Effects of Ion Irradiation on the Microstructure of an ODS/Fe12Cr Alloy
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Effects of Ion Irradiation on the Microstructure of an ODS/Fe12Cr Alloy V. de Castro1, S. Lozano-Perez1, M. L. Jenkins1 1 Department of Materials, University of Oxford, 16 Parks Road, OX1 3PH, Oxford, UK. ABSTRACT This study describes the microstructures of an ODS/Fe12Cr alloy before and after Fe+ irradiation at 500ºC to a dose of 2.5 dpa and compares these microstructures with those of a reference Fe12Cr alloy produced following the same pulvimetallurgical route. The grain and dislocation structures of the alloys did not seem to change after irradiation. The main difference after irradiation was the appearance of small loops, which had similar sizes, distribution and densities in the ODS and reference alloys. The size distributions and chemical compositions of the ODS particles were similar to those found before irradiation.
INTRODUCTION Oxide dispersion strengthened (ODS) reduced-activation ferritic/martensitic steels (RAFMS) are among the most promising structural materials for future fusion reactors. The homogeneous dispersion in the steel matrix of hard nanosized oxide particles, such as Y2O3, should allow an increase in the maximum service temperature of these materials by more than 100ºC as compared with conventional RAFMS. It has been demonstrated that ODS-RAFMS have better tensile and creep properties than their unreinforced counterparts [1]. However, they exhibit a high ductile-brittle transition temperature (DBTT), which seems to be related to the processing route [2]. ODS steels are normally produced by mechanical alloying (MA) and consolidated by hot isostatic pressing (HIP) or hot extrusion, but these production routes lead to very complex microstructures. MA is necessary in order to disperse the starting oxide nanopowders homogeneously, but leads to an excess of oxygen, micron-sized carbide precipitation and residual porosity. Despite these detrimental effects, recent studies have shown that the DBTT can be shifted towards lower temperatures by applying suitable thermomechanical treatments [3]. If ODS-RAFMS are to be used in fusion reactors it is indispensable to know how these materials behave under irradiation, which is directly linked to irradiation-induced changes in their microstructure. The 14 MeV neutrons resulting from the fusion reaction will cause a substantial amount of radiation damage as well as the production of He and H transmutation gases. This may result in changes in the dislocation structures, radiation-induced precipitation and solute segregation, and bubble formation. The oxide particles present in ODS-RAFMS could act as sinks for irradiation induced point defects and He, so lowering the rate of damage accumulation in the material and inhibiting the growth of bubbles and migration of He to grain boundaries [4]. However, the particles could undergo dissolution or coarsening and chemical or crystallographic changes as a consequence of irradiation. These changes could alter their effectiveness as trapping sites and affect the mechanical behaviour of the material. In the
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