Abstract
Radiation-induced segregation (RIS) of solutes in materials exposed to irradiation is a well-known problem. It affects the life-time of nuclear reactor core components by favouring radiation-induced degradation phenomena such as hardening and embrittlement. In this work, RIS tendencies in face-centered cubic (fcc) Ni-X (X = Cr, Fe, Ti, Mn, Si, P) dilute binary alloys are examined. The goal is to investigate the driving forces and kinetic mechanisms behind the experimentally observed segregation. By means of ab initio calculations, point-defect stabilities and interactions with solutes are determined, together with migration energies and attempt frequencies. Transport and diffusion coefficients are then calculated in a mean-field framework, to get a full picture of solute-defect kinetic coupling in the alloys. Results show that all solutes considered, with the exception of Cr, prefer vacancy-mediated over interstitial-mediated diffusion during both thermal and radiation-induced migration. Cr, on the other hand, preferentially migrates in a mixed-dumbbell configuration. P and Si are here shown to be enriched, and Fe and Mn to be depleted at sinks during irradiation of the material. Ti and Cr, on the other hand, display a crossover between enrichment at lower temperatures, and depletion in the higher temperature range. Results in this work are compared with previous studies in body-centered cubic (bcc) Fe, and discussed in the context of RIS in austenitic alloys.
Highlights
Ni-based alloys and austenitic stainless steel are common structural materials in current and future generation nuclear power plants (NPPs)
Flux coupling, and radiation-induced segregation (RIS) of Fe, Cr, P, Si, Ti, and Mn in fcc Ni have here been investigated by coupling first-principles calculations with the self-consistent mean field theory
The goal has been to improve the current understanding of radiationinduced segregation processes of materials commonly used in today’s and future generation nuclear power plants
Summary
Ni-based alloys and austenitic stainless steel are common structural materials in current and future generation nuclear power plants (NPPs). Novel material classes such as highentropy alloys (HEAs), or concentrated solid solution alloys (CSAs) are materials that can be Ni based with a face centered cubic (fcc) structure. Such materials can in the nuclear technology sector be exposed to intense radiation fields, which can have a great impact on their properties. Experimental characterization of irradiated material provides important insight on the processes involved in its degrada-
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