Microstructural Processes in Irradiated Materials: Ferritic and Ferritic-Martensitic Alloys II
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee
Program Organizers: Thak Sang Byun, Pacific Northwest National Laboratory; Chu-Chun Fu, Commissariat à l'énergie atomique et aux énergies alternatives (CEA); Djamel Kaoumi, University of South Carolina; Dane Morgan, University of Wisconsin-Madison; Mahmood Mamivand, University of Wisconsin-Madison; Yasuyoshi Nagai, Tohoku University

Tuesday 2:00 PM
February 28, 2017
Room: Del Mar
Location: Marriott Marquis Hotel

Session Chair: Frederic Soisson, CEA Saclay; Maylise Nastar, CEA Saclay

2:00 PM  Invited
Understanding the Multiple Functions of Point Defects in Fe-based Alloys under Irradiation: Maylise Nastar1; Thomas Schuler2; Luca Messina1; Chu Chun Fu1; Frédéric Soisson1; Pär Olsson3; 1CEA; 2University of Illinois; 3KTH
    Irradiation of alloys yields a super-saturation and permanent fluxes of point defects (vacancies and self-interstitials). The various thermodynamic and kinetic functions of these point defects are illustrated with the specific case of Fe-based dilute alloys including metallic solute atoms and interstitials (C, N, and O). Starting from DFT data of point defect migration barriers, the Self-Consistent Mean Field (SCMF) theory is used to study the stability and mobility of point defect-solute clusters, radiation induced segregations of solutes in the vicinity of point defect sinks, and the role of point defects on the dynamic phase diagrams established under irradiation.

2:30 PM  
Effect of Neutron Irradiation on the Microstructure of a Series of Fe-Cr Alloys: Dhriti Bhattacharyya1; Peter Wells2; Mukesh Bachhav3; Alan Xu1; Emmanuelle Marquis3; G. Robert Odette2; 1ANSTO; 2UCSB; 3University of Michigan
    A series of Fe-Cr alloys with 3%, 6%, 9%, 12%, 15% and 18%Cr, respectively, were irradiated to a dose of 1.8 dpa at 290 °C in the Advanced Test Reactor (ATR) at Idaho National Laboratory. In this talk, the authors present the results of a transmission electron microscopy (TEM) investigation into the effects of varying Cr content on the radiation induced microstructural effects, including both Cr segregation into α’ clusters and displacement damage resulting in dislocation loops. It was found that increasing Cr content resulted in a monotonically increasing number density of Cr rich clusters. On the other hand, it was observed that the number density of dislocation loops first decreased rapidly, reached a minimum at about 9% Cr, and then increased gradually, with the increase in Cr content. The type of dislocation loops observed was found to be predominantly <100> on {200} planes, with some ½<111> on {111} planes.

2:50 PM  
Diffusion Mechanisms of Solutes in Ferritic Steels: Effects ofIirradiation: Caroline Barouh1; Chu-Chun Fu1; Thomas Jourdan1; 1SRMP, CEA-Saclay
    Interstitial solutes (carbon, nitrogen, oxygen) are frequnetly present in steels. We have investigated their diffusion mechanisms in the bcc-iron lattice via Density Functional Theory and Cluster Dynamics simulations. These solutes interact strongly with vacancies. Opposite to common thoughts, some of the resulting solute-vacancy complexes are highly mobile. We show that an oversaturation of vacancies (e.g. under irradiation) may enhance the diffusion of the interstitial solutes. On the other side, Titanium and Ytrium are key alloying elements of ODS (Oxide Dispersion Strengthened) steels, which are promising materials for future nuclear reactors. We predict that the relative mobility of Ti and Y in iron changes as a function of the amount of vacancies present in the material. The decrease of diffusivity of Y with increasing vacancy concentration plays an essential role to dictate the microstructure of these steels.

