Radiation Effects in Metals and Ceramics: Mechanical Assessment of Irradiated Microstructures
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee
Program Organizers: Djamel Kaoumi, North Carolina State University; Thak Sang Byun, Oak Ridge National Laboratory; Dane Morgan, University of Wisconsin-Madison; Maria Okuniewski, Purdue University; Mahmood Mamivand; Geoffrey Beausoleil, Idaho National Laboratory; Philip Edmondson, The University Of Manchester; Khalid Hattar, University of Tennessee Knoxville; Aurelie Gentils, Université Paris-Saclay; Joel Ribis, Cea
Wednesday 8:30 AM
February 26, 2020
Room: Theater A-7
Location: San Diego Convention Ctr
Session Chair: Peter Hosemann, University of California Berkeley; Maria Okuniewski, Purdue University
8:30 AM Invited
A New Numerical Method to Simulate Dislocation Self-climb: Fengxian Liu1; Edmund Tarleton1; Alan Cocks1; 1University of Oxford
Dislocations can provide short circuit diffusion paths for atoms. This fast atomic transport, along the dislocation core region, known as core diffusion, accelerates the diffusion of impurities by more than three orders of magnitude compared to lattice diffusion. This allows dislocation motions perpendicular to the original slip system, known as dislocation self-climb (conservative climb) and is of particular importance in low-temperature creep and post-irradiation annealing. A variational principle is presented for the analysis of problems in which fast dislocation core diffusion is the dominant mechanism for material redistribution. A new finite element discretization method is employed to accelerate the simulation of dislocation self-climb. Self-climb dominated processes, including the evolution of elongated loops, dislocation dipoles and dislocation patterning during annealing of a large population of loops, involving both self-climb and glide motion are investigated.
9:00 AM
Three Dimensional Rate Theory Models of Radiation Damage with Mechanical Fields: Aaron Kohnert1; Laurent Capolungo1; 1Los Alamos National Laboratory
Coupled cluster dynamics and discrete dislocation dynamics can study the interactions between point defects and particular dislocation microstructures. In particular it allows a complete description of the collective effects of long and short range strain field interactions on point defect flux. The sink strength of representative 3d dislocation networks for a variety of defects is computed in this study using the coupled framework. A probability distribution of the dislocation bias factor is found, which accounts for curvature and configuration effects, the character of the segments, and the superposition of strain fields. In addition to global properties like the bias factor, these effects motivate wide variations in local quantities such as local defect supersaturation and the current of point defects into dislocations.
9:20 AM
Characterizing Self-ion Irradiated Tungsten: Nano-indentation, Multi-technique Microscopy and Crystal-plasticity Modeling: Suchandrima Das1; Hongbing Yu1; Kenichiro Mizohata2; Edmund Tarleton1; Felix Hofmann1; 1University of Oxford; 2University of Helsinki
Tungsten is the front-runner material for armour components in future fusion reactors. In-service irradiation with fusion neutrons will generate displacement cascades, leaving behind lattice defects. Understanding irradiation-induced evolution of mechanical properties, as a function of dose, is essential to accurately predict component performance and lifetime. Self-ion-implantation provides a convenient approach for mimicking neutron-irradiation damage. Here, we characterise the deformation behaviour of self-ion-implanted tungsten exposed to different damage levels, using nano-indentation, atomic force microscopy, and high-resolution electron backscatter diffraction. <001>-grains of implanted tungsten show increased hardness, pile-up and confined lattice distortions around nano-indents; all of which increase with dose until 0.1 dpa, and saturate thereafter. These observations suggest that dislocation motion is initially obstructed by self-ion-induced defects, but that the obstacle strength is reduced by progressive passage of dislocations, leading to strain-softening. Experimental results are directly compared with 3D crystal plasticity finite element calculations we have developed that capture these effects.
9:40 AM
Nanoscale Characterisation of Re/Os Precipitation and Mechanical Degradation In Neutron Irradiated Tungsten: Matthew LLoyd1; Duc Nguyen1; Michael Moody1; Paul Bagot1; Robert Abernethy1; David Armstrong1; 1University of Oxford
In this research, neutron irradiated single and polycrystalline, pure-W was irradiated to 1.67dpa at 1174K, producing a post irradiation composition of W-1.4at.%Re-0.1at.%Os. Characterisation of this material was carried out using Atom Probe Tomography (APT) and Scanning Tunnelling Electron Microscopy (STEM). The change in brittle to ductile transition was measured using 4 point bending and hardening using nanoindentationPrecipitates, mostly comprised on Re but with an Os rich central region were identified using APT. The presence of voids associated with these precipitates was indicated by regions of higher atomic density in the reconstruction of the APT data. High Angle Annular Dark Field (HAADF) imaging and STEM EDX of the samples identified voids decorated with both Re and Os, as well as smaller precipitates without visible voids. The brittle to ductile transition temperature is seen to increase by 600K. The mechanisms for this will be discussed through an energetics based model.
10:00 AM Break
10:20 AM Invited
High Temperature Small Scale Mechanical Testing and Nanoindentation of Advanced Accident Tolerant Fuels: David Frazer1; Joshua White1; Tarik Saleh1; Darrin Byler1; 1Los Alamos National Laboratory
The Fukushima accident prompted a significant amount of work on developing advanced accident tolerant fuels for light water reactors. Through these campaigns a variety of different fuel forms have been developed and proposed such as UN, U3Si2, and UO2 with additives. While the thermophysical properties of these fuels are important in understanding the fuel performance, there is a need for mechanical properties of these advanced fuel forms to evaluate phenomenon like the pellet clad mechanical interaction during operation. Important mechanical properties that are needed of these advanced accident tolerant fuel forms are the elastic modulus, fracture properties and creep of the fuel at the operating temperature. In this work high temperature small scale mechanical testing and nanoindentation is performed on these accident tolerant fuels to measure the mechanical properties over temperature such as the elastic modulus, hardness, and fracture stress which will be discussed in the context of fuel performance.
