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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]

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]

• 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]

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]