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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]