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EUROFER97 is the European Reduced Activation Ferritic-Martensitic (RAFM) candidate steel to be used as a structural material in future nuclear fusion devices. Neutron irradiation will degrade the mechanical properties, setting limits in terms of operational temperature and maximum allowed dose. It is anticipated that the European DEMO will utilize a first blanket with a 20 dpa damage limit in the first-wall and then switch to a second set of blankets with a 50 dpa damage limit. In addition to mechanical properties, tritium retention is a crucial consideration for any plasma-facing as well as structural material. In general, tritium retention for EUROFER97 is expected to be small. A very recent study used 20 MeV tungsten (W) ion beam irradiation as a surrogate for the displacement damage that neutrons will cause [A. Theodorou et al., Nucl. Mater. Energy 38, 101595 (2024)]. After, W irradiation, a low-temperature deuterium (D) plasma was used to decorate the created defects. Nuclear Reaction Analysis (NRA) with $^{3}$He was then employed to measure the trapped D within and beyond the damage region. The study showed that the retention of D is substantially increased by the displacement damage. However, post-irradiation annealing at 350°C recovered all radiation-induced defects. As no data was available so far for relevant irradiation doses and temperatures we adopted the methodology of the previous study, modifying it in four essential aspects: Firstly, rather than post-irradiation annealing of a microstructure that was irradiated at room temperature, high temperature irradiation was applied. Secondly, a continuous, broad beam was applied. Thirdly, irradiations were conducted with damage doses ranging between 0.7 and 100 dpa with dose rates ranging from $10^{-5}$ to $10^{-3}$ dpa/s. Fourthly, X-ray diffraction (XRD) analysis was applied to measure the lattice and micro strain induced by the ion irradiations.
XRD analysis revealed increased tensile lattice strain within the ion range, compared to the non-irradiated reference sample for all irradiation temperatures up to 400°C. Post-irradiation annealing at 600°C resulted in defect recovery and strain relaxation. NRA gives a very similar level for D retention within the damage region as compared to previous studies involving 0.6 dpa irradiations with scanned beam. D retention returns back to the level of pristine EUROFER97 when irradiation damaging is performed at 300°C. The present experiments support the initial assumptions about irradiation-induced defect densities that were made to predict tritium loss in a DEMO first wall in [K. Schmid, Nucl. Fusion 65, 026039 (2025)].