Carbonate of strontium anchored at the edge of graphite

Abstract

With the continuous growth of the fourth-generation nuclear reactor, the oxidation of irradiated graphite (i-graphite) has become a great concern worldwide in recent years, not only for the far-reaching significance in the treatment of i-graphite waste, but also for the safety evaluation of the reactor. Fission products (FPs) and activation of impurities are generated in i-graphite during reactor operation and inevitably released as radioactive dust when i-graphite is oxidized. The investigation of the binding mechanism between nuclides and graphite dust is vital for further separating procedures of off-gas decontamination. Herein, the mechanism of FP release during the oxidation of High-Temperature Gas-Cooled Reactors’ matrix graphite was explored with Strontium (Sr) as an example. The Sr-loaded matrix graphite was prepared by impregnation and characterized by secondary ion mass spectrometry (SIMS), X-ray powder diffraction (XRD), scanning electron microscope-energy dispersive spectrometer (SEM-EDS), and Raman spectra. The released Sr chemical speciation was extrapolated as SrCO3 through the investigation of the temperature and air flow rate effects on oxidation products. Additionally, the main morphology of the released fragment was identified as SrCO3 clusters anchored at the edge of graphite. The ab initio calculations revealed the mechanism for the anchoring of SrCO3 at the edge of graphite via adsorption. These findings provide a novel understanding of Sr-releasing behavior when i-graphite is oxidized.

Keywords

• Strontium
• Irradiated nuclear graphite
• High-temperature oxidation
• Release
• Graphite edge