Cerium incorporation in zirconolite glass-ceramic systems, consolidated by hot isostatic pressing, was investigated. Samples were formulated to target Ce incorporation on the Ca and / or Zr sites. Results show successful incorporation of Ce into the ceramic phase although complete waste digestion was not achieved. The formation of a Ce-bearing perovskite phase when targeting Zr substitution, was associated with reduction of Ce4+ to Ce3+. This study is part of preliminary work towards fabricating Pu-bearing glass-ceramic HIP samples.

High Level Waste (HLW) glasses are to be stored in deep geologic repositories around the world. Some potential repository geologies have oxidizing groundwaters while some have reducing or anoxic groundwaters. The differences in the oxidizing potential of the groundwater, expressed as groundwater Eh, which has a profound impact on the release of multivalent species such as iron and uranium from the glass. Static leach testing of monolithic glass samples (ASTM C1220) doped with uranium were performed at 90°C in an Ar glovebox under anoxic and oxic conditions. Tests were performed in both deionized water and in simulated basaltic groundwater that was pre-equilibrated at low Eh. Geochemical modeling of the measured Eh-pH conditions from the oxic and anoxic experiments using Geochemist’s Workbench software, suggested that different colloidal species control the release of uranium species under oxic and anoxic conditions.

In a scenario where ground water enters a canister for spent nuclear fuel in a deep geological repository, the presence of dissolved ions in the water could possibly influence the fuel dissolution due to effects on radiolysis yields. One species of particular interest in this context is bromide, which has a proven ability to scavenge hydroxyl radicals much faster than molecular hydrogen does. As a result, bromide could inhibit the beneficial effect of dissolved hydrogen, which has been shown in γ-radiolysis experiments. However, already a few hundred years after repository closure, α-decay starts to dominate in the radiation field from the spent fuel. Hence, the effects of α-radiolysis are expected to govern the fuel dissolution over the geological timeframes of the repository. In the present work, α-radiolysis experiments have been performed to determine the effect of bromide ions on the yield of hydrogen peroxide by mass spectrometric measurement of its decomposition product oxygen. The use of high activity 238Pu solutions has made it possible to study this effect during pure α-radiolysis from a homogeneously distributed radiation field. To simulate deep bedrock repository conditions, and to minimize the influence of in-leaking O2 from air, the studies were performed using graphite sealed stainless steel autoclaves with an initial atmosphere of 10 bar H2. The results show that addition of 1 mM Br- to the solution gives no significant effect on the O2 yield for radiation doses up to 2 MGy. This lack of effect is most likely explained by the limited radical escape yields from radiation tracks in pure α-radiolysis.

At the Savannah River Site (SRS) low activity salt solution is stabilized via encapsulation within a grout termed saltstone. Saltstone is emplaced into large (multi-million gallon) concrete storage facilities referred to as Saltstone Disposal Units (SDUs). Technetium-99 (99Tc) is a long-lived radionuclide contained in the low activity salt waste and subsequently incorporated into the grout waste form: it is considered a significant contributor to risk with respect to the long-term radiation exposure of the environment surrounding the SDUs. In the reducing, high pH environment within the grout, 99Tc is expected to be relatively immobile since it exists in a reduced Tc(IV) oxidation state in the form of sparingly soluble sulfides (TcSx) or hydrated oxides (TcO2.xH2O). However, in the presence of O2 (associated with the future infiltration of air or oxygenated ground waters into the saltstone monolith) it is possible for redox-sensitive Tc(IV) to transition into highly soluble (and mobile) Tc(VII) species, pertechnetate (TcO4-), which is more readily transported to the surrounding environment. Traditional approaches to quantifying the leaching behavior of 99Tc from cementitious matrices have involved partitioning experiments using size-reduced (crushed/ground) saltstone samples, and determination of the 99Tc fraction immobilized by the cementitious solids. Such experiments create artificially high solid-solution contact areas that likely result in higher 99Tc leachate concentrations than would be expected for intact, monolithic samples. In the current study a new technique, termed the Dynamic Leaching Method (DLM), is being used to investigate the 99Tc leaching behavior of monolithic saltstone samples. The data derived using this technique is intended to inform the SRS Saltstone Disposal Facility (SDF) Performance Assessment (PA) which models the long-term transport of radionuclides from the SDUs to the environment. The DLM utilizes a flexible-wall permeameter to achieve saturated leaching under an elevated hydraulic gradient in an effort to simulate the transport of groundwater through saltstone. Initial findings indicate that the 99Tc concentrations in the leachate are on the order of 1E-08 mol/L which suggests that the saltstone leaching behavior is controlled by the solubility of TcO2.xH2O compounds.

