[en] Full text: The project concerns recovery of ion irradiation damage in nuclear graphite during annealing. Nuclear graphite is employed as a moderator material in generation IV nuclear reactors such as Molten Salt reactors, to slow fast neutrons and sustain the fissile chain reaction. Ion irradiation produces a similar cascade damage effect to neutron irradiation and is more convenient to control for experimentation. Interstitials and vacancies in crystalline graphite cause anisotropic microstructural distortions which cause shrinkage and expansion as well as electrical property changes. Annealing at high temperatures has been shown to recover these changes to some extent and is used to manage the release of potentially destructive Wigner energy in current reactor operation. The research of this project endeavours to improve the understanding of the recovery process in PCIB nuclear graphite, which is a candidate material for current generation reactor designs. With greater understanding of the recovery process and improved estimation of damage, moderator and hence reactor operational lifetime may be extended. Raman, TEM and XRD techniques were used to characterise samples that had been irradiated with Carbon ions accelerated at 35MeV at a fluence of 4.50x1017 ions/cm². Using SRIM to predict the damage profile allowed the correlation of crystallite size to dpa reflecting the highly concentrated maximum damage at 30dpa at approximately 40µm from the irradiated surface. TEM images showed the arrangement of binder matrix and rosette-shaped filler particles with pores at a range of scales from the microscopic to nano-scale Mrozowski cracks between basal planes. Qualitative evidence of increasing disorder with damage was seen in HRTEM images of filler particles. The trend in d-spacing, measured by diffraction patterns, supported this observation between 2dpa and 30dpa. Dislocations were identifiable in the IFFTs. Raman spectroscopy will also be conducted in-situ during annealing to give a time-dependent kinetic model at 200°C and a temperature dependent model for a range of temperatures up to 600°C, approximating molten salt reactor conditions. Activation energies for the expected fast and slow processes will be deduced and compared with literature values for HOPG, PCIB and other nuclear graphite grades. A Molecular Dynamics simulation is being run to independently predict activation energies and gauge the validity of assumptions regarding which defect arrangements are involved under the given conditions. Using multiple characterisation techniques the present inquiry seeks to enrich the body knowledge of nuclear graphite and clarify the microstructural mechanisms of irradiation damage recovery. (author)