Description
This thesis aims to explore CuCrZr alloys for use in the next generation of fusion reactors, such as ITER, and those that come after. Numerous challenges are presented by the inclusion of CuCrZr in fusion reactors, both in terms of its manufacture and the unique challenges materials face in the high energy fusion radiation environment. Study will be focused on two key aspects: the evolution of precipitates during aging, with a focus on the role of Zr in optimising alloy performance, and the microstructural response to irradiation and the associated changes in properties. A study into precipitation evolution in CuCrZr at 480 °C was carried out using positron annihilation spectroscopy and microhardness techniques to measure bulk changes in precipitate structure and the associated mechanical properties. This was supported with 3D atom probe, SEM, TEM, and STEM techniques to determine changes in crystal structure and precipitate morphology and density. The study focused on the early stages of aging as it was found that this is when the most rapid changes occur, however testing of overaged alloy tested aging times up to 24 hours, well beyond the recommended heat treatment times for ITER-grade CuCrZr. A link was found between positron lifetime, hardness, and precipitate structure, showing that Zr segregates at the matrix-precipitate boundaries. While previous studies have shown this to occur at longer aging times at elevated temperatures, this study confirmed that the phenomenon occurs even in the earliest stages of aging at 480 °C. Transient grating spectroscopy was used to characterise changes in thermal diffusivity in CuCrZr during the aging process and proton irradiation. Unirradiated solution annealed and prime aged material was characterised, as were samples irradiated with 2MeV protons to between 0.01dpa to 0.1dpa. The conductive properties of CuCrZr under irradiation were found to be highly resistant to change in this damage range, however thermal diffusivity was found to be slightly lower (around 1m^2/s) than in the unirradiated prime aged material (around 1. 2m^2/s). STEM characterisation found there to be no change in the precipitate structure across this range, however, showing the TGS results to likely be accurate as the dislocation loops formed had no effect on the alloy's thermal properties. TGS shows promise, however as a method for monitoring changes in precipitate structure due to thermal annealing. Dislocation loops were found to form under irradiation with 2MeV protons at 200 °C between 0.05 and 0.075dpa. 200 °C. Samples were irradiated from 0.01 to 0.6dpa, and characterised using STEM-EDX to monitor changes in precipitate structure, with bright field STEM techniques to characterise the formation and evolution of type dislocation loops. Radiation damage defects were found to nucleate as black spot damage at precipitate sites before forming dislocation loops above 0.05dpa. Loops were found to be pinned by the Cr-rich precipitates in CuCrZr, agglomerating over time, increasing in diameter while reducing in density.| Period | 31 Dec 2020 |
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| Examinee | Jonathan Hughes |
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| Degree of Recognition | International |