Microstructure of Simulated High-Level Waste Glass Doped with Short-Lived Actinides, 238Pu and 244Cm

The Influence of Radiation on Confinement Properties of Nuclear Waste Glasses

Self-irradiation can affect durability of glasses used to immobilize high-level nuclear waste (HLW). The stability of glasses can also be indirectly affected by the radiolytic changes in contact water leading to decrease in its pH although this is unlikely to occur for disposal systems where the interaction of groundwater with glass is delayed to times when radiation dose rates are decreased to levels insignificant to cause such effects. Besides, interaction of the water influenced by radiation with other repository protective elements (container and bentonite) will suppress the radiolytic changes. Literature analysis shows practical absence or very weak effect of self-irradiation on structure and characteristics of borosilicate glasses with typical content of nuclear waste. Data for aluminophosphate glass used in Russia have showed that, after γ-irradiation with a dose of 6.2·107 Gy, the leaching rates of elements were decreased approximately twice relatively to pristine samples.

Mechanism of powellite crystallite expansion within nano-phase separated amorphous matrices under Au-irradiation

A fundamental approach was taken to understand the implications of increased nuclear waste loading in the search for new materials for long-term radioisotope encapsulation. This study focused on the formation and radiation tolerance of glass ceramics with selectively induced CaMoO₄ as a form to trap the problematic fission product molybdenum. Several samples were synthesised with up to 10 mol% MoO₃ within a soda lime borosilicate matrix, exhibiting phase separation on the nano scale according to thermal analysis, which detected two glass transition temperatures. It is predicted that these two phases are a result of spinodal decomposition with Si-O-Ca-O-Si and Si-O-Ca-O-B units, with the latter phase acting as a carrier for MoO₃. The solubility limit of molybdenum within this matrix was 1 mol%, after which crystallisation of CaMoO₄ occurred, with crystallite size (CS) increasing and cell parameters decreasing as a function of [MoO₃]. These materials were then subjected to irradiation with 7 MeV Au³⁺ ions to replicate the nuclear interactions resulting from α-decay. A dose of 3 × 10¹⁴ ions per cm² was achieved, resulting in 1 dpa of damage within a depth of ∼1.5 μm, according to TRIM calculations. Glasses and glass ceramics were then analysed using BSE imaging, XRD refinement, and Raman spectroscopy to monitor changes induced by accumulated damage. Irradiation was not observed to cause any significant changes to the residual amorphous network, nor did it cause amorphisation of CaMoO₄ based on the relative changes to particle size and density. Furthermore, the substitution of Ca²⁺ to form water-soluble Na₂/NaGd-MoO₄ assemblages did not occur, indicating that CaMoO₄ is resilient to chemical modification following ion interactions. Au-irradiation did however cause CaMoO₄ lattice parameter expansion, concurrent to growth in CS. This is predicted to be a dual parameter mechanism of alteration based on thermal expansion from electronic coupling, and the accumulation of defects arising from atomic displacements.

Radionuclides containment in nuclear glasses: an overview

Radioactive waste vitrification has been carried out industrially in several countries for nearly 40 years. Research into the formulation and long term behavior of high and intermediate level waste glasses, mainly borosilicate compositions, is still continuing in order to (i) safely condition new types of wastes and (ii) design and demonstrate the safety of the disposal of these long-lived waste forms in a deep geological repository. This article presents a summary of current knowledge on the formulation, irradiation resistance and the chemical durability of these conditioning materials, with a special focus on the fate of radionuclides during glass processing and aging. It is shown that, apart from the situation for certain elements with very low incorporation rate in glass matrices, vitrification in borosilicate glass can enable waste loadings of up to ~20 wt% while maintaining the glass homogeneity for geological time scales and guaranteeing a high stability level in spite of irradiation and water contact.