Technetium-99m Transport and Immobilisation in Porous Media: Development of a Novel Nuclear Imaging Technique

Abstract

Technetium-99, a β-emitting radioactive fission product of 235 U, formed in nuclear reactors, presents a major challenge to nuclear waste disposal strategies. Its long half-life (2.1 x 10 5 years) and high solubility under oxic conditions as the pertechnetate anion [Tc(VII)O 4 ] is particularly problematic for long-term disposal of radioactive waste in geological repositories. In this study, we demonstrate a novel technique for quantifying the transport and immobilisation of technetium-99m, a γ-emitting metastable isomer of technetium-99 commonly used in medical imaging. A standard medical gamma camera was used for non-invasive quantitative imaging of technetium-99m during co-advection through quartz sand and various cementitious materials commonly used in nuclear waste disposal strategies. Spatial moments analysis of the resulting 99m Tc plume provided information about the relative changes in mass distribution of the radionuclide in the various test materials. 99m Tc advected through quartz sand demonstrated typical conservative behaviour, while transport through the cementitious materials produced a significant reduction in radionuclide centre of mass transport velocity over time. Gamma camera imaging has proven an effective tool for helping to understand the factors which control the migration of radionuclides for surface, near-surface and deep geological disposal of nuclear waste.

Blast furnace slag-Mg(OH)2cements activated by sodium carbonate

“…13 GGBS rich-binders are benecial in the encapsulation of radioactive wastes, as the slag contains sulphide and potentially Fe(II) which act as reducing agents, controlling the solubility of key radionuclides. 14,15 The hydration of GGBS in cements also oen forms layered double hydroxides with a hydrotalcite-type structure (general formula M x 2+ M y 3+ (OH) 2x+3yÀnz (A nÀ ) z $mH 2 O, where A nÀ is oen OH À , Cl À or CO 3…”

Design and Application of Materials for Sequestration and Immobilization of 99Tc

“…Migration of 99 Tc from a waste repository to the environment is impeded by the efficient immobilization of Tc chemical species in durable waste forms, including glass, cements, iron minerals, and synroc/ceramics. − Tc­(VII) as TcO 4 – is a highly stable Tc species under aerobic conditions; however, it is also unstable at high vitrification temperatures (∼1200 °C), resulting in low Tc incorporation in the final glass waste form. Cementation is one of the most common low temperature nuclear waste treatment methods and has also been widely applied for retention and effective immobilization of 99 Tc. − The addition of reducing agents to the cementitious waste form results in the reduction of TcO 4 – to the more stable Tc­(IV) species. − However, the immobilization of 99 Tc via reduction methods is quite challenging because of the possibility of Tc­(IV) reoxidation to the mobile chemical species Tc­(VII)­O 4 – as the waste form interacts with the oxidizing environment. The application of various synroc/ceramics matrixes for immobilization of different radionuclides, including 99 Tc, has been recently reviewed by Gregg et al…”

Use of WetSEM® capsules for convenient multimodal scanning electron microscopy, energy dispersive X-ray analysis, and micro Raman spectroscopy characterisation of technetium oxides

“…Technetium (Tc) is a high-yield fission product of both plutonium and uranium (6% fission yield for 235 U). In the safety case for geological disposal of radioactive wastes, 99 Tc makes a significant contribution to the long term dose risk, due its long half-life (t 1/2 = 2.1 × 10 5 years), and the high solubility and poor sorption of the pertechnetate species, TcO 4 − [1][2][3] In the context of some advanced nuclear fuel reprocessing flowsheets, such as the UREX process [4], an objective is to separate and immobilise Tc in a durable glass, ceramic or metal alloy wasteform for geological disposal [5,6]. Consequently, there is considerable research directed at the synthesis and characterisation of Tc wasteforms and understanding the sorption and migration behaviour of Tc in engineered barrier systems relevant to their geological disposal [3,7].…”