A multiscale approach to study the effect of chromium and nickel concentration in the hardening of iron alloys I.N. Mastorakos ⇑ , H.M. Zbib School of Mechanical and Materials Engineering, Washington State University, Pullman, WA 99164, United States article info Article history: Received 18 November 2013 Accepted 2 March 2014 Available online 11 March 2014 abstract A multiscale approach is presented to study the effect of chromium and nickel concentration on the deformation behavior of iron systems. A combination of molecular dynamics (MD) and dislocation dynamics (DD) simulations are employed. In this framework, the 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 mobility is obtained from MD sim- ulations. Since accurate measurements of dislocation velocity are difficult to obtain through experiment, atomistic simulations constitute an adequate alternative tool. Then this information is used by DD to sim- ulate large systems with high dislocation and defect densities. In particular, we study the effect of nickel and chromium concentration on the strength, as well as the effect of dislocation loops concentration on the yield stress of the aforementioned systems. The results reveal, among other, defect free zones, in accordance to experimental observations, and an evolution law for the defect density. Ó 2014 Elsevier B.V. All rights reserved. 1. Introduction The development of new-generation nuclear reactors depends on the availability of materials that can operate safely in severe environments for an extended service lifetime [1,2]. Materials operating in such a harsh environment are subjected to high doses of irradiation which cause changes in microstructure. These changes are responsible for dimensional instabilities, such as swelling and irradiation creep, and mechanical property evolution and degradation, such as irradiation hardening and post-yield deformation behavior including plastic flow and subsequent local- ization, which impact component performance and reliability. The evolution of microstructural features with irradiation dose and temperature involves coalescence of vacancies and interstitials into voids and dislocation loops that cause swelling. In steels, void swelling can occur at temperatures up to about 800 K [3,4]. Modern steels used in nuclear reactors include high strength, fer- ritic, martensitic and oxide dispersion strengthened steels. These steels reduce the occurrence of void swelling and are more stable in the presence of defects that arise due to their severe functional environment. These steels are used in conditions subjected to creep and stress corrosion in the reactor environment that ad- versely affect their useful service life. Although ferritic–martensitic steels are quite resistant to swelling and maintain good fracture toughness at irradiation above 673 K [5,6], they are prone to loss of ductility at lower irradiation temperatures [7,8]. 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 mech- anisms [9]. For example, it is well established that the effect of irra- diation on ferritic/martensitic alloys at low to intermediate temperatures is to increase yield stress, reduce strain hardening capacity and initiate flow localization at lower strains [10]. Furthermore, the predominant microstructural features include dislocation loops, voids, regions of solute segregation and second phase precipitates. The initial density and evolution of these fea- tures depends on some key variables, such as irradiation tempera- ture, dose and dose rate, helium production rate and alloy composition. The mechanical properties on their part depend on the interaction of dislocations with the defects as well as the interaction between dislocations. Due to the cost of the development and testing of such new materials, the use of computational techniques has been proven to be highly preferable and cost-effective. These techniques in- clude, among others, first-principle, molecular and dislocation dynamics and were used by a number of authors [11–14], to model the behavior of prospective materials in future nuclear energy systems. Previous empirical models are being replaced by more http://dx.doi.org/10.1016/j.jnucmat.2014.03.005 0022-3115/Ó 2014 Elsevier B.V. All rights reserved. ⇑ Corresponding author. Tel.: +1 509 335 3215. E-mail address: mastorakos@wsu.edu (I.N. Mastorakos). Journal of Nuclear Materials 449 (2014) 101–110 Contents lists available at ScienceDirect Journal of Nuclear Materials journal homepage: www.elsevier.com/locate/jnucmat