Nanostructured Materials for Nuclear Applications II: Session I
Sponsored by: TMS Materials Processing and Manufacturing Division, TMS Structural Materials Division, TMS: Nanomechanical Materials Behavior Committee, TMS: Nuclear Materials Committee
Program Organizers: Cheng Sun, Idaho National Laboratory; Khalid Hattar, Sandia National Laboratories ; Celine Hin , Virginia Tech; Fei Gao , University of Michigan; Osman Anderoglu , Los Alamos National Laboratory; Mitra Taheri , Drexel University; Haiming Wen , Idaho State University

Monday 8:30 AM
February 27, 2017
Room: Pacific 24
Location: Marriott Marquis Hotel

Session Chair: Khalid Hattar, Sandia National Laboraory ; Mitra Taheri , Drexel University

8:30 AM  Invited
Understanding and Predicting Nanoscale Precipitate Formation and Associated Reactor Pressure Vessel Embrittlement: Dane Morgan1; Huibin Ke1; Mahmood Mamivand1; Shipeng Shu1; Henry Wu1; Peter Wells2; Nicholas Cunningham2; Nathan Almirall2; G. Robert Odette2; 1University of Wisconsin - Madison; 2University of California, Santa Barbara
    Irradiation enhanced precipitation hardening is the primary cause of in-service embrittlement of reactor pressure vessel (RPV) steels. In this work we describe recent modeling approaches based on cluster dynamics to understand and predict the Cu and Mn-Ni-Si precipitates that dominate precipitation hardening in RPVs. We demonstrate that Mn-Ni-Si precipitates are thermodynamically stable, driven by both homogeneous and cascade damage nucleation, significantly influenced by the presence of Cu, sensitive to local composition and temperature variations, and can play a critical role in RPV embrittlement under life extension conditions. We also describe machine learning regression approaches to predicting embrittlement effects, including kernel methods, neural networks, and decision trees. We demonstrate that these methods have excellent interpolation capability, although extrapolation is a significant challenge. Models are compared to extensive irradiation data from the ATR1 and Irradiation VARiable (IVAR) databases developed at UC Santa Barbara.

9:00 AM  
Search for Radiation Resistance Materials: As Revealed by Computer Simulations: Fei Gao1; Liangliang Liu1; Nanjun Chen1; Chenyang Lu1; Lumin Wang1; 1University of Michigan
    Recently, there have been a large number of researches dealing with radiation resistance by introducing high-densities of defect sinks, such as secondary phase, grain boundaries and multi-layers, is a popular way for reducing residual defects in irradiated materials. However, we will demonstrate unprecedented mechanisms of radiation tolerance and suppression of void formation in multicomponent single-phase alloys by simulating the migration behavior of point defects and nano defect clusters in Ni, NiCo and NiFe. A striking feature is that the similar migration energy of vacancy clusters and interstitials in NiFe enhances defect recombination, thus improving its radiation resistance. Another important mechanism for the enhanced swelling resistance in NiFe is attributed to interstitial defect cluster motion in the alloys from a long-range one-dimensional mode to a short-range three-dimensional mode that leads to a dramatically enhanced recombination of interstitials and vacancies, significantly suppressing void formation as compared with that in Ni.

9:20 AM  
Kinetic Mote Carlo Simulation of Radiation-induced Segregation in Quaternary Fe-Ti- Y-O: Christopher Nellis1; Celine Hin1; 1Virginia Tech
    Nanostructured ferritic alloys (NFAs) are materials being studied for use in nuclear reactors as cladding material to separate nuclear fuel rods from the plant’s coolant. An area of concern for these materials is radiation induced segregation of alloying elements to grain boundaries of NFAs, which would embrittle the material. A kinetic Monte Carlo model was created to model the migration of atoms and point defects in an Fe-Ti- Y-O alloy around a grain boundary during irradiation. The model will notably incorporate the migration of oxygen, an element that exists in the octahedral sites in the iron matrix. Additionally, the energy of segregation at the grain boundary is added to the system. The study will look at the enrichment or depletion of alloying elements at the grain boundaries under various irradiation conditions and alloy compositions.

9:40 AM  
Molecular Dynamic Cascade Simulations of Yttria Nanoclusters in an Alpha Fe Matrix: Mike Higgins1; Fei Gao1; 1University of Michigan
    Molecular dynamic (MD) simulations of Y2O3 in a α-Fe matrix were used to understand the structure and radiation resistance of Y2O3 Nanoclusters (NCs) in oxide dispersion strengthened (ODS) steels. A core-shell model was predicted using our modified Buckingham potential combined with a Ziegler, Biersack, and Littmark potential. ODS steels have shown increased irradiation resistance, and this increased resistance is due to the introduction of Y2O3 NCs. The NCs will vary in size from 3 nm to 5 nm, along with a change in temperature from 100K to 600K. The energy transferred to the cluster versus to the matrix, tends to be a major factor in the stability of the Y2O3 clusters during a cascade. The smaller NCs have a higher chance of transferring less damage to the cluster with the energy able to easily transfer through the cluster compared to the larger NCs.

