The Evolution of Nuclea Graphite through Ion Irradiation

Abstract

Nuclear graphite is a cornerstone material for current and next-generation reactor systems, where damage resistance and radionuclide retention critically influence reactor performance, safety, and waste management options. This thesis investigates the evolution of nuclear graphite, from irradiation damage and in-service recovery to radionuclide retention and sustainable decontamination, using ion irradiation as a surrogate for radiation damage and graphite contamination. This novel approach overcomes the historical difficulty of simulating and investigating irradiated reactor graphite, which is challenging to access and handle.

Low-energy cobalt ion implantation (25 keV, 1.5 – 30 dpa) was employed to examine the structural recovery and chemical bonding of nuclear graphite grade IG-110, at temperatures representative of advanced reactor operation (550 – 1100 °C). Multi-technique characterisation combining Raman spectroscopy, transmission electron microscopy (TEM), and electron energy-loss spectroscopy (EELS) revealed severe near-surface lattice disorder followed by partial reformation of graphitic domains above 750 °C. Approximately 50% restoration of the sp2/sp3 ratio was achieved after heating to 1100 °C, demonstrating significant self-healing capability under reactor-relevant conditions.

A first-of-its-kind real-time imaging approach was developed using in situ heating within a Scanning Electron Microscopy (SEM) to visualise nanoparticle migration at elevated temperatures. This approach revealed the clustering of implanted cobalt into nanoparticles, followed by fragmentation and regrowth at temperatures above 850 °C. Complementary ex situ heating studies using time-of-flight secondary ion mass spectrometry (ToF-SIMS) and TEM confirmed these observations at both atomic and microscales. These investigations provide a holistic view of cobalt migration at elevated temperatures, with no significant surface dicusion observed at 750 - 850 °C, and a substantial loss of Co at 1100 °C. These findings demonstrate that ion mobility varies with graphite microstructure, underscoring its key role in radionuclide retention, and establish a novel methodology for studying ion migration in graphite.

Cobalt-implanted graphite was utilised to investigate an electrochemical decontamination process using deep eutectic solvents (choline chloride–urea). A mechanistic understanding of the treatment process was achieved utilising the characterisation framework developed in the previous studies. Systematic testing on cobalt-implanted graphite revealed that cycles of intercalation and deintercalation induce lattice strain, leading to surface exfoliation as the dominant decontamination pathway. Application of this method to irradiated Magnox graphite demonstrated rapid removal of 20 - 30% of 60Co, with higher efficiencies for other fission products and no measurable weight loss or material degradation. This represents the first mechanistic study of graphite electrochemical treatment, providing key insights for the development of this waste-management route, which enables graphite waste recycling.

Together, these studies provide an integrated understanding of the evolution of nuclear graphite, from radiation-induced damage recovery to ion mobility and ultimate decontamination, supporting the development of more regenerative and efficient applications in advanced nuclear systems.