Ceramic Immobilization Options for Technetium
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Ceramic Immobilization Options for Technetium Martin C. Stennett1, Tae-Hyuk Lee1,2, Daniel J. Bailey1, Erik V. Johnstone1, Jong Heo2, and Neil C. Hyatt1 1 NucleUS Immobilisation Science Laboratory, Department of Materials Science and Engineering, The University of Sheffield, Mappin Street, Sheffield, S1 3JD, UK. 2 Department of Materials Science and Engineering and Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), Pohang, Gyeongbuk, 37673, Republic of Korea. ABSTRACT 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. INTRODUCTION Of the many challenges facing the nuclear industry one of the most important is the issue of waste management; waste arising from nuclear fuel cycle operations must be rendered passively safe prior to disposal. Safe sequestration is very challenging for 99Tc because it has a half-life of approximately 210000 years and it is present in spent fuel rods as the technetium (VII) oxide (Tc2O7) which is extremely water-soluble, forming the mobile anionic pertechnetate species (TcO4-) in solution. Additionally 99Tc is produced during nuclear fission of 235U, with a yield of 6.06 % [1], which means that approximately 1 kg of 99Tc is produced for every tonne of enriched 235U that is used in a nuclear reactor [2]. Ceramic and vitreous materials offer promising host matrices for radionuclides however for safe immobilization the difficulties associated with processing 99Tc must be addressed. Many higher oxidation state technetium species, such as Tc2O7, are highly volatile and unstable so care must be taken to retain 99Tc in its Tc4+ oxidation state. TcO2 is considerably less volatile than Tc2O7; TcO2 has a sublimation temperature of approximately 900°C, in contrast to the melting and boiling points of Tc2O7, which are 119.5°C and 311°C, respectively [3]. It is possible to immobilize Tc4+ by incorporating it into the structure of a durable ceramic phase which has suitably sized crystallographic sites. In this work CaZrTi2O7, which crystallizes with the zirconolite structure, w
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