Enhanced Radiation-tolerant Oxide Dispersion Strengthened Steel and its Microstructure Evolution under Helium-implantation and Heavy-ion Irradiation

Abstract

The world eagerly needs cleanly-generated electricity in the future. Fusion reactor is one of the most ideal energy resources to defeat the environmental degradation caused by the consumption of traditional fossil energy. To meet the design requirements of fusion reactor, the development of the structural materials which can sustain the elevated temperature, high helium concentration and extreme radiation environments is the biggest challenge for the entire material society. Oxide dispersion strengthened steel is one of the most popular candidate materials for the first wall/blanket applications in fusion reactor. In this paper, we evaluate the radiation tolerance of a 9Cr ODS steel developed in China. Compared with Ferritic/Martensitic steel, this ODS steel demonstrated a significantly higher swelling resistance under ion irradiation at 460 °C to 188 displacements per atom. The role of oxides and grain boundaries on void swelling has been explored. The results indicated that the distribution of higher density and finer size of nano oxides will lead a better swelling resistance for ODS alloy. The original pyrochlore-structured Y2Ti2O7 particles dissolved gradually while fine Y-Ti-O nano clusters reprecipitated in the matrix during irradiation. The enhanced radiation tolerance is attributed to the reduced oxide size and the increased oxide density.

α′ formation kinetics and radiation induced segregation in neutron irradiated 14YWT nanostructured ferritic alloys

“…Lu et al . 53,54 have reported that NO distribution in fine and coarse grains of spark plasma sintered 9Cr ODS alloys having bimodal grain size distribution varies considerably, similar to the present study (see Figs 1 and 3). On the other hand, this is the first study reporting the α′ formation kinetics based on the NO distribution in ODS alloys, to the best of authors’ knowledge.…”

Amorphous intergranular films mitigate radiation damage in nanocrystalline Cu-Zr

“…nm and 5 nm when found at the grain boundary and in the grain interior, respectively, according to prior measurements on these samples from Grigorian and Rupert [62]. The inset in Figure 1 of ZrC precipitates could impact the radiation damage tolerance of the nanocrystalline Cu-Zr samples since precipitates can both limit grain boundary mobility through pinning [39,40] and serve as defect sinks [18,69]. However, both the ordered grain boundary and AIF-containing samples were sourced from the same ball milled powder and contained the same size and density of ZrC precipitates.…”

“…Interfaces in the form of phase boundaries have been observed to improve radiation tolerance in oxide dispersion strengthened (ODS) steels by serving as efficient trapping sites, which enhance defect recombination and He dispersion [17][18][19]. Interfaces in the form of grain boundaries are thought to serve as defect sinks by absorbing interstitials, and have been observed to facilitate defect recombination through emission of the stored interstitials to vacancies, annihilating the defect pairs [10,20].…”

Effect of Helium on Dispersoid Evolution under Self-Ion Irradiation in A Dual-Phase 12Cr Oxide-Dispersion-Strengthened Alloy

“…However, to deploy this alloy into a fast reactor environment, the effect of helium addition produced by (n, α) transmutation and subsequent He embrittlement needs to be further studied [14]. There have been a few He implantation studies on various ODS alloys [11,15,16,17,18]. Yamamoto et al observed small-size He bubbles trapped on YTiO clusters during simultaneous neutron and He irradiation of MA957 [15].…”