Reduced-activation ferritic steels are considered as the primary candidate materials for structural applications within nuclear fusion power plants. It is known that by mechanically alloying ferritic steel powder with Y (usually in the form of Y₂O₃) then consolidating the material by hot isostatic pressing, a nanoscale dispersion of oxygen rich nanoclusters as small as ~2nm is introduced into the microstructure. This vastly improves high temperature strength and creep resistance, and the nanoclusters also act as trapping sites for helium and point defects produced under irradiation. In this thesis, the evolution of the oxide nanoclusters in a Fe-14Cr-2W-0.3Ti & 0.3Y₂O₃ ODS alloy was investigated primarily using atom probe tomography. The microstructure was characterised at various points during processing to give an insight into the factors influencing the formation of the nanoclusters. It was found that the nanoclusters nucleated during the mechanical alloying stage, then followed near classical nucleation and growth mechanisms keeping the same composition of ~8%Y, ~12%Ti,~25%O and ~45%Cr throughout. The formation and evolution of 5-15nm grain boundary oxides was also observed, and these were shown to form first as Cr₂O₃ particles that subsequently transform into a Y-Ti-O based oxide on further processing. The influence of mechanical alloying with 0.5wt.%Fe₂Y rather than 0.3wt.%Y₂O₃ was also investigated, and this showed that there was no difference in the final microstructure produced provided the level of Ti in the starting powder was tightly controlled. Without sufficient Ti, the nanoclusters were Y-O based and ~6nm diameter. Both the Y-O and Y-Ti-O nanoclusters were moderately stable on annealing at 1200°C for up to 100 hours, with only minimal coarsening observed. Ti was found not to influence the coarsening rate of the nanoclusters significantly. The stability of the oxide nanoclusters under irradiation was investigated by using Fe²⁺ ion irradiation to simulate displacement cascade damage in the ODS-Eurofer material (the official European candidate material for testing in the ITER fusion test reactor). Doses up to ~6 dpa at 400°C were used, and there was no significant change to the nanocluster distribution. However segregation of Mn to dislocations was observed after irradiation. These results indicate that ODS steels are good candidate structural materials, as the microstructure is stable at high temperature and under irradiation. The starting powders, and processing parameters need to be tightly controlled in order to produce the optimal material for use in service.
| Identifer | oai:union.ndltd.org:bl.uk/oai:ethos.bl.uk:559835 |
| Date | January 2012 |
| Creators | Williams, Ceri Ann |
| Contributors | Smith, George D. W. ; Roberts, Steve ; Marquis, Emmanuelle |
| Publisher | University of Oxford |
| Source Sets | Ethos UK |
| Detected Language | English |
| Type | Electronic Thesis or Dissertation |
| Source | http://ora.ox.ac.uk/objects/uuid:4f03864f-4fe1-4005-ac28-6d9e8244989b |
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