Towards High Performance Oxide-Dispersion-Strengthened Alloys
- Department of Nuclear Engineering, Texas A and M University, College Station, TX 77843 (United States)
- Department of Material Science and Engineering, Hokkaido University: N13, W-8, Kita-ku, Sapporo 060-0808 (Japan)
The current first-generation candidate swelling-resistant alloys envisioned to replace austenitics are ferritic and ferritic-martensitic steels, with second-generation alloys being yttria dispersion-hardened variants of these steels. In ferritic-martensitic alloys, ferrite grains always swell earlier than martensite. Currently, most oxide dispersion-strengthened (ODS) alloys have only a ferrite matrix, thus giving up the first line of defense against swelling provided by tempered martensite. Towards development of better swelling-resistant ODS alloys, dual phase (ferrite + martensite) ODS alloys with oxide particles uniformly distributed in both ferrite and martensite phases is needed. Furthermore, fundamental understanding on mechanisms governing dispersoid stability and its correlation with void swelling is greatly needed. We reported here further alloy improvement by exploring new 9Cr ODS alloy with majority of oxide dispersoids uniformly distributed in tempered martensite grains. The alloy was irradiated by 3.5 MeV Fe self ions up to 600 dpa (displacements per atom) and at temperatures ranging from 325 to 625 deg. C. According to SRIM code, the dpa profile peaks at about 1000 nm. First of all, ion irradiation reduces the dispersoid sizes. The dispersoid sizes begin to increase beyond ∼1000 nm. Second, higher irradiation temperatures result in larger dispersoids on average. Third, there is no significant dispersoid size difference between 100 and 200 dpa irradiation. The equilibrium sizes are quickly saturated. Size and coherency of a dispersoid is closely related. In martensite grains, fine dispersoids tend to be coherent while coarse dispersoids tend to incoherent. We found that at a low irradiation temperature, coherent small dispersoids become smaller and their interface coherency remains. In a comparison, incoherent large dispersoids disappear after irradiation. At an intermediate irradiation temperature, the final equilibrium dispersoid sizes increase. At a high irradiation temperature, the final dispersoid sizes further increase and some large semi-coherent dispersoids survive. In ferrite grains, as-received dispersoids are relatively smaller in comparison with martensite grains. Their equilibrium size changes after irradiation follow a similar trend but large dispersoids completely disappear even under high temperature ion irradiation. We propose a model to explain the above dispersoid size changes under various irradiation conditions. The model considers the competing effects between ballistic-collision- assisted dispersoid dissolution into matrix and thermally driven back-diffusion for recovery. When the competing reaches a balance, an equilibrium size distribution curve is determined. For dispersoids smaller than the equilibrium size, they grow upon ion irradiation. Large dispersoids may shrink and reach the equilibrium size but ultra-large ones will completely disappear if they shrinkage paths are far away from the curve. The model further considers the complexity from the dpa rates, and damage cascade morphology.
- OSTI ID:
- 22992139
- Journal Information:
- Transactions of the American Nuclear Society, Vol. 114, Issue 1; Conference: Annual Meeting of the American Nuclear Society. Embedded topical meeting 'Nuclear fuels and structural material for the next generation nuclear reactors', New Orleans, LA (United States), 12-16 Jun 2016; Other Information: Country of input: France; 3 refs.; Available from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 United States; ISSN 0003-018X
- Country of Publication:
- United States
- Language:
- English
Similar Records
Oxide dispersoid coherency of a ferritic-martensitic 12Cr oxide-dispersion-strengthened alloy under self-ion irradiation
Influence of thermomechanical treatment on irradiation microstructure in an ODS ferritic steel