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The adsorption of U(VI) onto dried and pyrolyzed tea and coffee wastes was investigated. The adsorption properties of the materials were characterized by measuring uranium uptake as a function of solution pH, kinetics and adsorption isotherms. pH profile of uranium adsorption where UO is expected to be the predominant species was measured between pH 0 and 4. Both Langmuir and Freundlich adsorption models were used to describe adsorption equilibria, and corresponding constants evaluated. Using the Langmuir model, the maximum adsorption capacity of uranium by dried tea and coffee wastes was 59.5 and 34.8 mg/g, respectively at 291 K. Adsorption thermodynamic constants, Δ H° Δ S° and Δ G° were also calculated from adsorption data obtained at three different temperatures. Adsorption thermodynamics of uranyl ions on dried tea and coffee systems indicated spontaneous and endothermic processes. Additionally, a Lagergren pseudo-second-order kinetic model was used to fit the kinetic experimental data for both adsorbents and the constants evaluated. Dried tea and coffee wastes proved to be effective adsorbents with high capacities and significant advantage of a very low cost. [ABSTRACT FROM AUTHOR]

In a geopolymer matrix prepared by mixing a metakaolinite precursor with a solution of potassium hydroxide and silicate, 1 wt% Pb as the nitrate has been immobilized. Under the United States Environmental Protection Agency test protocol the Pb release was less than 5 ppm, the acceptable limit for landfills in the U.S.A. Electron microscopy showed Pb was present in the major amorphous phase and a minor (∼1 vol%) Ca-rich potassium silicate phase. Heating cured samples to several hundred degrees Centigrade was not significantly advantageous for immobilizing Pb. [ABSTRACT FROM AUTHOR]

Synroc, a candidate nuclear wasteform and Synroc technology, a waste treatment solution utilizing hot‐isostatic pressing (HIPing) have significant potential for the immobilisation of challenging nuclear wastes from both current and innovative reactors and fuel cycles. Hot isostatic press (HIP) consolidation is undertaken within sealed metal HIP canisters, where metal buffers (e.g., Ti, Fe and Ni) can be incorporated to control the redox environment within the canister. This study, for the first time, reports the effect of varying Ti‐metal addition (0, 2, 4, and 8 wt.%) on phase formation, microstructural characteristics, and wasteform performance for HIP consolidated Synroc‐C containing 20 wt.% simulated PUREX type (PW‐4b) high level waste. Quantitative X‐ray diffraction analysis, scanning electron microscopy‐energy dispersive X‐ray spectroscopy (EDS) and transmission electron microscopy‐EDS analyses were undertaken for analytical investigations. The chemical durability of the samples was assessed using ASTM C1220‐21 standard test. Hot‐isostatically pressed (HIPed) samples with 0 and 8 wt.% Ti added for redox control produced unfavourable phase formation. However, the HIPed samples with Ti additions of 2 and 4 wt.% as a redox buffer showed the desired phase formation of Synroc‐C without any significant change to the partitioning of waste elements among the phases along with compatible durability results, when compared to previous literature for hot uniaxial pressing (HUPed) or sintered materials. [ABSTRACT FROM AUTHOR]

[Display omitted] Pyrochlore glass ceramics have been fabricated via in-situ crystallization under reducing conditions by both sintering and hot isostatic pressing (HIPing) as candidate wasteforms for the acidic waste biproduct of Mo-99 radiopharmaceutical production. The tailored wasteform demonstrates flexibility in the wasteform design to receive the required waste variability, it also suitably passes high-level waste performance requirement, and successfully scales to 1 kg scale with 45 wt.% waste loading. U-rich pyrochlore as the major phase was confirmed by X-ray diffraction, scanning electron microscopy and energy dispersive X-ray spectroscopy, with residual glass and minor secondary phases. The presence of both U4+ and U5+ valences in the wasteforms was revealed by diffuse reflectance spectroscopy. Addition of glass content had little influence on the pyrochlore composition but facilitated minor perovskite formation. The up-scaled dense, HIPed sample showed elemental releases of < 2 g/L for all elements in durability experiments. [ABSTRACT FROM AUTHOR]

