Simulation of Self-Irradiation of High-Sodium Content Nuclear Waste Glasses
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Simulation of Self-Irradiation of High-Sodium Content Nuclear Waste Glasses Alexey S. Pankov1, Olga G. Batyukhnova2, Michael I. Ojovan1, and William E. Lee3 1 Immobilisation Science Laboratory, Department of Engineering Materials, University of Sheffield, Sir Robert Hadfield Building, Mappin Street, Sheffield, S1 3JD, United Kingdom 2 International Education Training Centre, SUE SIA “Radon”, The 7-th Rostovsky Lane 2/14, Moscow, 119121, Russian Federation 3 Department of Materials, Imperial College London, South Kensington Campus,Exhibition Road, London, SW7 2AZ, United Kingdom ABSTRACT Alkali-borosilicate glasses are widely used in nuclear industry as a matrix for immobilisation of hazardous radioactive wastes. Durability or corrosion resistance of these glasses is one of key parameters in waste storage and disposal safety. It is influenced by many factors such as composition of glass and surrounding media, temperature, time and so on. As these glasses contain radioactive elements most of their properties including corrosion resistance are also impacted by self-irradiation. The effect of external gamma-irradiation on the short-term (up to 27 days) dissolution of waste borosilicate glasses at moderate temperatures (30° to 60°C) was studied. The glasses studied were Magnox Waste glass used for immobilisation of HLW in UK, and K-26 glass used in Russia for ILW immobilisation. Glass samples were irradiated under γ-source (Co-60) up to doses 1 and 11 MGy. Normalised rates of elemental release and activation energy of release were measured for Na, Li, Ca, Mg, B, Si and Mo before and after irradiation. Irradiation up to 1 MGy results in increase of leaching rate of almost all elements from both MW and K-26 with the exception of Na release from MW glass. Further irradiation up to a dose of 11 MGy leads to the decrease of elemental release rates to nearly initial value. Another effect of irradiation is increase of activation energies of elemental release. INTRODUCTION Vitrification is the main option of managing the high-level radioactive wastes (HLW) arisen at the reprocessing of spent nuclear fuel. Alkali-borosilicate glasses are widely used for the immobilisation of these wastes because at the time when first vitrification plants were being planned and constructed this material was the best compromise between production efficiency and relative durability. There were developed many other immobilisation technologies since then. However at the present day immobilisation of HLW in form of borosilicate glass is the most common approach. Durability or corrosion resistance of materials used for immobilisation of radioactive waste is one of the key parameters in waste disposal safety. To secure long-term disposal, it is very important to understand and be able to predict the evolution of the glass and its’ properties due to the self-irradiation. Formation of oxygen bubbles and volume change, for instance, are the most known effects of α- and β- irradiation [1, 2]. Relatively recent studies of irradiation (mainly el
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