Dislocation-Radiation Obstacle Interactions: Developing Improved Mechanical Property Constitutive Models Page: 3 of 23
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I. Executive Summary
Radiation damage to structural and cladding materials, including austenitic stainless
steels, ferritic steels, and zirconium alloys, in nuclear reactor environments results in significant
mechanical property degradation, including yield strength increases, severe ductility losses and
flow localization, which impacts reliability and performance. Generation IV and advanced fuel
cycle concepts under consideration will require the development of advanced structural
materials, which will operate in increasingly hostile environments. The development of
predictive models is required to assess the performance and response of materials in extreme Gen
IV reactor operating conditions (temperature, stress, and pressure), to decrease the time to
rapidly assess the properties of new materials (e.g., new alloys, nanostructured materials) and
insert them into technological applications (Gen IV and Advanced Fuel Cycle Operations). In
this project we focus on identifying the controlling mechanisms of dislocation interactions with a
range of obstacles commonly produced in material microstructures under irradiation (stacking fault
tetrahedron and dislocation loops, and radiation-induced precipitates) through a combination of
atomistic modeling and dynamic in situ and static ex-situ transmission electron microscopy
experiments. Molecular dynamics was used to determine time evolution dynamics. Correlation of
the simulation and experimental results provide information of the governing dynamics that are
inaccessible to either approach alone, but that dominate the macroscopic mechanical response. The
material systems investigated focused on copper and iron-based alloys, representative of the most-
promising austenitic and ferritic steels for advanced reactor applications, although the findings are
likely to be generic to many material systems.
The results are able to determine the obstacle strengths, including the observation of new
interaction/strengthening mechanisms, and provide insight into the localized deformation in the
form of defect free channels observed in irradiated structural materials. Molecular dynamics
simulations of the interaction between edge, screw or mixed dislocations and a SFT show that
the SFT is a strong barrier to dislocation motion. Further, these observations lead to the
conclusion that dislocation channel formation is much more complicated than defect absorption
in a single interaction, and may result from a combination of i) decreased defect cluster size due
to shear, ii) partial absorption leaving isolated vacancies and smaller defect clusters of
presumably reduced obstacle resistance, and iii) partial to complete absorption and the
subsequent dragging and re-emission of defect clusters at a different location. The study of
dislocation - precipitate interactions has revealed a new dislocation interaction mechanism with
coherent precipitates that can explain the strong irradiation hardening observed in reactor
pressure vessel steels. Ultimately, the dislocation-radiation obstacle interaction rules will be
incorporated in higher length scale models to predict the post-yield constitutive properties of
irradiated materials required for the design of advanced material systems for advanced Nuclear
technology.
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WIrth, B.D. & Robertson, Ian M. Dislocation-Radiation Obstacle Interactions: Developing Improved Mechanical Property Constitutive Models, report, November 29, 2007; United States. (https://digital.library.unt.edu/ark:/67531/metadc895040/m1/3/: accessed April 23, 2024), University of North Texas Libraries, UNT Digital Library, https://digital.library.unt.edu; crediting UNT Libraries Government Documents Department.