23 research outputs found
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
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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