Multiscale Modeling of Irradiation Induced Hardening in a-Fe, Fe-Cr and Fe-Ni Systems
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Multiscale Modeling of Irradiation Induced Hardening in a-Fe, Fe-Cr and Fe-Ni Systems Ioannis N. Mastorakos1, Ngoc. Le1, Melody Zeine1, Hussein M. Zbib1 and Moe Khaleel2 1 School of Mechanical and Materials Engineering, Washington State University, Pullman, WA 99164-2920, U.S.A. 2 Pacific Northwest National Laboratory, Richland, WA 99352, U.S.A. ABSTRACT Structural materials in the new Generation IV reactors will operate in harsh radiation conditions coupled with high levels of hydrogen and helium production, thus experiencing severe degradation of mechanical properties. The development of structural materials for use in such a hostile environment is predicated on understanding the underlying physical mechanisms responsible for microstructural evolution along with corresponding dimensional instabilities and mechanical property changes. As the phenomena involved are very complex and span in several length scales, a multiscale approach is necessary in order to fully understand the degradation of materials in irradiated environments. The purpose of this work is to study the behavior of Fe systems (namely a-Fe, Fe-Cr and Fe-Ni) under irradiation using both Molecular Dynamics (MD) and Dislocation Dynamics (DD) simulations. Critical information is passed from the atomistic (MD) to the microscopic scale (DD) in order to study the degradation of the material under examination. In particular, information pertaining to the dislocation-defects (such as voids, helium bubbles and prismatic loops) interactions is obtained from MD simulations. Then this information is used by DD to simulate large systems with high dislocation and defect densities. INTRODUCTION The development of new generation fission and fusion nuclear reactors depends on the availability of materials to operate safely in severe environments for an extended service lifetime. Structural materials in nuclear reactors will function in harsh radiation conditions coupled with point defects and defect clusters of high density, and thus will experience severe degradation of mechanical properties. Over the past two decades, significant advances have been made in understanding the effects of irradiation on materials microstructure and mechanical properties by focusing theory, experiments and modeling on the basic underlying physical mechanisms [1]. For example, it is well established that the effect of irradiation on ferritic/martensitic alloys at low to intermediate temperatures (T < ≈850 K) is to increase yield stress, reduce strain hardening capacity and initiate flow localization at lower strains [2]. Furthermore, the predominant microstructural features include dislocation loops, voids, regions of solute segregation and second phase precipitates. The initial density and evolution of these features depends on some key variables, such as irradiation temperature, dose and dose rate, helium production rate and alloy composition. However, the prediction of the material behavior based on this knowledge is still an open problem, mainly due to its multiscale na
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