Dependence on grain boundary structure of radiation induced segregation in a 9 wt.% Cr model ferritic/martensitic steel Kevin G. Field a,⇑ , Leland M. Barnard a , Chad M. Parish b , Jeremy T. Busby c , Dane Morgan a , Todd R. Allen a a Materials Science Program, University of Wisconsin, Madison, WI 53706, USA b Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37832, USA c Fuel Cycle & Isotopes Division, Oak Ridge National Laboratory, Oak Ridge, TN 37832, USA article info Article history: Received 26 July 2012 Accepted 13 December 2012 Available online 21 December 2012 abstract Ferritic/Martensitic (F/M) steels containing 9 wt.% Cr are candidates for structural and cladding compo- nents in the next generation of advanced nuclear fission and fusion reactors. Although it is known these alloys exhibit radiation-induced segregation (RIS) at grain boundaries (GBs) while in-service, little is known about the mechanism behind RIS in F/M steels. The classical understanding of RIS in F/M steels presents a mechanism where point defects migrate to GBs acting as perfect sinks. However, variation in grain boundary structure may influence the sink efficiency and these migration processes. A proton irradiated 9 wt.% Cr model alloy steel was investigated using STEM/EDS spectrum imaging and GB mis- orientation analysis to determine the role of GB structure on RIS at different GBs. An ab initio based rate theory model was developed and compared to the experimental findings. This investigation found Cr preferentially segregates to specific GB structures. The preferential segregation to specific GB structures suggests GB structure plays a key role in the mechanism behind radiation-induced segregation, showing that not all grain boundaries in F/M steels act as perfect sinks. The study also found how irradiation dose and temperature impact the radiation-induced segregation response in F/M steels. Ó 2012 Elsevier B.V. All rights reserved. 1. Introduction Advanced fusion and fast neutron fission reactors demand materials that remain stable at high temperatures and perform to high irradiation doses. Advanced 9 wt.% Cr Ferritic/Martensitic (F/ M) steels, including oxide dispersion strengthened steels, are a promising material class for cladding and structural components in these advanced reactor designs due to their corrosion resistance, low swelling rate under irradiation and creep strength [1]. A mat- ter of concern for F/M steels for advanced nuclear reactor applica- tions is the lack of information regarding the radiation-induced segregation (RIS) mechanisms at internal interfaces, including grain boundaries. Segregation of Cr to grain boundaries under irra- diation is a concern as Cr provides strength and corrosion resis- tance to the steel matrix [2]. The current body of work on this topic [3–25] presents large variability in the Cr segregation behav- ior under irradiation at grain boundaries for F/M steels. Historically, the variability in literature was assumed to be the result of the wide range of irradiation conditions and alloy systems studied. However, recent systematic studies of Cr segregation to grain boundaries show variability in segregation from boundary to boundary in samples with the same experimental conditions [17,18]. A plausible explanation for the variability observed is changes in the atomic structure of grain boundaries depending on the individual grain boundary structure. Grain boundary struc- ture, through the use of misorientation axis/angle pair and coinci- dent site lattice convention descriptions, has been shown to dictate the material properties of many polycrystalline metals including corrosion resistance, fracture, recrystallization, and diffusion behavior [26,27]. Current RIS theory on irradiated steels predicts the resulting grain boundary chemistry is a competition between vacancy-solute and interstitial-solute diffusion to and interaction with a grain boundary where the grain boundary acts as a perfect sink for these complexes [28]. A more rigorous approach to RIS the- ory would include how alterations in the grain boundary structure change the defect sink efficiency and defect interaction and hence the observed RIS response, but such concepts have yet to be ap- plied to F/M steels. Furthermore, limited work has been completed to develop RIS models for BCC systems and these models do not ac- count for complex phenomena such as impurities or evolving microstructures under irradiation [21,25,29,30]. This work pro- vides the critical first steps in developing such a model by estab- lishing the relationship between RIS and grain boundary structures as well as irradiation temperature and dose using a sys- tematic irradiation campaign for an irradiated 9 wt.% Cr F/M model steel and an ab initio based rate theory model. 0022-3115/$ - see front matter Ó 2012 Elsevier B.V. All rights reserved. http://dx.doi.org/10.1016/j.jnucmat.2012.12.026 ⇑ Corresponding author. Address: Materials Science Program, 1500 Engineering Drive, ERB719, Madison, WI 53706, USA. Tel.: +1 608 262 4711. E-mail addresses: kgfield@wisc.edu (K.G. Field), lmbarnard@wisc.edu (L.M. Barnard), parishcm@ornl.gov (C.M. Parish), busbyjt@ornl.gov (J.T. Busby), ddmor- gan@wisc.edu (D. Morgan), allen@engr.wisc.edu (T.R. Allen). Journal of Nuclear Materials 435 (2013) 172–180 Contents lists available at SciVerse ScienceDirect Journal of Nuclear Materials journal homepage: www.elsevier.com/locate/jnucmat