Since the year 2000, Synroc has evolved from the titanate full-ceramic waste forms developed in the late 1970s to a hot isostatic pressing (HIP) technology platform that can be applied to produce glass, glass–ceramic, and ceramic waste forms and where there are distinct advantages over vitrification in terms of, for example, waste loading and suppressing volatile losses. This paper describes recent progress on waste form development for intermediate-level wastes from 99Mo production at ANSTO, spent nuclear fuel, fluoride pyroprocessing wastes and 129I. The microstructures and aqueous dissolution results are presented where applicable. This paper provides perspective on Synroc waste forms and recent process technology development in the nuclear waste management industry.

Zirconolite glass-ceramics are being developed as potential wasteforms for the disposition of Pu wastes in the UK. Previous studies utilised a variety of surrogates whilst this work uses both cold-press and sinter and hot isostatic press methods to validate the wasteform with PuO2. A cold press and sinter sample was fabricated as part of a validation study for plutonium incorporation in hot isostatically pressed (HIPed) wasteforms. The results confirmed the cold-press and sinter, achieved successful waste incorporation and a microstructure and phase assemblage that was in agreement with those expected of a HIPed equivalent. A HIP sample was fabricated of the same composition and characterised by SEM and XRD. Results were in agreement with the sintered sample and achieved complete waste incorporation into the glass-ceramic wasteform. These samples have demonstrated successful incorporation of PuO2 into glass-ceramic HIPed wasteforms proposed for processing Pu-based waste-streams in the UK.

Nagelschmidtite, Ca7P2Si2O16, is an end-member of continuous solid solution Ca2SiO4 – Ca3(PO4)2⋅2Ca2SiO4 within the pseudo-binary system Ca3(PO4)2 – Ca2SiO4 (whitlockite – larnite). This phase is capable to wide isomorphic exchanges in Ca, P and Si sites: Ca2+ = Sr2+; Ca2+ = Eu2+; Ca2+ + P5+ = (RE,An)3+ + Si4+, 2Ca2+ = Na+ + (RE,An)3+; 2Ca2+ = An4+ + ☐; Ca2+ + Si4+ = (RE,An)3+ + (Al,Fe)3+; Ca2+ + Si4+ = Na+ + P5+; 2Ca2+ = Na+ + (Al,Fe)3+; Ca2+ + P5+ = Na+ + S6+. It was found in metallurgical slags and geological formations. We revealed nagelschmidtite-type phase in vitrified phosphorus-bearing radioactive incinerator slags. The materials were glass-crystalline and contained nano-sized nagelschmidtite crystals distributed in vitreous matrix phase. Average chemical composition of the largest (few microns) crystals was recalculated to formula Na1.21K1.05Ca2.22Al2.02Fe0.46Si2.69P1.26U0.08O15.76. Significant oxygen misbalance suggests higher than U(IV) oxidation state for uranium – U(V) or U(VI). Capability of nagelschmidtite to be crystallized from melt makes it promising phase for actinides, rare earths and some other fission and corrosion products at using a melting route to nuclear waste forms including cold crucible induction melting and self-propagating high-temperature synthesis.

Slow cooling of phosphate melt at liquid nuclear waste solidification yields glass-crystalline material. Partial crystallization during melt solidification results in elemental partitioning among crystalline phase and glass: Al, Cr, Fe are concentrated in the crystalline phosphate phase while Ca, Ni, La, U enter predominantly in the residual glass. Glass dissolution rate and leach rate of La and U as rare earth and actinide surrogates depends strongly on the glass composition, for example reduction of Al2O3 content in the glass to ∼10-12 wt.% increases leachability by three orders of magnitude as compared to the glass with specified composition (∼18-22 wt.% Al2O3).