10:00 AM Break

10:20 AM  Invited
Irradiation Response of Nanostructured Oxides to Ionization and Displacement Damage: Yanwen Zhang1; Dilpuneet Aidhy2; Tamas Varga3; Philip Edmondson1; Fereydoon Namavar4; William Weber5; 1Oak Ridge National Laboratory; 2University of Wyoming; 3Pacific Northwest National Laboratory; 4University of Nebraska Medical Center; 5University of Tennessee
    Understanding microstructural change of nanostructured materials to irradiation is important for advanced nuclear energy systems. Nanostructured cubic ceria and zirconia with grain size of a few nanometers are used as non-radioactive surrogates to evaluate irradiation behavior of actinide dioxides with the fluorite structure. Irradiations are performed over a range of ion mass, energy, fluence and temperature. Defect production, grain growth, oxygen stoichiometry, and phase stability are investigated. While the cubic phase is stable under irradiation, an additive effect on grain growth from both collision cascades and ionization is revealed. The observed grain growth is disorder-driven that is triggered by the active interaction between irradiation-induced disorder and grain boundary. It has also clarified the controlling mechanisms of grain growth by minimizing the defect activity or concentration in small grains and by pinning grain boundary movement in the larger grains. This work was supported by the U.S. DOE, BES, MSED.

10:50 AM  
Evolution of Microstructures and Mechanical Properties of Zr-containing Ferritic Alloys under Self-ion Irradiation: Tianyi Chen1; Mo-Rigen He2; Lizhen Tan1; Ying Yang1; Beata Tyburska-Püschel2; Kumar Sridharan2; 1Oak Ridge National Laboratory; 2University of Wisconsin, Madison
    A series of Zr-containing ferritic alloys were developed to have dispersed fine intermetallic particles, which are believed to enhance the radiation tolerance and mechanical strengths of the materials. Microstructures of the alloys with different chemical compositions were characterized using transmission electron microscopy (TEM), energy-dispersive X-ray spectroscopy and X-ray diffraction techniques. Alloy compositions were found to affect the distributions and crystal structures of the intermetallic phases. Consequently, the radiation-induced microstructure changes are different in the alloys because of the various types of intermetallics. The resultant mechanical property changes were studied by nanoindentation in irradiated and unirradiated samples. TEM characterizations at the vicinity of the indents revealed deformation microstructures. This study provides insights into the radiation tolerance of intermetallic-strengthened alloys, interfase-defect interactions and localized deformation mechanisms of these novel materials.

11:10 AM  
Stability of 14YWT Nanostructured Ferritic Alloys under Irradiation and Thermal Aging: Eda Aydogan1; Stuart Maloy1; Osman Anderoglu1; Sven Vogel1; Clarissa Yablinsky1; Nathan Almirall2; G. Robert Odette2; Jonathan Gigax3; Lloyd Price3; Di Chen3; Lin Shao3; Frank Garner3; 1Los Alamos National Laboratory; 2University of California Santa Barbara; 3Texas A&M University
    Nanostructured ferritic alloys (NFAs) are attractive materials for core components in Gen-IV reactors because of their excellent high temperature strength, stability, and radiation tolerance. In this research, two sets of experiments were conducted. First, the stability of the 14YWT microstructure with varying deformation was investigated at temperatures >800°C using in-situ neutron diffraction. It was found that the microstructure is very stable and recrystallization starts at times depending on temperature and deformation. Second, 14YWT NFAs were irradiated with 3.5 MeV Fe2+ ions up to 1100 peak dpa at 450 °C. Their initial and post irradiation microstructures were investigated using transmission electron microscopy (TEM) and atom probe tomography (APT). It was found that even after high dose irradiation, these alloys show almost zero swelling. Both TEM and APT indicate that the nano-oxides, having the size <5 nm, are quite stable under irradiation. A theoretical model was established to explain the nano-oxide stability.

11:30 AM  
In-situ TEM Study of Defect-grain Interactions under Irradiation in Bulk Severe Plastically Deformed Model Ni Alloys: Christopher Barr1; Marquis Kirk2; Meimei Li2; Mitra Taheri1; 1Drexel University; 2Argonne National Laboratory
    The use of ultrafine grain and nanocrystalline alloys produced by bulk severe plastic deformation have recently gained extensive interest in nuclear energy and extreme environment applications due to their high volume of grain boundaries and excess dislocation density. In this study, we examine the interaction of irradiation induced defect clusters and dislocation loops with grain boundaries in a model Ni fcc alloy after severe plastic deformation through in-situ ion irradiations in the TEM. The outcomes include a higher defect absorption rate at grain boundaries in severe plastic deformed grains compared to coarse grain counterparts. In addition, the average irradiation induced dislocation size is shown to decrease in the ultrafine grain size regime. The implications for a reduced radiation induced defect size and increased defect-GB interactions is discussed in the context of improvements in radiation response by tailoring microstructures with both specific grain boundary character and grain boundary density.