Mechanical Behavior of Nuclear Reactor Components: Defect Evolution
Sponsored by: TMS Materials Processing and Manufacturing Division, TMS Structural Materials Division, TMS: Nanomechanical Materials Behavior Committee, TMS: Nuclear Materials Committee
Program Organizers: Clarissa Yablinsky, Los Alamos National Laboratory; Assel Aitkaliyeva, University of Florida; Eda Aydogan, Middle East Technical University; Laurent Capolungo, Los Alamos National Laboratory; Khalid Hattar, University of Tennessee Knoxville; Kayla Yano, Pacific Northwest National Laboratory; Caleb Massey, Oak Ridge National Laboratory
Monday 2:00 PM
March 15, 2021
Room: RM 50
Location: TMS2021 Virtual
2:00 PM Invited
Simulation of Intergranular Void Growth under the Combined Effects of Surface Diffusion, Grain Boundary Diffusion, and Bulk Creep: John Sanders1; Negar Jamshidi1; Niloofar Jamshidi1; Mohsen Dadfarnia2; Sankara Subramanian3; James Stubbins4; 1California State University, Fullerton; 2Seattle University; 3PhotoGAUGE; 4University of Illinois at Urbana-Champaign
Creep rupture is currently a major concern for next-generation nuclear reactor components, and many commonly used lifetime estimates are based on how quickly intergranular voids grow. Void growth is caused by three processes: diffusion along the void surface, diffusion along the grain boundary, and creep of the surrounding grains. Previous modeling efforts have only considered two of these three processes at a time. Here we present finite element simulations of void growth under the influence of all three mechanisms simultaneously. To our knowledge, these are the first such simulations to be reported in the literature. Based on our simulations, we develop quantitative criteria for quasi-equilibrium and crack-like void growth, and compare them to previous results. Furthermore, we find that void growth is highly accelerated during the primary creep regime. Our results promise to aid in the development of microstructure-sensitive material strength models for next-generation nuclear reactor components.
2:30 PM
A Novel Displacement Cascade Driven Irradiation Creep Mechanism in Pure Copper: Nargisse Khiara1; Fabien Onimus1; Laurent Dupuy1; Jean-Paul Crocombette1; Stéphanie Jublot-Leclerc2; Thomas Jourdan1; Thomas Pardoen3; Jean-Pierre Raskin4; Yves Bréchet5; 1CEA Saclay; 2Université Paris–Saclay; 3Ecole Polytechnique de Louvain, Institute of Mechanics, Materials and Civil Engineering (IMMC), Materials and process engineering, Belgium; 4Ecole Polytechnique de Louvain; 5SIMAP - Science et Ingénierie des MAtériaux et Procédés, Grenoble-INP, France
Metals and alloys used as structural materials in the nuclear core of pressurized water reactors undergo irradiation creep deformation. The mechanical behavior is well characterized at the macroscopic level. Yet, the underlying microscopic mechanisms are still unclear. In situ straining experiments under ion irradiation were conducted on pure copper. We observed that, at high stress levels, dislocations pinned on irradiation induced point defects clusters start to glide once under irradiation. The usual irradiation creep climb induced glide mechanisms were not sufficient to explain the experimental observations. Another mechanism was proposed suggesting that the dislocation glide assisted by irradiation is due to a direct interaction between the displacement cascade and the pinned dislocation. A molecular dynamics study was conducted to test this hypothesis showing that the unpinning induced by a displacement cascade of a dislocation pinned on a loop can be observed under certain conditions of stress, cascade energy and position.
2:50 PM
Controlling Helium Morphology in Pure Metals: Dislocation-helium Interactions: Calvin Lear1; Jonathan Gigax1; Nan Li1; Saryu Fensin1; 1Los Alamos National Laboratory
Prolonged irradiation of structural components results in the accumulation of excess microstructural defects, diminished confidence in performance, and increased overall cost. Although the evolution of helium is a vital aspect of structural material degradation, only grain boundary effects on materials strength are well understood. A systematic study of temperature-dose-dose rate interactions was thus performed for helium morphologies in pure metals. Helium implanted samples were probed using a combination of electron microscopy, micro-mechanical, and nano-mechanical techniques. In situ testing was used extensively to better understand changes in deformation mode with implantation and to directly observe dislocation interactions with the helium defects. These findings are considered in terms of materials selection and control of helium defect nucleation and growth.