3:10 PM  
Understanding the Formation and Growth Behavior of Alpha-prime Precipitates in Neutron-Irradiated FeCrAl Alloys Using SANS and APT: Philip Edmondson1; Samuel Briggs2; Yukinori Yamamoto1; Ken Littrell1; Richard Howard1; Charles Daily1; Kurt Terrani1; Kumar Sridharan2; Kevin Field1; 1Oak Ridge National Laboratory; 2University of Wisconsin
    FeCrAl alloys have been proposed as a potential accident-tolerant fuel cladding material for light-water reactors (LWRs). However, their role may be limited due to the formation of the Cr-rich alpha-prime phase that leads to embrittlement in high-Cr ferritic alloys. Four model FeCrAl alloys of varying Cr and Al contents have been neutron-irradiated to doses up to 13.8 displacement per atom at 320 C. Small-angle neutron scattering (SANS) and atom probe tomography (APT) have been used to characterize the alpha-prime precipitates to obtain accurate number densities, volume fractions, size distributions, and chemical compositions. These results have been compared to data obtained from alpha-prime formation in binary FeCr systems. The results are also used to provide insights into the kinetics of the coarsening behavior of the alpha-prime precipitates to enable a better understanding of the mechanism for their formation and growth, and how to mitigate their production through alloy design.

3:30 PM  Cancelled
Strain and Self-ion Irradiation Changes in Cr Atoms Distribution in Fe-Cr Alloys: Stanislaw Dubiel1; Jan Zukrowski1; 1AGH University of Science and Technology
    Distribution of Cr atoms within the first two neighbor shells in strained and strain-free samples of Fe(100-x)Cr(x) alloys (x=5.8, 10.75 and 15.15) irradiated with Fe-ions was studied by Mossbauer spectroscopy. It was quantify in terms of short-range parameters and, alternatively, local concentration of Cr. Strong degree of ordering was revealed in the strained samples viz. a weak overpopulation in 1NN and a strong one in 2NN. In the strain-free samples a weak underpopulation was found for 1NN and a weak overpopulation for 2NN leading to a fairly random distribution on average. The irradiation at 573 K with 2 MeV Fe ions to the dose of 0.6 dpa caused hardly detectable changes, except for the 1NN where some weak decrease of the Cr atoms population was revealed. The latter may be indicative of radiation-induced clustering.

3:50 PM Break

4:05 PM  
Deformation Microstructure of Ferritic/Martensitic Steels Irradiated in Spallation Environment: Kun Wang1; Yong Dai1; Philippe Spatig1; Maximo Victoria1; 1Paul Scherrer Institute
    Three different Ferritic/martensitic steels were tensile tested to understand the mechanisms of embrittlement induced by the combined effects of displacement damage and helium after irradiation in SINQ, the Swiss spallation source. The irradiation conditions were in the range: 10.7 – 20.4 dpa with 850-1750 appm He at 160-300 °C. After tensile testing, TEM observation was employed to investigate the deformation microstructures. Defect free channels with {110} and {112} slip planes were found in deformed area of some specimens, indicating plastic flow localization. Regarding the brittle samples, the TEM-lamella were extracted directly below intergranular or cleavage fracture surfaces by focused ion beam. Strikingly, deformation twinning was observed as the main feature in three irradiated specimens at high dose. Twins with {112} planes were observed in all of these samples. As deformation twinning was first observed in irradiated FM steels, the novel features of deformation twinning are summarized in this presentation.

4:25 PM  
APT Characterization of Post-irradiation Microstructural Changes in T91 Steel: Guma Yeli1; Maria Auger1; Steve Roberts1; Paul Bagot1; Michael Moody1; 1University of Oxford
    T91 steel is a candidate for applications in fast reactors and for replacement alloys for LWRs at high dose. As part of a US-UK Integrated Research Project Project, the aim of this study is to understand the capability to predict the evolution of microstructure of T91 in-reactor and at high doses, using ion irradiation as a surrogate for neutrons. Atom probe tomography (APT) has been used to characterize the microstructural changes in T91 steel for three different conditions, single beam irradiation (Fe++ to 1dpa, 10dpa and 20dpa), dual beam irradiation (Fe++ and He++) and BOR-60 reactor irradiation. Grain boundary segregation, modification of pre-existing precipitates and irradiation-induced precipitation has been compared between different irradiation conditions.

