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Polymeric dopant strategy distinctly suppressed the molecular aggregations and surface dewetting of the NPB small molecule hole transport layer in perovskite photovoltaics.
Altering the CaO to ZnO ratio of alkali borosilicate glass, in favour of ZnO, causes phase separation with hardystonite-like (Ca 2 ZnSi 2 O 7 ) and willemite-like (ZnSi 2 O 4 ) structures identified. This is concurrent with increased glass durability.
The chemical pathway for synthesizing covalent organic frameworks (COFs) involves a complex medley of reaction sequences over a rippling energy landscape that cannot be adequately described using existing theories. Even with the development of state-of-the-art experimental and computational tools, identifying primary mechanisms of nucleation and growth of COFs remains elusive. Other than empirically, little is known about how the catalyst composition and water activity affect the kinetics of the reaction pathway. In this work, for the first time, we employ time-resolved in situ Fourier transform infrared spectroscopy (FT-IR) coupled with a six-parameter microkinetic model consisting of ~10 million reactions and over 20 000 species. The integrated approach elucidates previously unrecognized roles of catalyst pK a on COF yield and water on growth rate and size distribution. COF crystalline yield increases with decreasing pK a of the catalysts, whereas the effect of water is to reduce the growth rate of COF and broaden the size distribution. The microkinetic model reproduces the experimental data and quantitatively predicts the role of synthesis conditions such as temperature, catalyst, and precursor concentration on the nucleation and growth rates. Furthermore, the model also validates the second-order reaction mechanism of COF-5 and predicts the activation barriers for classical and non-classical growth of COF-5 crystals. The microkinetic model developed here is generalizable to different COFs and other multicomponent systems.
During nuclear fuel reprocessing, radioiodine, can be released.
Pure manganese product is recovered from carbonaceous polymetallic nodules via pH cycling.
Annealing Mg-implanted homoepitaxial GaN at temperatures above 1400 °C eliminates the formation of inversion domains and leads to improved dopant activation efficiency. Extended defects, in the form of inversion domains, contain electrically inactive Mg after post-implantation annealing at temperatures as high as 1300 °C (one GPa N 2 overpressure), which results in a low dopant activation efficiency. Triple-axis x-ray data reveal that implant-induced strain is fully relieved after annealing at 1300 °C for 10 min, indicating that strain-inducing point defects formed during implantation have reconfigured and inversion domains are formed. However, annealing at temperatures of 1400–1500 °C (one GPa N 2 overpressure) eliminates the presence of the inversion domains. While residual defects, such as dislocation loops, still exist after annealing at and above 1400 °C, chemical analysis at multiple dislocation loops shows no sign of Mg segregation. Meanwhile, an overall decreasing trend in the dislocation loop density is observed after annealing at the higher temperatures and longer times. Additionally, once inversion domains are formed and the samples are cooled to room temperature, they are shown to dissolve with subsequent annealing above 1400 °C. While such defects have been observed before, the important finding that such defects can be dissolved with a short, higher temperature step is key. Earlier work [Breckenridge et al., J. Appl. Phys. Lett. 118, 022101 (2021)] addressing electrical measurements of these types of samples showed that annealing at 1400 °C leads to a dopant activation efficiency that is an order of magnitude higher than that observed at 1300 °C. In conclusion, this work complements earlier work by identifying the inversion domains, which incorporate Mg, and points to the benefits, in terms of defect density and p-type dopant activation, of using higher temperature (>1400 °C) annealing cycles to activate Mg in GaN, even if the Mg-containing inversion domains had been formed during lower temperature annealing.
There is considerable interest in molten chloride salts for several applications including thermal storage and next-generation molten salt reactors (MSRs). Several studies have concluded that Cl salts are highly corrosive to structural materials. Using Oak Ridge National Laboratory’s established methodology, Cl salt compatibility was assessed for candidate Ni-based alloys 230, 600 and 740 H at 600°–800°C in static welded capsules and in flowing thermal convection loop (TCL) salt experiments. Simply drying commercial Cl salt at 550°–650°C and adding~0.05 wt.%Mg was able to produce small specimen mass changes and limited surface attack after 100–1000 h exposures. Intentional additions of H 2 O, NiCl 2 and undried salt were used to better understand the role of impurities and achieve the>50 µm levels of attack reported in other studies. Further, characterisation of Cr depletion and oxide formation in conjunction with pre- and post-test salt chemistry measurements are being used to understand salt compatibility of structural materials.
Elevated uranium concentrations in groundwater remain a persistent challenge at contaminated sites. Several sites rely on natural flushing of uranium as a remediation strategy. Uncertainties in conceptual models can cause remediation strategies to underestimate timeframes required to reach remediation goals. In this study, laboratory experiments were conducted to investigate uranium flushing of sediments from a site with persistent groundwater contamination. Six columns were used to simulate alkalinity-enhanced flushing of uranium from sediments, by switching influent alkalinity after ~12 pore volumes. About 20% to 31% of sediment uranium was consistently flushed from the sediments. K d values varied greatly (5.3 to 117 L/kg) but were consistently lower (20% to 50%) during influents with elevated alkalinity. The mass of uranium recovered from the columns also was consistently higher (5% to 80%) during alkalinity-enhanced flushing periods. However, column experiments with sequentially increasing alkalinity showed diminishing returns in uranium elution at higher carbonate concentrations. The loss of flushing efficiency is attributed to significant calcite precipitation at higher alkalinities. Furthermore, the results show that alkalinity-enhanced flushing of uranium could be employed as a viable remediation scheme if calcite precipitation could be minimized in a field application.