We report experimental and computational studies of Ba doping for Cs in Cs2TiNb6O18, a material with potential to be an exceptional ceramic waste form for Cs sequestration. Three co-doping (simultaneous metal reduction for charge balance of Ba2+ for Cs+) schemes have been experimentally tested: Ti4+ for Nb5+, Ti3+ for Ti4+ and Nb4+ for Nb5+. Unfortunately, none showed conclusively that the co-substitution was successful. Atomistic modelling was then performed on all three schemes using novel potentials to assess the energetic feasibility, from these the most favourable scenario is reduction of Nb5+ to Nb4+.

Pure and Nb-doped zirconium germanate materials of composition K2-xZr1-xNbxGe3O9.H2O where x = 0, 0.1, 0.2 and 0.3 with the structure of the natural mineral umbite have been prepared in high yield using hydrothermal synthesis methods. The parent material displays virtually no ion exchange of the K+ for Cs+ but the doped materials show rapidly enhanced exchange with replacement of ca. 70% of the K+ by Cs+ for the 30% doped material. Rietveld analysis of the powder X-ray diffraction data is consistent with no change in the unit cell parameters or K+ bonding prior to the exchange, hence we propose the improved property is due to the creation of cation defect sites within the pores of the material that facilities greater cation mobility and leads to exchange.

Key drivers for the application of thermal treatment processes include the reduced volume, improved passive safety, and superior long term stability of the vitrified wasteform products. These advantages have led to a renewed interest in thermally treating various UK ILW streams. To support the increased investment in thermal treatment technologies a fundamental understanding of the processes and the impact of waste inventory needs to be established. The research described in this report aims to provide the evidence necessary to support a major investment in thermal treatment of Plutonium Contaminated Materials (PCM). The report describes the use of laboratory scale waste simulants to develop an understanding of the waste and matrix interactions during thermal treatment of PCM waste. The report includes the thermal treatment process and characterization of the vitrified product.

Borate is present in natural groundwaters and borate is also released into groundwaters when borosilicate glass, waste form for high level nuclear waste, is corroded. Borate can form an aqueous complex, AmHB4O72+, with actinides in +III oxidation state. In this work, we present our evaluation of the equilibrium constant for formation of AmHB4O72+ and the associated Pitzer interaction parameters at 25°C.

Using Nd(III) as an analog to Am(III), solubility data of Nd(OH)3(s) in NaCl solutions in the presence of borate ion from the literature, is used to determine Am(III) interactions with borate. The log10K for the formation reaction is 37.34. This evaluation is in accordance with the Waste Isolation Pilot Plant (WIPP) thermodynamic model in which the borate species include B(OH)3(aq), B(OH)4–, B3O3(OH)4–, B4O5(OH)42–, and NaB(OH)4(aq). The WIPP thermodynamic database uses the Pitzer model to calculate activity coefficients of aqueous species.

In addition, the equilibrium constant for dissolution of AmB9O13(OH)4(cr) at 25oC is evaluated from the solubility data on NdB9O13(OH)4(cr) in NaCl solutions, again using Nd(III) as an analog to Am(III). The log10K for the dissolution reaction is –79.30. In the evaluation for log10K for the dissolution reaction, AmHB4O72+ is also considered.

The equilibrium constant and Pitzer parameters evaluated by this study will be important to describe the chemical behavior of Am(III) in the presence of borate in geological repositories.

While conducting siderite (FeCO3) solubility experiments in NaCl-Na2CO3 brines, evidence for a second phase was detected. Experiments, in which synthesized siderite was reacted with high ionic strength (0.18 – 7.5 m) solutions at room temperature and high pH (>10), were conducted in a glovebox. As the aging time of siderite-bearing experiments increased, the pH of the solution decreased, signaling formation of a hydroxyl-bearing phase. Decreasing pH values are interpreted to indicate that a hydroxyl-bearing phase, such as chukanovite, is the reaction controlling solid in the solid assemblage. Chukanovite was tentatively identified by XRD analysis. We set out, therefore, to determine the thermodynamic stability of chukanovite under the experimental conditions. Aqueous thermodynamic model parameters were determined with experimentally analyzed Fe(II) solubility data, and subsequently yielded a proposed formation free energy of chukanovite (-1149.8 kJ/mol).

Long half-life biologically active fission products, such as technetium-99, present particular problems for the disposal of spent nuclear fuel (SNF). Technetium is present in relatively high concentrations in fuel (approx. 1kg tonne-1 SNF) and has very high mobility in oxidizing environments. Technetium is therefore generally removed from SNF either by solvent extraction and reduction, during the PUREX process, or by sorption via ion exchange processes. Historically technetium has been disposed of via dilution and dispersion in the sea but stringent regulations now mean that the preferred long term option is immobilization in a highly stable and durable matrix. In this contribution we have looked at the synthesis of fluorite derivative crystalline host phases based on the zirconolite structure. Samples have been characterised by X-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersive spectroscopy (EDX), thermo-gravimetric analysis (TG), and mass spectroscopy (MS). We have used Mo as an inactive surrogate for Tc.