Nuclear energy is one option to meet rising electricity demands, although one concern of this technology is the proper capture and storage of radioisotopes produced during fission processes. One of the more difficult radioisotopes is 129 I due to its volatility and poor solubility in traditional waste forms such as borosilicate glass. Iodosodalite has been previously proposed as a viable candidate to immobilize iodine due to high iodine loading and good chemical durability. Iodosodalite was traditionally synthesized using solid state and hydrothermal techniques, but this paper discusses an aqueous synthesis approach to optimize and maximize the iodosodalite yield. Products were pressed into pellets and fired with glass binders. Chemical durability and iodine retention results are included.

Sodalites have been investigated experimentally for the capture and long-term containment of 129I, a significant and hazardous waste product of the nuclear fuel cycle. Sodalites are zeolite-type structures commonly occurring in nature in alkaline igneous rocks and having the prototype formula Na8(AlSiO4)6Cl2. The crystal structure is based around β-cages consisting of corner-sharing SiO4 and AlO4 tetrahedra. In the centre of the β-cage is an anion X. Iodine captured by sodalites sits in the centre of the β-cages as iodide anions. Silver iodide (AgI) plays an important role in the capture and subsequent processing of 129I in the nuclear fuel cycle. Using ab initio density functional theory (DFT) modelling, we investigate the energetics and feasibility of iodine capture and containment in mixed cation sodalites Na8-xAgx(AlSiO4)6I2, and compare the results with experimental observations.

Immobilization of bulk liquid high level radioactive waste (HLW) in Synroc ceramic is well-known reliable way for final isolation of dangerous long-lived radionuclides from biosphere. The alternative method of Synroc-like ceramic synthesis has been proposed. Radionuclide incorporation into crystalline titanate host-phases can be provided as a result of direct radionuclide sorption from liquid HLW using non-selective sorbent – layered hydrazinium titanate (LHT-9). Such an approach allows excluding expensive multi-stage procedure of precursor preparation. The precipitate obtained after sorption can be easily transformed into Synroc-like ceramic by cold pressing followed with sintering in air at 1000-1200°C. The highly radioactive samples of titanate ceramic loaded with real HLW have been synthesized and preliminary studied at KRI hot-cell facility. Chemical durability of this sample has been studied using static leach test in distilled water at 90°C and the leach rates for 154Eu, 241Am, 244Cm were (in g·m-2·day-1) from 2·10-4 to 5·10-3. Normalised 137Cs mass loss was 0.3 g·m-2 for 110 days at the same conditions.

With the aim of creating novel ceramics for applications in nuclear materials with high radiation tolerance, multiple samples with A-B-O stoichiometries ranging from 215 to 227 were synthesized and characterized by a combination of SEM, XRD, and TEM methods. Single-phase defect-fluorite-type compounds with A = Sm or Yb and B = Ti, Zr, and/or Sn are reported; whereas, pyrochlore structured compounds and lanthanide sesquioxide phases were found as major phases in numerous samples. A series of Y-b-Sn-O samples was successfully prepared as nearly single phase or single phase materials. These are essentially defect fluorites with extra weak peaks, most likely due to X-ray scattering from oxygen or oxygen and metal cations. We describe some interesting TEM data and describe selected area diffraction patterns with complex modulations.

There is growing interest in reducing the use of ordinary Portland cement (OPC) owing to its high energy consumption and CO2 emissions. An environmentally-friendly alternative is the use of geopolymers, which can potentially reduce direct CO2 emissions through the appropriate choice of raw materials, mix design, and curing regimes. In this regard geopolymer mortars are also realistic candidates for the replacement of OPC mortars in nuclear waste immobilisation applications as they provide a more durable incorporation matrix as well as suppressing the formation of radiolytic hydrogen. The advantages of geopolymers over OPC for nuclear waste immobilisation include i) lower water content as alkaline activator is the main component that drives geopolymerisation, ii) higher thermal stability (<600°-800°C) compared to OPC concrete (<300°C), iii) higher compressive strength (50-80 MPa), and iv) lower leachability of radioactive ions when the mix design and curing temperature are appropriately balanced. UNSW and ANSTO have embarked on a long-term research program to investigate the possibility of using geopolymers for the immobilisation of Intermediate Level Liquid Waste (ILLW), the focus of which will be around the influence of gamma-irradiation on the durability.