Removal of TcO4- from Representative Nuclear Waste Streams with Layered Potassium Metal Sulfide Materials

Many efforts have focused on the sequestration and immobilization of 99Tc because the radionuclide is highly mobile in oxidizing environments and presents serious health risks due to its radiotoxicity and long half-life (t1/2 = 213 000 a). One of the more common methods for Tc removal from solution and immobilization in solids is based on reducing Tc from highly soluble Tc(VII) to sparingly soluble Tc(IV). Here, we report results obtained with two potassium metal sulfides (KMS-2 and KMS-2-SS) that are capable of reducing Tc(VII) to Tc(IV). Batch sorption experiments were performed in both oxic and anoxic conditions for 15 d in both deionized water (DIW) and a highly caustic (pH ∼ 13.6), high ionic strength (8.0 mol L-1), low-activity waste (LAW) stream simulant solution. Tc removal for both materials in DIW is improved in anoxic conditions compared to oxic conditions as a result of a higher solution pH. In DIW and anoxic conditions, KMS-2 is capable of removing ∼45% of Tc, and KMS-2-SS is capable of removing ∼90% of Tc. Both materials perform even better in the LAW simulant and remove more than 90% of available Tc after 15 d of contact in anoxic conditions. Postreaction solids analyses indicate that Tc(VII) is reduced to Tc(IV) and that Tc(IV) is bonded to S atoms in a Tc2S7 complex. Examination of the materials after Tc removal by X-ray diffraction shows that the initially crystalline KMS-2 materials lose much of their initial long-range order. We suggest a Tc removal mechanism wherein the TcO4- enters the interlayer of the KMS-2 materials where it is reduced by sulfide, which results in a distorted crystalline structure and a solid-state Tc2S7 complex

Corrosion of the International Simple Glass under acidic to hyperalkaline conditions

Assessment of glass dissolution kinetics, under disposal relevant temperature and pH environments, is required to credibly estimate radionuclide release rates from vitrified radioactive waste. Leaching of the International Simple Glass (ISG) under acidic to hyperalkaline conditions was examined. Forward rate measurements have been obtained using the dynamic leaching SPFT protocol and rate parameters for B, Na and Si in the basic regime; errors in rates predicted using these parameters at high pH and temperature are significant because the fitting uses logarithmic data. Longer term behaviour under hyperalkaline conditions, representative of some disposal environments, was investigated using the PCT and MCC-1 static leaching protocols with Ca(OH)2 solutions for up to 120 days (PCT) and 720 days (MCC-1). In hyperalkaline conditions dissolution was incongruent for all elements and the presence of alternating zirconia-rich and zirconia-poor alteration layers was observed on all leached monoliths, indicating the occurrence of a self-organisation phenomenon during leaching

Chemical Durability of Glasses

International audienceThe chemical durability of silicate glasses has long been studied for many applications, in particular when glasses are subjected to environmental weathering and aqueous corrosion. Typical applications include optical instruments, glass vessels, radioactive waste confinement, and bone reparation. Glass corrosion involves ion exchange, water diffusion, network dissolution-recondensation, and secondary phase precipitation. These reactions may impact, among other things, the release of contaminants from waste glasses, and the glass mechanical, optical and catalytic properties. The glass corrosion mechanisms and alteration product formation have been well studied as a function of many environmental parameters (temperature, pH, water composition, etc.).The present chapter describes the general phenomena behind glass corrosion and details glass dissolution in aqueous conditions on one hand and glass vapor hydration on the other hand. The latter phenomenon has not received the same level of attention in the literature relative to the corrosion in aqueous solutions. Research and development needs, in particular in complex systems such as radioactive waste geological repositories, are discussed in the conclusion of the chapter