4:45 PM  
Understanding Deformation Dynamics in Neutron-irradiated Fe-based Alloys with High-Energy X-rays: Meimei Li1; Xuan Zhang1; Yiren Chen1; Jonathan Almer1; Jun-Sang Park1; Peter Kenesei1; Hemant Sharma1; Yong Yang2; Chi Xu2; Lizhen Tan3; 1Argonne National Laboratory; 2University of Florida; 3Oak Ridge National Laboratory
    This paper will present recent results of in situ characterization of tensile deformation of neutron-irradiated ferritic and austenitic alloys with wide-angle X-ray scattering (WAXS) and small-angle X-ray scattering (SAXS), and ex situ 3D characterization of microstructure with far-field high-energy X-ray diffraction microscopy (ff-HEDM). High-energy X-rays offer new opportunities for probing mm-sized specimens exposed to external stimulants, e.g. stress, temperature, with multiple probes, which allow characterization of microstructural features ranging from the nano-scale to microscale, linked directly with the macroscale stress-strain behavior of a bulk material. This talk will discuss the new findings of neutron irradiation effects on the deformation process in Fe-based alloys with current in situ thermal-mechanical loading capability of activated materials, and the outlook for in situ time- and spatial- resolved (4D) characterization of grain dynamics in neutron- irradiated materials with high-energy X-rays.

5:05 PM  
Investigation of Elevated Temperature Tensile Deformation of Neutron-irradiated Fe using High-Energy X-ray Techniques: Xuan Zhang1; Chi Xu2; Meimei Li1; Jun-Sang Park1; Jonathan Almer1; 1Argonne National Laboratory; 2University of Florida; Argonne National Laboratory
    We report here the first experiment of an in situ tensile test of a neutron-irradiated pure Fe at 300C to probe the tensile deformation and damage process using combined high-energy X-ray diffraction and small-angle scattering techniques. The Fe specimen was neutron-irradiated to 0.01 dpa at 300C at the Advanced Test Reactor, Idaho National Laboratory. The tensile test was performed at 300C in vacuum using the in-situ Radiated Materials (iRadMat) testing apparatus interfaced with a mechanical load frame at the beamline 1-ID at the Advance Photon Source. Results of the tensile deformation of the irradiated pure Fe will be discussed. The unique experimental capability opens up a new opportunity for understanding the structure-property relationship of nuclear materials.

5:25 PM  
Radiation Effects in RAFM Steels: Ermile Gaganidze1; Christian Dethloff2; Benjamin Kaiser2; Jarir Aktaa2; Daniel Brimbal3; Mickaël Payet3; Lucile Beck3; 1Karlsruhe Institute of Technology, Institute for Applied Materials ; 2Karlsruhe Institute of Technology, Institute for Applied Materials; 3CEA, DEN, Service de Recherches de Metallurgie Physique, Laboratoire JANNUS
    High displacement damage combined with high helium content accumulated in reduced activation ferritic/martensitic (RAFM) structural steels of a fusion reactor will strongly degrade their mechanical properties. We summarize the mechanical properties of the RAFM steels for different neutron irradiation conditions. The irradiation induced defects are assessed by quantitative TEM and linked to the irradiation hardening in EUROFER97. Helium effects are simulated by dual beam irradiations carried out at 330, 400 and 500°C at Jannus Saclay facility. The helium bubble size distribution and swelling are studied by TEM. Low temperature ion implantation lead to spatially homogeneous bubble distribution, whereas spatially heterogeneously nucleated bubbles additionally appeared at 500°C. The study of the helium bubble evolution is supplemented by a kinetic rate theory model based on diffusion governed helium nucleation and growth. The model provides good qualitative description of the observed bubble microstructure though peak bubble diameters are overestimated in comparison to experiment.

5:45 PM  
A Predictive Model for Irradiation-induced Nanocluster Evolution in b.c.c. Fe-based Alloys: Matthew Swenson1; Janelle Wharry2; 1Boise State University; 2Purdue University
    The objective of this work is to advance a model that predicts irradiation-induced nanocluster evolution in b.c.c. Fe-based alloys. Ferritic oxide dispersion strengthened (ODS) steels, ferritic-martensitic (F/M) steels, and other b.c.c. Fe-based alloys are leading candidates for advanced nuclear reactor structural and cladding materials. However, experiments have shown variable irradiation evolution of nanoclusters in these alloys. For example, ODS oxide nanoclusters either coarsen or dissolve, while F/M alloys exhibit nucleation of Si-Mn-Ni-Cu and Cr-rich nanoclusters. Thus there is a critical need for a predictive model of irradiation-induced nanocluster evolution. This work develops an empirically-validated model, depending on both the irradiation conditions (i.e. irradiating particle, temperature, and dose rate) and the pre-existing cluster morphology and alloy composition, to predict irradiation-induced cluster evolution in ODS and F/M alloys. This model is a potentially versatile and valuable tool to aid b.c.c. Fe-based and nanofeatured alloy development for nuclear reactor environments.