Micro mechanical testing of candidate structural alloys for Gen-IV nuclear reactors

Ion irradiation is often used to simulate the effects of neutron irradiation due to reduced activation of materials and vastly increased dose rates. However, the low penetration depth of ions requires the development of smallscale mechanical testing techniques, such as nanoindentation and microcompression, in order to measure mechanical properties of the irradiated material. In this study, several candidate structural alloys for Gen-IV reactors (800H, T91, nanocrystalline T91 and 14YWT) were irradiated with 70 MeV Fe9+ ions at 452 °C to an average damage of 20.68 dpa. Both the nanoindentation and microcompression techniques revealed significant irradiation hardening and an increase in yield stress after irradiation in austenitic 800H and ferritic-martensitic T91 alloys. Ion irradiation was observed to have minimal effect on the mechanical properties of nanocrystalline T91 and oxide dispersion strengthened 14YWT. These observations are further supported by line broadening analysis of X-ray diffraction measurements, which show a significantly smaller increase in dislocation density in the 14YWT and nanocrystalline T91 alloys after irradiation. In addition, good agreement was observed between cross-sectional nanoindentation and the damage profile from SRIM calculations

Nonsingular dislocation and crack fields: Implications to small volumes

The simplest, yet robust, gradient elasticity theory (GRADELA) as first introduced by the last author is used to deduce nonsingular expressions for the stress and strain fields near dislocation lines and crack tips. These expressions are particularly useful for small volumes where the details of the deformation field need to be known for interpreting related experimental observations. Various implications are discussed in relation to the determination of the size of dislocation cores, the size of maximum stress or maximum strain in crack tips, and the interpretation of X-ray line profile measurements in determining internal stresses. © 2008 Springer-Verlag

Evolution of dislocation structure in neutron irradiated Zircaloy-2 studied by synchrotron x-ray diffraction peak profile analysis

Dislocation structures in neutron irradiated Zircaloy-2 fuel cladding and channel material have been characterized by means of high-resolution synchrotron x-ray diffraction combined with whole peak profile analysis and by transmission electron microscopy (TEM). The samples available for this characterization were taken from high burnup fuel assemblies and offer insight into the evolution of the dislocation structure after the formation of dislocation loops containing a c component. Absolute dislocation density values are about 4e15 times higher for the whole peak profile compared to TEM analysis. Most interestingly, the diffraction analysis suggests that the total dislocation density, as well as the a loop density, increases with fluence for the cladding material type. This trend is also inferred from a Williamson-Hall representation but contradicts the TEM observations. The c loop density evolution is more complicated and doesn't display any particular trend. In addition, the diffraction analysis highlights the presence of well-developed shoulders adjacent to the basal reflections and noticeable peak asymmetry particularly for the channel samples that experienced slightly lower operation temperatures than the clad. The findings are discussed in respect of the perceived irradiation induced growth mechanisms in Zr alloys

Micro mechanical testing of candidate structural alloys for Gen-IV nuclear reactors

Ion irradiation is often used to simulate the effects of neutron irradiation due to reduced activation of materials and vastly increased dose rates. However, the low penetration depth of ions requires the development of small-scale mechanical testing techniques, such as nanoindentation and microcompression, in order to measure mechanical properties of the irradiated material. In this study, several candidate structural alloys for Gen-IV reactors (800H, T91, nanocrystalline T91 and 14YWT) were irradiated with 70 MeV Fe9+ ions at 452 °C to an average damage of 20.68 dpa. Both the nanoindentation and microcompression techniques revealed significant irradiation hardening and an increase in yield stress after irradiation in austenitic 800H and ferritic-martensitic T91 alloys. Ion irradiation was observed to have minimal effect on the mechanical properties of nanocrystalline T91 and oxide dispersion strengthened 14YWT. These observations are further supported by line broadening analysis of X-ray diffraction measurements, which show a significantly smaller increase in dislocation density in the 14YWT and nanocrystalline T91 alloys after irradiation. In addition, good agreement was observed between cross-sectional nanoindentation and the damage profile from SRIM calculations

Micro mechanical testing of candidate structural alloys for Gen-IV nuclear reactors

Ion irradiation is often used to simulate the effects of neutron irradiation due to reduced activation of materials and vastly increased dose rates. However, the low penetration depth of ions requires the development of small-scale mechanical testing techniques, such as nanoindentation and microcompression, in order to measure mechanical properties of the irradiated material. In this study, several candidate structural alloys for Gen-IV reactors (800H, T91, nanocrystalline T91 and 14YWT) were irradiated with 70 MeV Fe9+ ions at 452 °C to an average damage of 20.68 dpa. Both the nanoindentation and microcompression techniques revealed significant irradiation hardening and an increase in yield stress after irradiation in austenitic 800H and ferritic-martensitic T91 alloys. Ion irradiation was observed to have minimal effect on the mechanical properties of nanocrystalline T91 and oxide dispersion strengthened 14YWT. These observations are further supported by line broadening analysis of X-ray diffraction measurements, which show a significantly smaller increase in dislocation density in the 14YWT and nanocrystalline T91 alloys after irradiation. In addition, good agreement was observed between cross-sectional nanoindentation and the damage profile from SRIM calculations