Microstructure and Mechanical Properties of Forged High Yttria 18Cr-ODS Steels
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JMEPEG https://doi.org/10.1007/s11665-020-05106-z
Microstructure and Mechanical Properties of Forged High Yttria 18Cr-ODS Steels Ratnakar Singh, Ujjwal Prakash, Deepak Kumar, and Kinkar Laha (Submitted April 6, 2020; in revised form August 14, 2020) Oxide dispersion strengthened (ODS) ferritic steels are candidate materials for clad tubes in the upcoming Generation IV nuclear reactors. In the present work, a powder forging consolidation technique has been used for fabrication of ODS steels. Two alloys having nominal compositions (in weight %) of Fe-18Cr-2W0.285Ti-0.5Y2O3 and Fe-18Cr-2W-0.571Ti-1Y2O3, respectively, have been studied in this work. The alloys were prepared by mechanical alloying of elemental powders with yttria in a Simoloyer high energy horizontal attritor. The milled powders were consolidated at 1473 K by powder forging in a flowing hydrogen gas atmosphere. Yttria to titanium ratio was kept constant at 1.75 for both the alloys. TEM micrographs of the forged alloys showed fine recrystallized grains with a dispersion of nano-size Y-Ti-O oxide particles. High-resolution transmission electron microscope fringes and the corresponding fast Fourier transformation confirmed the presence of orthorhombic Y2TiO5 oxide particles in a ferrite matrix. These were the predominant oxide particles in the forged alloys. The Y2TiO5 particles were incoherent with the matrix and exhibited a cuboidal morphology. Despite their high yttria content, both the alloys showed high tensile strength and ductility at room temperature and 973 K. Reasons for this are discussed. Keywords
ferritic ODS steels, HR-TEM, nano-sized oxide particles, powder forging, STEM, tensile properties
1. Introduction Generation IV advanced nuclear reactors for high energy production are being developed to achieve minimum wastage and a longer reactor life (Ref 1). These reactors may operate at high temperatures (773-1273 K) and intense neutron flux of up to 200 dpa (displacement per atom) (Ref 2). The effective operation of advanced reactors mainly relies on the performance of core structural materials that are exposed to severe neutron irradiation and high temperatures (Ref 3). Austenitic stainless steels (ASS) are being used in core material and structural parts of the present-day nuclear reactors because of their superior creep resistance (Ref 4). Void swelling restricts the use of austenitic stainless steels where the radiation dose is higher than 120 dpa (Ref 5). Hence, ferritic steels are being developed for the new generation nuclear reactors because they offer high swelling resistance (< 2% swelling up to 250 dpa) compared to austenitic stainless steels (Ref 5). Ferritic steels also exhibit relatively high thermal conductivities and lower thermal expansion coefficients (Ref 6). The main drawback of ferritic steels is that they exhibit inferior long-term creep Ratnakar Singh and Ujjwal Prakash, Department of Metallurgical and Materials Engineering, Indian Institute of Technology (IIT) Roorkee, Roorkee, Uttarakhand 247667, India; Deepak Kumar, Dep
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