Abstract
Microstructure and nanohardness evolution in 18Cr10NiTi and 18Cr10NiTi-ODS steels after exposure to argon ion irradiation has been studied by combination of nanoindentation tests, XRD analysis, TEM and SEM observation. ODS-modified alloy was produced on the basis of conventional 18Cr10NiTi austenitic steel by mechanical alloying of steel powder with Y(Zr)-nanooxides followed by mechanical-thermal treatment. XRD analysis has showed no significant changes in the structure of 18Cr10NiTi steel after irradiation at room and elevated temperatures (873 K) and in ODS-steel after irradiation at 873 K, whereas the evidences of domains refinement and microstrain appearance were revealed after irradiation of 18Cr10NiTi-ODS steel at room temperature (RT). Layer-by-layer TEM analysis was performed to investigate the microstructure of alloys along the damage profile. The higher displacement per atom (dpa) and Ar concentration clearly lead to increased cavities size and their number density in both steels. The swelling was estimated to be almost half for 18Cr10NiT-ODS (4.8%) compared to 18Cr10NiTi (9.4%) indicating improved swelling resistance of ODS-steel. The role of oxide/matrix interface as a sink for radiation-induced point defects and inert gas atoms is discussed. The fine dispersed oxide particles are considered as effective factor in suppressing of cavity coarsening and limiting defect clusters to small size. The hardness behavior was investigated in both non-irradiated and irradiated specimens and compared to those at RT and elevated temperature of irradiation. The hardness increase of unirradiated ODS-steel is associated mainly with grain refinement and yttrium oxides particles addition. The hardening of 18Cr10NiTi-ODS after Ar ion irradiation at RT was found to be much lower than 18Cr10NiTi. Black dots and dislocation loops are observed for both steels in the near-surface area; however, the main hardening effect is caused by the cavities. Oxide dispersion strengthened steel was found to be less susceptible to radiation hardening/embrittlement compared with a conventional austenitic steel.
Highlights
Igor Kolodiya,*, Oleksandr Kalchenkoa, Sergiy Karpova, Victor Voyevodina,b, Mikhail Tikhonovskya, Oleksii Velikodnyia, Galyna Tolmachovaa, Ruslan Vasilenkoa, Galyna Tolstolutskaa
The swelling was estimated to be almost half for 18Cr10NiT-oxide dispersion-strengthened (ODS) (4.8%) compared to 18Cr10NiTi (9.4%) indicating improved swelling resistance of ODS-steel
By the collection of properties, oxide dispersion strengthened 18Cr10NiTi austenitic stainless steel is less susceptible to radiation hardening/embrittlement compared to conventional austenitic steel
Summary
Igor Kolodiya,*, Oleksandr Kalchenkoa, Sergiy Karpova, Victor Voyevodina,b, Mikhail Tikhonovskya, Oleksii Velikodnyia, Galyna Tolmachovaa, Ruslan Vasilenkoa, Galyna Tolstolutskaa. Exposure of materials under irradiation in nuclear reactors leads to changes in the crystalline structure at the atomic level due to the nucleation of various kinds of defects (such as voids, bubbles, dislocation loops and stacking faults). These defects can dramatically degrade the physical properties of reactor materials through swelling, irradiation hardening, embrittlement, irradiation creep, etc. Oxide dispersion-strengthened (ODS) austenitic stainless steels can be an attractive material because of their corrosion resistance, high-temperature strength and irradiation properties [2]. Studies mainly focused on the macroscopic mechanical properties of steels after irradiation, but studies on crucial aspects such as microstructural evolution, the effect of inert gas implantation, and the interaction with different microstructural characteristics are rare
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