An Internal State Variable Elastoviscoplasticity-Damage Model for Irradiated Metals

This study presents an irradiation-dependent internal state variable (ISV) elastoviscoplasticity-damage constitutive model that accounts for nuclear irradiation hardening and embrittlement of the irradiated polycrystalline materials. The irradiation effects were added to the coupled plasticity-damage kinetics with consideration of the structure–property relationships. The present irradiation-dependent elastoviscoplasticity-damage model was compared with the lab deformation experimental data of irradiated oxygen-free high conductivity (OFHC) copper, modified 9Cr-1Mo steel, and Ti-5Al-2.5Sn. The results show excellent agreement over the entire stress–strain curves at various irradiation doses. Because the ISV model, before the irradiation plasticity-damage addition, had been used on over 80 different metal alloys, it is anticipated that this nuclear irradiation supplement will also allow for application to many more irradiated metal alloys.

Correlation between Microstructural Change and Irradiation Hardening Behavior of He-Irradiated V–Cr–Ti Alloys with Low Ti Addition

V–4Cr–xTi (x = 0 to 4) alloys were used to investigate the additional effect of Cr, Ti and interstitial impurities on the microstructural evolution in He-irradiated V–Cr–Ti alloys to minimize radioactivity after fusion neutron irradiation. Transmission electron microscopy and atom probe tomography were carried out to the He-irradiated specimens at 500 °C with 0.5 dpa at peak damage. A flash electro-polishing method for the FIB-extracted specimen was established for the ion-irradiated vanadium alloys. The microstructural evolution of the irradiation-induced titanium-oxycarbonitride, Ti(CON) precipitates was observed and was influenced by the effect of Ti addition on the Ti(CON) precipitation. Apparent Ti(CON) precipitates formed in V-4Cr-xTi with 2% addition of Ti. In the V-4Cr-1Ti alloy, a high density Ti enriched cluster was formed. The origin of the irradiation hardening increase resulted from the size distribution of Ti(CON) precipitation from the dispersed barrier-hardening theory.

Nanoindentation Test of F321 Austenitic Stainless Steel Under Fe-ion Irradiation

Nuclear technology, as a high quality, clean and reliable energy supply, is attracting broad interest from countries across the world. F321 austenitic stainless steel (F321SS) is widely utilized in key components of nuclear power plant due to its excellent corrosion resistance and high temperature mechanical properties. Irradiation can easily lead to the degradation behaviors of materials, such as irradiation hardening, irradiation embrittlement and high-temperature He embrittlement, etc. Understanding such degradation is important for predicting the evolution of material behavior under irradiation and extending the lifespan of existing nuclear reactors. Ion irradiation is most commonly used to model neutron-induced damage since the irradiation conditions (temperature, flux, spectrum, etc.) can be regulated more accurately and flexibly. In this paper, the Fe-ion irradiation experiments of F321SS at different temperatures and doses were carried out, and the nanoindentation experiments under different conditions were further conducted. Irradiation hardening is observed in all specimens and strongly depending on irradiation temperature and damage dose. The hardness after irradiating increases with doses and saturates for at least 1dpa under low temperature regimes (< 300°C). However, at higher temperature (450°C and 560°C), nano-hardness reaches the peak at ∼0.5dpa and then declines. Moreover, the hardness of all specimens has a similar trend with temperature, that is, it first increases, reaches the peak, and then decreases.