The Role of Nb and Impurities on Nano-oxide Retention under Neutron Irradiation

A range of Fe-based alloys using various compositions and processing techniques are currently being considered for both accident tolerant fuel (ATF) applications in currently operating commercial light water reactors (LWRs), advanced reactor and small modular reactor (SMR) applications. Any of these reactor environments require materials, which can perform under extreme environments including elevated temperatures, high displacement damage and corrosive conditions. Oxide dispersion strengthened (ODS) alloys form nano-clusters and nano-oxides within the microstructure that provide elevated temperature creep strength and radiation tolerance. However, large fluctuations in impurity content during processing have led to inconsistent properties and significant heat to heat variations. The objective of this work is to evaluate the effectiveness of impurity sequestration under neutron irradiation conditions relevant to current and advanced reactors using detailed post irradiation examination. Under irradiation, nano-oxide retention is a balance between ballistic dissolution, back diffusion of solutes to the original oxide, and diffusion of solutes from one oxide through the lattice to another. We hypothesize that the addition of a more stable carbonitride former, such as Nb, will reduce the amount of Ti loss from (Y,Ti,O) nano-oxides through the reduction of Ti capture in the lattice by impurities following ballistic dissolution, thereby retaining nano-oxide density and size, and thus enhance nano-oxide effectiveness as a defect sink across all temperatures. The outcome of this work will provide quantitative analysis of the irradiated microstructure using comprehensive post-irradiation transmission electron microscopy including dislocation loops, nano-oxides, and any secondary precipitate phases as a function of temperature to compare with the as-fabricated alloys. The availability of this dataset will support ongoing development activities in determining composition windows for ODS steels that will provide both acceptable creep resistance and irradiation resistance for current reactors and advanced reactor applications. The alloys in this work were irradiated in the High Flux Isotope Reactor (HFIR) for 4 cycles (~8 dpa) at nominal temperatures of 300 °C, 385 °C, and 525 °C to represent conditions for LWR and advanced reactor designs. While five ODS variants were included in the irradiation campaign, the primary focus of this RTE will be a comparison of the “legacy” ODS steel (14YWT) with an ODS steel designed for impurity sequestration (OFRAC) through the addition of Mo, Ti, and Nb to form (Nb,Ti)-rich carbonitrides. There are 2 alloys, irradiated at 3 temperatures, and thus, in total, the proposed experiments will require an estimation of about 48 hours for lamella preparation and 72 hours for post-irradiation examination with transmission electron microscopy. This project will provide reactor relevant neutron irradiation data demonstrating a promising alloy composition strategy to improve radiation tolerance, a novel and valuable area which is of growing interest to the Department of Energy Office of Nuclear Energy, as advanced clad materials are selected and eventually manufactured for SMRs.