Molten pyroprocessing salts can be used to dissolve used nuclear fuel from a reactor allowing recovery of the actinides. Previously, ANSTO have demonstrated hot isostatically pressed (HIPed) sodalite glass‐ceramic wasteforms for eutectic (Li,K)Cl salts containing fission products, but this system cannot be used for the analogous molten alkali fluoride salts (eg, FLiNaK), which have utility in the application of the next generation of nuclear reactors. In this work, a novel glass‐ceramic composite wasteform has been prepared by HIPing, as a candidate for the immobilization of fission product‐bearing FLiNaK salts. The wasteform has been tailored to immobilize the high fluoride content of the waste within fluorite, whereas the waste alkali elements are incorporated in a durable sodium aluminoborosilicate glass, with total waste loadings of ~17‐21 wt% achieved. It was also demonstrated that the speciation of Mo‐ and Sb‐simulated fission products was altered by adding Ti metal due to a controlled redox environment. The resulting candidate wasteform has been studied by X‐ray diffraction and scanning electron microscopy, including the HIP canister‐wasteform interaction zone, and its performance assessed via leaching studies using the PCT and ASTM C1220 leaching protocols. Dr Vance very much enjoyed the challenge of wasteform design for problematic nuclear wastes, for which fission product‐bearing FLiNaK salts are a clear example. His ability to hone in on a wasteform solution with viable waste loadings that meet performance requirements was testament to his nearly 40 years experience in nuclear waste immobilization. The samples discussed in this work represent the last wasteform materials that he prepared. [ABSTRACT FROM AUTHOR]

Generally, pharmaceuticals (PhAs) have been extensively detected in various food and wastewater samples. In this regard, solid-phase extraction (SPE) is an excellent choice for their preconcentration, extraction, and fractionation. Typical, graphene-based nanomaterials (GBNMs) have been spaciously utilized as the efficient SPE adsorbents for the extraction of PhAs. Magnetic solid-phase extraction (MSPE) is found to reduce the time and cost of the operation by skipping the centrifugation step. Of note, this review summarizes the current research on the preconcentration of PhAs by GBNMs and provides valuable information about their applications in the determination and analysis of PhAs in various samples of health and environmental importance. Finally, the different analytical techniques used for the preconcentration of PhAs as well as their adsorption mechanism, and different environmental factors affecting the adsorption, have been studied and discussed in details. [ABSTRACT FROM AUTHOR]

CuI is a candidate wasteform for the immobilization of the fission product 129 I. CuI can be made simply by the addition of CuCl to an I − bearing solution such that exchange of Cl − with I − takes place. The CuI material can then be consolidated into a wasteform by sintering at approximately 550 °C in argon or by hot isostatically pressing at 550 °C with 100 MPa of pressure. A waste loading of greater than 60 wt.% is achievable with good water leach resistance, in keeping with the low solubility product of CuI. However, like the well known wasteform candidate AgI, CuI decomposes in water containing metallic Fe. To compensate this deficiency, the sintered CuI wasteform can be further protected by surrounding it by Sn powder and HIPing at the low temperature of 200 °C. [ABSTRACT FROM AUTHOR]

