Development of Oxide Dispersion Strengthened (ODS) Ferritic Steel Through Powder Forging
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JMEPEG DOI: 10.1007/s11665-017-2573-2
Development of Oxide Dispersion Strengthened (ODS) Ferritic Steel Through Powder Forging Deepak Kumar, Ujjwal Prakash, Vikram V. Dabhade, K. Laha, and T. Sakthivel (Submitted July 12, 2016; in revised form December 21, 2016) Oxide dispersion strengthened (ODS) ferritic steels are candidates for cladding tubes in fast breeder nuclear reactors. In this study, an 18%Cr ODS ferritic steel was prepared through powder forging route. Elemental powders with a nominal composition of Fe-18Cr-2 W-0.2Ti (composition in wt.%) with 0 and 0.35% yttria were prepared by mechanical alloying in a Simoloyer attritor under argon atmosphere. The alloyed powders were heated in a mild steel can to 1473 K under flowing hydrogen atmosphere. The can was then hot forged. Steps of sealing, degassing and evacuation are eliminated by using powder forging. Heating ODS powder in hydrogen atmosphere ensures good bonding between alloy powders. A dense ODS alloy with an attractive combination of strength and ductility was obtained after re-forging. On testing at 973 K, a loss in ductility was observed in yttria-containing alloy. The strength and ductility increased with increase in strain rate at 973 K. Reasons for this are discussed. The ODS alloy exhibited a recrystallized microstructure which is difficult to achieve by extrusion. No prior particle boundaries were observed after forging. The forged compacts exhibited isotropic mechanical properties. It is suggested that powder forging may offer several advantages over the traditional extrusion/HIP routes for fabrication of ODS alloys. Keywords
forging, isotropy, mechanical alloying, mechanical properties, ODS steel, powder metallurgy
1. Introduction Fast breeder generation IV nuclear reactors are being developed for higher energy production with minimum wastage, safety, reliability and useful reactor life (Ref 1). These reactors have higher operating temperatures. Future reactor structural materials and cladding materials should have adequate strength, ductility, creep rupture, fatigue, toughness and good swelling resistance under irradiation (Ref 1). Conventional reactor materials cannot meet these demanding requirements. Initially austenitic stainless steels were used as reactor material due to their high creep resistance. But, austenitic steels undergo unacceptable swelling when subjected to high (>50 dpa) radiation dose (Ref 2). Ferritic-martensitic (F/M) steels exhibit good thermal conductivity, lower thermal expansion and good resistance to void swelling under irradiation in comparison with austenitic stainless steel (Ref 3). To take the advantage of F/M steels, advanced F/M steels called reduced activation ferritic-martensitic (RAFM) steels have been developed. They exhibit quick decay of induced radioactivity and allow shallow burial of components after use (Ref 1). Like F/M steels, they also exhibit good swelling resistance than austenitic steels. However, these steels suffer from poor creep resistance around 873 K (Ref 4), and this limits their use a
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