10:50 AM
Studies of High dpa Ion Beam Irradiation fcc-bcc Duplex Steel 2205: Micromechanical Testing and Nanoindentation Examination of Hardness Variations: Michael Saleh1; Alan Xu1; Paul Munroe2; Dhriti Bhattacharyya1; 1ANSTO; 2School of Material Science and Engineering, University of NSW
Irradiation of metals and alloys causes the formation of defects such as vacancy and interstitial clusters, dislocation loops, voids, bubbles, etc. which results in pronounced hardening depending on the irradiation and testing temperature. The current study focuses on ion beam irradiation of Duplex steel 2205 with 12 MeV Au+5 ions. Post irradiation studies of the micromechanical behaviour are done through nano-indentation. This allows for estimates of moduli and relative estimates of the strengths and hardening of individual phases and individual grains within a multiphase alloy. The results show the increase in hardness of the Duplex steel 2205 (bcc/fcc phases) to be phase dependent with the fcc phase exhibiting greater hardening. Coupling these results to micromechanical tensile tests and crystal plasticity modelling, the authors describe the role of multi-scale modelling in complementing micromechanical testing for irradiation studies.
11:10 AM
Comprehensive Evaluation of the Degradation of Duplex Stainless Steels in Light Water Reactor Systems: Timothy Lach1; Shawn Riechers1; David Collins1; William Frazier1; Arun Devaraj1; Emily Barkley1; Thak Sang Byun1; 1Pacific Northwest National Laboratory
Multicomponent alloys like cast austenitic stainless steels (CASSs) with a duplex phase structure allow for the investigation of interdependent microstructural evolution mechanisms during thermal aging or irradiation, where simpler model systems may not show the co-dependency of different evolution pathways. We aim to provide a comprehensive knowledge base for the integrity assessment of CASS components for extended operations in nuclear power systems, with a focus on degradation mechanisms in multiple alloys at reactor-relevant temperatures of 290-400°C up to 30,000 hours. A comprehensive evaluation of the microstructural and mechanical behavior evolution of the ferrite and austenite was conducted. The microstructure was studied using atom probe tomography and electron microscopy. The mechanical behavior was investigated on two scales: on the bulk scale using tensile, Charpy impact, and fracture toughness testing, and on the nanoscale using relative hardness mapping via atomic force microscopy. The results were further interpreted by kinetic Monte Carlo simulation.
11:30 AM
Mechanical Properties of Irradiated Cladding Material from BOR-60 Irradiations: Tarik Saleh1; Benjamin Eftink1; Stuart Maloy1; Gary Was2; 1Los Alamos National Laboratory; 2University of Michigan
Multiple engineering materials were irradiated in the BOR-60 reactor to doses of 14 dpa to 35 dpa, at a number of temperatures ranging from 375 to 525 C, as part of the Simulating Neutrons with Accelerated Particles (SNAP) project, funded through the U.S. Department of Energy, Office of Nuclear Energy. The purpose of this project is to compare fast reactor generated neutron irradiation damage to that generated from ion irradiation. The samples were in the form of 3mm diameter by 0.25mm thick TEM specimens. Mechanical properties were measured by shear punch test method, yielding effective shear stress vs displacement curves. This data can be correlated to and compared to yield and UTS data generated from tensile tests. The alloys HT9, T91, 14YWT, and 800H were tested, along with a number of model alloys. This data will be compared to control specimens, as well as samples from other irradiations.
11:50 AM
Comparison of In-situ Micro- and Ex-situ Meso-scale Tensile Testing of As-fabricated HT-9 Steels: Tanvi Ajantiwalay1; Assel Aitkaliyeva1; Peter Hosemann2; 1University of Florida; 2University of California, Berkeley
HT-9 steels have lower swelling rate, which makes them a preferred structural material for Generation IV reactors. It is known that microstructural defects can cluster and degrade the overall mechanical properties over the material’s lifetime in a reactor environment. Thus, it is imperative to establish microstructure-property relationships for these alloys and establish the fundamental mechanisms governing material failure. In-situ micro-tensile tests have the potential to improve statistics necessary to evaluate material performance, while reducing the operator dose. However, the representativeness of this technique and its correlation with the meso-scale tests have not been established. In this study, the micro- and meso-tensile properties of as-fabricated HT-9 steels are correlated with the microstructure of the steels using techniques such as, in-situ micro-tensile testing, scanning electron microscopy, and electron backscattering diffraction (EBSD). The yield and ultimate tensile stress of HT-9 calculated from both scales are compared and representativeness of micro-tensile testing is accessed.
12:10 PM Cancelled
Effects of Irradiation Defect and Strain on the Morphology and Kinetics Evolution of Nanoscale Phase: Yongsheng Li1; Zhengwei Yan1; Xinwen Tong1; Dong Wang1; 1Nanjing University of Science and Technology, Nanjing
The microstructure of the alloy can be affected by the irradiation defects and stress state. The decomposition rate and morphology evolution of the precipitates will be changed when alloys serve in irradiation environment. The transformation kinetics and microstructures of nanoscale precipitates suffered irradiation point defect and strain was studied by modified continuous phase-field model. Vacancies and interstitial atoms provide a priority position for phase precipitation, thus accelerate phase decomposition and accumulate to a loop at the phase boundary. Similar as the energy induced by point defects, the strain energy also promotes phase decomposition, and changes the morphology into strip-like under mixed triaxial strain. The effects of irradiation defects and strain on phase decomposition kinetics were revealed by means of micro-morphology and quantitative changes of particle size and volume fraction.