Role of ion exchange in the corrosion of nuclear waste glasses

Abstract

The ion exchange phase of corrosion of nuclear waste glasses was modelled using Doremus' theory of interdiffusion and numerically analysed for British Magnox waste and Russian K-26 glasses. It is shown that even in non-silica -saturated conditions the ion exchange phase plays a significant role in the overall radionuclide release inventory particularly for shortlived radionuclides. INTRODUCTIONStudies of archaeological artefacts show that the principal processes in the natural corrosion of silicates are diffusion-controlled ion exchange reactions leading to selective leaching of alkalis and proton entering the silicate structure to produce a hydrated alkali-deficient layer and congruent dissolution with destruction of the silicate network and subsequent precipitation of hydrous silica-gel layers as secondary alteration products [1,2]. At ambient temperatures the ion exchange reactions persist for extended periods of time and govern alteration processes for thousands [1] and tens of thousands of years [2]. The role of ion exchange in the corrosion behaviour of nuclear waste glasses is also considered to be important [3,4]. This paper analyses the ion exchange phase in the corrosion behaviour and radionuclide release inventory for two nuclear waste glasses: British Magnox waste and Russian high sodium waste glass K-26.Radionuclide immobilisation in glass can be affected by corrosion processes which may lead to partial mobilisation and release of radionuclides into the environment. The potential contact of water with glass is deferred in actual disposal systems to times after the waste container has been breached. For vitrified high-level waste (HLW) these times may be of the order of many hundreds or even thousands of years. High temperatures and radiation dose rates are likely only for the first few hundred years after HLW vitrification so that container temperatures will be close to those of the ambient rock by the expected time of contact with ground water. Moreover the role of βγ-radiolysis will also become negligible because of low radiation dose rates after these long times. Vitrified low and intermediate waste (LILW) is almost invariably at the ambient temperature of a repository environment. This type of waste is also expected to be disposed of in near-surface repositories which are characterised by relatively low host rock temperatures. Thus the most likely temperatures of future corrosion events are expected to be close to those of the disposal environment which makes them similar to those that have occurred with archaeological artefacts.Aqueous corrosion of nuclear waste glasses is a complex process which depends on many parameters such as time, temperature, groundwater chemical composition and pH. Although corrosion of silicate glasses including nuclear waste-containing borosilicate glasses is a complex process it involves two major reactions -diffusion-controlled ion exchange and glass network hydrolysis [1][2][3][4][5][6][7]. In dilute solutions hydrolysis of the glass network controls the late stages...

The Influence of Radiation on Confinement Properties of Nuclear Waste Glasses

“…e irradiation influences on the stability of the matrix have two factors: (1) a change in the structure and properties of the matrix under radiation and (2) radiogenic effects at the water-matrix interface, in particular, change in characteristics of groundwater due to its radiolysis (decrease in pH and so on). e authors in [5,20,44] have analysed the effect of selfirradiation on two corrosion mechanisms of nuclear waste glasses: (1) diffusion-controlled ion exchange and (2) hydrolysis.…”

“…Tests of radioactive glasses showed that initial leaching rates are significantly higher for radioactive samples compared to nonradioactive simulants although the long-term hydrolytic stage was unaffected [49]. Summarizing the data available [2,5,20,42,44] led to the conclusion that, in cases when corrosion occurs via diffusion-controlled ion exchange, the irradiation has a detectable and significant impact. is occurs at relatively low temperatures (≤50°C), low and medium pH (≤8), relatively high absorbed doses (>6•10 6 Gy), and for diluted solutions at initial stages of corrosion.…”