3:10 PM
Correlating the Neutron-irradiation Induced Hardening and Solute Nano-clustering in Oxide Dispersion Strengthened Alloys: Samara Levine1; Arunodaya Bhattacharya2; Andrew Lupini2; David Hoelzer2; Yutai Katoh2; Steven Zinkle1; 1Oak Ridge National Laboratory, University of Tennessee; 2Oak Ridge National Laboratory
Oxide dispersion strengthened (ODS) alloys are candidate materials for the first wall/blanket structures in fusion reactors. However, solute nano-cluster behavior under neutron damage (ballistic dissolution, reprecipitation, nanocluster alterations) is not well-understood in these alloys. At ORNL, two ODS alloys: 20%Cr-5.5%Al based PM2000 and 14%Cr based MA957 were irradiated in the High Flux Isotope Reactor at 300°C to ~3-80 dpa. Previously, we revealed transformation of nano-dispersoids into amorphous, nano-porous structures in ~80 dpa irradiated PM2000. Here, we reveal that in addition to Cr-rich α’ phase separation, unexpected Al/Ti nano-clustering occurred in the PM2000 matrix. Likewise, irradiated MA957 suffered Ni/Ti nano-clustering. By combining analytical scanning electron transmission microscopy (STEM), atom probe tomography (APT) and Vickers micro-hardness tests with hardening models, we will correlate the contribution of solute nano-clustering to hardening of ODS alloys. Research sponsored by the U.S. Department of Energy, Office of Fusion Energy Sciences, under contract DE-AC05-00OR22725 with UT-Battelle, LLC.
3:30 PM Invited
Effect of Cr Concentration On <111> and <100> Dislocation Loop Formation in Fe-Cr Alloys: Yaxuan Zhang1; Ziqi Xiao1; Xian-Ming Bai1; 1Virginia Polytechnic Institute and State University
Radiation can produce both 1/2<111> and <100> type interstitial dislocation loops in Fe-based ferritic alloys. They can hinder the dislocation motion and lead to radiation hardening. Although the observation of the two loop types in Fe-Cr alloys has been widely reported, contradictory experimental results exist in terms of the Cr effect on their relative abundance, partially because different irradiation conditions and characterization methods were used. In this work, we use molecular dynamics simulations to study the effects of Cr concentration on the relative abundance of two types of loops in Fe-Cr alloys. Our results show that the formation probability of <100> type of loops decreases significantly with the increasing Cr concentration while that of 1/2<111> type of loops is less affected. Our independent molecular statics calculations show that Cr can suppress <100> loop formation more strongly than 1/2<111> loops. The possible effects of other alloying elements are also discussed.
4:00 PM
Void Swelling and Transmutation in Tungsten Metals and Alloys after Fusion Relevant Neutron Irradiation: Daniel Morrall1; John Echols1; Josina Geringer1; Lauren Garrison1; Chad Parish1; 1Oak Ridge National Lab
PHENIX is a U.S.-Japan collaboration for the technological assessment of plasma facing components for experimental fusion power plants. One of the main goals of this experiment is to understand the thermomechanical properties of tungsten irradiated with a transmutation-to-dpa ratio relevant to fusion plasma facing components (PFCs). PFC candidate materials were irradiated in HFIR at ORNL at fusion relevant temperatures in gadolinium shielded capsules, reducing the thermal-to-fast neutron flux ratio ~10x. Four samples were chosen for post irradiation analysis; rolled foil tungsten (RFW), single crystal tungsten (SCW), polycrystalline tungsten (PCW), and W-Re alloy (WRA). In this work we have analyzed these samples using conventional and scanning transmission electron microscopy ((S)TEM) and energy dispersive x-ray spectroscopy (EDS). Preliminary comparisons of each sample reveal the fewest cavities in WRA affecting the Orowan hardening calculations which may offer an explanation to the mechanism of reduced embrittlement of other W-Re alloys.
4:20 PM
Irradiation Resistance in Several Multi-principal Element Alloys: Yanqing Su1; 1Utah State University
Multi-principal element alloys (MPEAs) are alloys that form solid solution phases and consist of three or more principal metallic elements. They are considered to have intermediate structural and chemical complexities between single-element regular metals and multi-element disordered metallic glasses. Due to their unique microstructures and chemical compositions, MPEAs exhibit excellent mechanical properties such as high strength at elevated temperatures and excellent irradiation resistance. Since radiation damage is caused by radiation-induced defects and their evolution such as voids and dislocation loops, the primary radiation damage process is an atomic-level material phenomenon. Here, the generation and evolution of radiation-induced point defects in several MPEAs are characterized by atomistic simulations. Results show that the numbers of residual defects after cascade and evolution in these MPEAs are smaller than those in their constituent pure metals. The origin of the higher radiation damage tolerance in MPEAs is discussed.