• A glass wasteform was designed for Mo-99 production Li 2 SO 4 -rich wastes. • Incorporation of 2.78 wt% SO 3 was achieved in the glass melted at 1200 °C. • At low temperature an immiscible layer of BaSO 4 and Na 2 SO 4 formed. • Formation of Na 2 SO 4 lowered the sulphur incorporation rate compared of BaSO 4. • PCT leach values showed satisfactory aqueous durability as glass wasteform. This study has systematically investigated the effect of Li 2 SO 4 addition (2.75 −16.5 wt%) in barium borosilicate glass, to provide a pathway to optimise the glass composition and maximise sulphate incorporation. The work also provides a mechanistic understanding as to how SO 4 2- is incorporated within the glass structure. The highest sulphate incorporation of 2.78 wt% SO 3 (from 11 wt% Li 2 SO 4 addition) was achieved without crystallisation following melting at 1200 °C. Sulphate incorporation in glass was confirmed by XRF, ICP, EDS and Raman analysis. Addition of Li 2 SO 4 along with sodium and barium oxides improved the sulphate incorporation by mixed alkali network depolymerisation and the larger Ba cations helped to create sufficient space within the boron-silicate network to incorporate sulphate ions into the glass. An immiscible sulphate layer rich in BaSO 4 and Na 2 SO 4 formed on top of the glass at lower temperature (800–1100 °C) and subsequent diffusion of Na, Ba oxides and sulphur from this layer increased with increasing time and temperature to form a sulphate incorporated amorphous glass. Addition of Na 2 O played an important role to improve sulphate incorporation in the glass, as well as formation of an immiscible layer on top of the glass however, the formation of Na 2 SO 4 lowered the sulphur incorporation rate at high temperature compared to BaSO 4. Increasing the Li 2 SO 4 content in the glass decreased the glass transition temperature. Aqueous durability testing using the standard PCT tests indicated the glass had satisfactory aqueous durability compared to benchmark environmental assessment glass. This study provides opportunities to convert Li+ and SO 4 2- rich nuclear wastes into appropriate glass wasteforms. [ABSTRACT FROM AUTHOR]

The mechanistic effects of long-term γ irradiation on the mineralogical, microstructural, structural, physical, and chemical properties of 40 wt% blast furnace slag + 60 wt% fly ash geopolymer pastes have been examined. Ambient curing for 28 days during normal equilibration was followed by exposure to 60Co irradiation (1574, 4822, 10,214 kGy). The material characteristics are controlled largely through the competing mechanisms of beneficial equilibration at initial lower dosages, which enhances gelation and crosslinking, and detrimental equilibration at subsequent higher dosages, which causes structural and microstructural destabilisation. Irradiation for 2 months (1574 kGy) increases the compressive strength ~45% (~57 to ~83 MPa) through conversion of less-crosslinked (Q0/Q1/Q1′) to more-crosslinked (Q2/Q3/Q4) silicate species. The transition between these regimes occurs after ~5 months of irradiation (~4000 kGy). Beyond this, the rates of beneficial equilibration and detrimental equilibration equalise upon completion of normal geopolymerisation. Additional geopolymerisation from γ irradiation is controlled by the rate-limiting release of Si4+ from the unreacted aluminosilicates and silicates and their rapid incorporation in the geopolymer network. The aqueous leaching of the geopolymer pastes is not affected significantly by γ irradiation. These data reveal the potential for these materials as intermediate-level wasteforms that can outperform Portland cement-based materials. ga1 • Geopolymer properties are affected by two competing mechanisms during γ irradiation. • Beneficial equilibration at lower doses enhances curing and crosslinking. • Detrimental equilibration at high doses cause structural/microstructural destabilisation. • γ irradiation for 2 months causes an increase in compressive strength. • These geopolymers can be used successfully for immobilisation of ILW. [ABSTRACT FROM AUTHOR]

Structured carbon adsorbents were prepared by carbonizing macroporous polyacrylonitrile beads whose pores were lined with a mesoporous phenolic resin. After activation, the beads were tested for minor actinide (Np and Am), major actinide (Pu and U) and lanthanide (Gd) adsorption in varying acidic media. The activation of the carbon with ammonium persulfate increased the surface adsorption of the actinides, while decreasing lanthanide adsorption. These beads had a pH region where Pu could be selectively extracted. Pu is one of the longest lived, abundant and most radiotoxic components of spent nuclear fuel and thus, there is an urgent need to increase its security of storage. As carbon has a low neutron absorption cross-section, these beads present an affordable, efficient and safe means for Pu separation from nuclear waste. [ABSTRACT FROM AUTHOR]