MOFs) can potentially address these challenges due to their tunable reactivity and ability to be incorporated as porous coatings on textile materials. Here, we report how incorporating titanium into the zirconium-pyrene-based MOF NU-1000, denoted as NU-1012, generates a highly reactive biocidal photocatalyst. This MOF features a rare ligand migration phenomenon, and both the Ti/Zr center and the pyrene linker act synergistically as dual active centers and widen the absorption band for this material that results in enhanced reactive oxygen species (ROS) generation upon visible light irradiation. Additionally, we found the ligand migration process is generally applicable to other csq topology Zr-MOFs. Importantly, NU-1012 can be easily incorporated onto cotton textile cloths as a coating, and the resulting composite material demonstrates fast and potent biocidal activity against Gram-negative bacteria (E. coli), Gram-positive bacteria (S. epidermidis), and T7 bacteriophage virus with up to a 7-log (99.99999%) reduction within one hour under simulated daylight.
Owing to their switchable spin states and dynamic electronic character, organic-based radical species have been invoked in phenomena unique to a variety of fields. When incorporated in solid state materials, generation of organic radicals proves challenging due to aggregation. Metal–organic frameworks (MOFs) are promising candidates for immobilization and stabilization of organic radicals because of the tunable spatial arrangement of organic linkers and metal nodes, which sequesters the reactive species. Herein, a flexible, redox-active tetracarboxylic acid linker bearing two imidazole units was chosen to construct a new Zr 6 -MOF, NU-910, with scu topology. By exploiting the structural flexibility of NU-910, we successfully modulate the dynamics between an isolated organic radical species and an organic radical π-dimer species in the MOF system. Single-crystal X-ray diffraction analysis reveals that through solvent exchange from N,N-diethylformamide to acetone, NU-910 undergoes a structural contraction with interlinker distances decreasing from 8.32 Å to 3.20 Å at 100 K. Organic radical species on the bridging linkers are generated via UV light irradiation. Direct observation of temperature-induced spin switches from an isolated radical species to a magnetically silent radical π-dimer in NU-910 after irradiation in the solid state was achieved via variable-temperature single-crystal X-ray diffraction and variable-temperature electron paramagnetic resonance spectroscopy. Ultraviolet–visible–near infrared spectroscopy and density functional theory calculations further substantiated the formation of a radical cation π-dimer upon irradiation. As a result, this work demonstrates the potential of using flexible MOFs as a platform to modulate radical spin states in the solid phase.
Tungsten has been chosen as the plasma-facing material (PFM) for the divertor region in ITER and also a candidate PFM for future plasma-burning nuclear fusion reactors. During fusion device operation, PFMs will be exposed to low-energy He irradiation at high temperatures, resulting in sub-surface bubbles and surface morphology changes such as pores and fuzz. Carbide dispersion-strengthened W materials may enhance the ductility of W, but their behavior under high flux He irradiation remains unclear. In this work, the response of dispersion-strengthened tungsten materials to high flux, low energy He irradiation at high temperature is examined. Tungsten alloyed with 1, 5, or 10 wt. % tantalum carbide or titanium carbide exposed to these conditions result in surface pores, coral-like feature growth and sub-surface helium bubbles. Reactor-relevant helium irradiation (5 x 10 26 m -2 fluence) combined with high powered laser pulses to simulate off-normal reactor events does not significantly alter the surface morphology, as the surface nanostructures appear stable and cracks are only observed on a localized region of one sample. However, specimens show the development of an impurity layer on the surface, likely impurity deposition from the sample holder during irradiation, resulting in a mixed material layer on the surface. Helium bubbles exist in this impurity layer, and obscure conclusions about helium interactions with the carbide dispersoids. Nonetheless, it is clear that the dispersoid microstructure limits He bubble formation and subsequent surface nanostructuring, attributed to the dispersoid composition.
The Enrico Fermi Atomic Power Plant Unit 1 (Fermi-1) was a sodium-cooled fast breeder reactor located in Monroe County, Michigan. The reactor was powered by a core of enriched uranium metal alloy driver fuel, which was enveloped by an axial and radial blanket material consisting of depleted uranium metal alloyed with 3 wt.% molybdenum. There are 406 axial and 559 radial irradiated sodium-bonded Fermi-1 blanket assemblies in storage at INL, totaling 34 metric tons of heavy metal. Disposal of the Fermi-1 blanket material directly into a geological repository is prohibited due to the reactive characteristic of its bond sodium. A melt drain evaporate (MEDE) process can effectively remove bond sodium from Fermi-1 blanket material. Consequently, planning is underway to apply a MEDE process to treat the 34 metric tons of heavy metal of irradiated Fermi-1 blanket material. Given the irradiated Fermi-1 blanket material’s relatively low power history and 50+ years of decay, it can be treated in a shielded glovebox. To assess the requisite shielding, the dose rates of the Fermi-1 blanket elements in various process configurations are needed. The method to perform the dose rate calculations is to first generate an average source term and associated photon source spectra for the Fermi-1 blanket material in SCALE and then use the associated photon spectra to calculate dose rates using MCNP6.2 in seven representative geometries. Dose rates in rem/h were calculated on contact (1 cm from outer geometry surface), 30 cm away from outer geometry surface, and 1 m from outer geometry surface at axial heights spanning the length of the blanket material within the geometry. The maximum average dose rate for the single Fermi-1 radial blanket element is ~0.60 rem/h, for the Fermi-1 radial blanket assembly ~1.10 rem/h, for the single Fermi-1 axial blanket element ~0.5 rem/h, for the Fermi-1 axial blanket assembly ~1.01 rem/h, for the MEDE can ~ 0.98 rem/h, for the Fermi Storage Canister ~ 0.83 rem/h, and for the MEDE cans in the DOE Standard Canister ~0.51 rem/h. Based on these maximum average values, the bounding dose rates are assumed to be 3× the average dose rates calculated for the axial blanket material and 7.6× the average dose rates calculated for the radial blanket material, based on the distribution of Fermi-1 blanket material assembly burnup.
MgK alpha X-rays (1254 eV) and 2 keV electrons irradiate the surface of polytetrafluoro ethylene (PTFE). The damage is confined to a few tenths of a micron below the surface, and the doses exceed 10 to the eight power rad. X-ray Photoelectron Spectroscopy (XPS) of the irradiated surfaces and mass spectroscopy of the gaseous products of irradiation indicate that the damaged layer is crosslinked or branched PTFE. After either type of irradiation, the surface has enhanced affinity for metals and a lower contact angle with hexadecane. Tape pull tests show that evaporated Ni and Au films adhere better to the irradiated surface. XPS shows the Ni interacts chemically with PTFE forming NiF2 and possibly NiC. However, the gold adhesion and contact angle results indicate that the interaction is, at least in part, chemically nonspecific. Decreased contact angles on FEP Teflon crystallized against gold were attributed to either the presence of a polar oxygen layer or increased physical forces due to greater density. In the case of irradiated PTFE, no oxygen on the surface was observed. The crosslinked structure might, however, have a greater density, thus accounting for the observed increase in adhesion and wettability.