We have assessed the local solute redistribution at defect sinks in 20Cr–25Ni Nb-stabilised austenitic stainless steel after proton irradiation at three temperatures, i.e. 420, 460 and 500 °C, up to a maximum damage level of 0.8 dpa. This material is currently being used as cladding in Advanced Gas-cooled Reactors (AGR), and potential local Cr depletions would compromise its resistance to intergranular corrosion attack during wet storage of spent fuel elements. Irradiation induces the depletion of Cr, Fe and, to a lesser extent, Mn from grain boundaries, whereas Ni and Si become enriched at those locations. The elemental profiles are symmetric and primarily W-shaped at 420 °C, whereas at higher temperatures asymmetric and double-peaked profiles are also detected, most likely as a result of grain boundary migration. High-angle grain boundaries with a misorientation angle ≥40° become mobile at 460 °C and especially at 500 °C, and also experience a relatively large solute redistribution, with local Cr contents in a significant number of boundaries falling below 12 wt% and profile widths ≥100 nm. However, coincidence site lattice boundaries (CSL) Σ3 boundaries prove to be resistant to Cr depletion and to boundary mobility. Local elemental patterns at radiation-induced dislocations seem to mimic those at grain boundaries, but do not trigger the formation of Ni 3 Si precipitates. Additionally, Ni and Si form a shell-like structure around the pre-existing Nb(C,N) precipitates, potentially leading to the transition into G-phase at higher damage levels.