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The U.S. Department of Energy’s National Nuclear Security Administration (NNSA) has entered into cooperative agreements with commercial businesses to establish a reliable domestic source of 99 Mo that will be produced without the use of highly enriched uranium (HEU).
The ability to produce transcurium elements and isotopes, such as 249 Bk and 252 Cf, requires the appropriate feed material and a high thermal neutron flux. Currently, at Oak Ridge National Laboratory (ORNL), the High Flux Isotope Reactor (HFIR) provides one of the world’s highest steady state thermal neutron fluxes. The feed material consists of recycled curium from the decades of production of 249 Bk, 252 Cf, 253 Es, and 255/257 Fm at the Radiochemical Engineering Development Center (REDC). Over time, the feed material has been depleted and needs to be replenished. The current feed material is called “heavy recycled curium” and consists of over 50 wt% of 246/248 Cm isotopes. In order to replenish this stock material, additional curium is needed. The only curium material that is currently available consists predominantly (greater than 80 wt%) of 244 Cm. This material also contains 240 Pu from the radioactive decay of the 244 Cm. Because of the amount of 244 Cm is greater than 50% of the total Cm, this material is termed “light curium.” Today, 249 Bk and 252 Cf are produced in only two places in the world: the United States and Russia. In the United States, the US Department of Energy and its predecessors have made these isotopes and others available to research agencies and commercial users since the late 1960s as part of the Department of Energy heavy-element production program located at ORNL. Currently these materials are provided through the DOE Office of Science Nuclear Physics-Isotope Development and Production for Research and Applications and the Basic Energy Science-Heavy Element Chemistry. Feedstock for the production of 252 Cf and other transcurium element by-product isotopes was originally prepared by means of irradiation of multi-kilograms of 239 Pu in the large production reactors at the Savannah River Site. Products of the Savannah River Site irradiation were predominantly 242 Pu, 243 Am, and 244 Cm. These materials were shipped to ORNL for production of multi-milligram amounts per year of 252 Cf and other transcurium element isotopes by means of irradiation in the HFIR, followed by separation and purification processing in the Radiochemical Engineering Development Center (REDC). The transcurium element isotopes produced were distributed to other US and collaborating foreign laboratories for research studies to determine the nuclear and chemical properties of these rare materials. The Cm feed material plays a key role in the efficient production of 252 Cf. The early targets irradiated in HFIR contained only the 242 Pu, 243 Am, and 244 Cm produced at Savannah River Site, and the rate of production of 252 Cf was less than 1 mg per 8 g of target material as shown in Figure 1. Also, shown in Figure 1 is the significant difference in 252 Cf production between the light and heavy curium; it differs by approximately 2 orders of magnitude.
Savannah River Mission Completion (SRMC) is currently preparing the second batch of material to be processed through the Tank Closure Cesium Removal (TCCR) 1A system. The feed for TCCR 1A consists of dissolved saltcake from Tank 9H. The third batch of salt to make up processing Batch 2 (Batch 2C) has recently been dissolved in Tank 9H and transferred to Tank 10H where it was composited with the first part of the batch (Batches 2A and 2B) in preparation for processing through the TCCR 1A unit. Savannah River National Laboratory (SRNL) received samples from the recent batch (2C) of dissolved salt for characterization.
For each sludge batch that is processed in the Defense Waste Processing Facility (DWPF), the Savannah River National Laboratory (SRNL) tests the applicability of the digestion methods used by the DWPF Laboratory for elemental analysis of Sludge Receipt and Adjustment Tank (SRAT) Receipt samples and SRAT Product process control samples. DWPF SRAT samples are typically dissolved using a method referred to as the DWPF Cold Chemical or Cold Chem Method (CC), (see DWPF Procedure SW4-15.201- 2.1).
Decontamination of 99 Tc, Nb, and Zr impurities in dissolved irradiated Mo disks can be accomplished using Fe(II) precipitation. Here, we report bench-scale experiments and large-scale demonstrations on removal of 99 Tc (0.1 mM), 95 Nb, and 95 Zr from dissolved Mo disk simulant solutions. Tc removal is accomplished through the reduction and simultaneous immobilization of Tc(VII) by Fe(II), producing an insoluble, Tcincorporated Fe(II)-Fe(III) solid that is removed by filtration. Large-scale testing showed that Tc (97.6% ± 1.7%) is removed without affecting 99 Mo yields. Other important side reaction product impurities in irradiated Mo targets, namely various isotopes of Nb and Zr, are also removed (>99%) in this process. The effects of metal (Fe, Tc, Mo) concentrations, Fe(II)/Fe(III) ratio, pH, temperature, Fe addition method, and hydrogen peroxide addition were tested. The minimum amount of Fe(II) needed to remove up to 99% Tc is 10 mM and the optimal pH value for simultaneous removal of Tc, Nb, and Zr impurities is pH 13.