Advanced Characterization of Materials for Nuclear, Radiation, and Extreme Environments: Radiation Effects & Materials Mechanics
Sponsored by: TMS Nuclear Materials Committee
Program Organizers: Samuel Briggs, Oregon State University; Christopher Barr, Department Of Energy; Emily Aradi, University of Huddersfield; Michael Short, Massachusetts Institute of Technology; Janelle Wharry, Purdue University; Cheng Sun, Clemson University; Dong Liu, University of Oxford; Khalid Hattar, University of Tennessee Knoxville
Wednesday 8:00 AM
November 4, 2020
Room: Virtual Meeting Room 13
Location: MS&T Virtual
Session Chair: Christopher Barr, Sandia National Laboratories; Janelle Wharry, Purdue University
8:00 AM
Benefits of Using High Energy Ions in Ion Irradiation Experiments to Evaluate Void Swelling: Peter Doyle1; Takaaki Koyanagi2; Steven Zinkle1; 1University of Tennessee; 2Oak Ridge National Laboratory
Because of its cost, speed, and convenience for post-irradiation examination, ion irradiation is a common technique for evaluation of irradiation effects in nuclear materials. However, ion irradiations require very careful planning and detailed knowledge of potential artifacts. Specifically, in void swelling studies, near the surface and at the end of the irradiation-affected zone, void formation is suppressed due to the surface sink and implanted ions. This work examined irradiation of pure Cr with 15MeV Ni5+ ions to midrange damage of 1, 8, and 22 dpa. Void swelling was shown to saturate at less than 3%. An artificial peak in void diameter was observed adjacent to the void-denuded zone, ~500nm from the surface and true swelling was not observed until a depth of >1µm, emphasizing the need for high energy ions in irradiation studies. Funding was provided by the U.S. Department of Energy Office of Nuclear Energy, Advanced Fuel Campaign.
8:20 AM
Swelling of Nuclear Reactor Steels: Modeling, Theory, and Accelerated Testing
: Michael Fluss1; Edward Moses2; 1Nuclear Materials Consultancy; 2Longview Consulting, Inc.
Understanding and predicting in-service changes and degradation of the internal materials in commercial nuclear energy facilities is important with respect to safety, licensing, operations, sustainability, and economics. Swelling is of particular importance with regards to predicting in core material service lifetimes. It is well known that heavy ions produce atomic displacement damage at rates five to six orders of magnitude greater than reactor neutrons. Although neutron induced swelling for nuclear reactor steels was discovered over 50 years ago there is still no accepted engineering methodology for accelerated studies using heavy ion beams. To fully take advantage of this highly accelerated testing for swelling requires an analytical methodology that accurately predicts swelling over a range of displacement rates from 10exp-2 to 10exp-9 displacements per atom per second. Here we will outline such a parametric methodology and its rate theory underpinnings.
8:40 AM
Response of an Additively Manufactured 316 Stainless Steel Subjected to High Temperature Heavy Ion Irradiations: Zhongxia Shang1; Cuncai Fan2; Jie Ding1; Sichuang Xue1; Adam Gabriel3; Thomas Voisin4; Jin Li1; Lin Shao3; Yinmin Wang4; Haiyan Wang1; Xinghang Zhang1; 1Purdue University; 2Oak Ridge National Laboratory; 3Texas A&M University; 4Lawrence Livermore National Laboratory
Additive manufacturing has become an appealing technique to fabricate three-dimensional metallic materials and components for nuclear reactors. However, response of additively manufactured alloys to high-dose heavy ion irradiations at elevated temperatures is still not well understood. Here, an additively manufactured 316 austenitic stainless steel with high-density solidification cells was irradiated using Fe ion to a peak dose of ~ 200 dpa at 450 ˚C. Microscopy studies show a smaller Frank loop density and size in the additively manufactured sample compared with its cold worked counterpart, and the cellular structures may largely suppress the formation of perfect loops and dislocation networks and reduce the magnitude of solute segregations than that along high angle grain boundaries. The present work advances the understanding on the high-temperature irradiation response of additively manufactured steels for nuclear reactor applications.
9:00 AM
In situ Crack Loading and Measurement Techniques for Gen IV Reactor Coolant Media: Peter Beck1; Andrew Brittan2; Dustin Mangus1; Jake Quincy1; George Young2; Guillaume Mignot1; Samuel Briggs1; Julie Tucker1; 1Oregon State University; 2Kairos Power
Next-generation nuclear reactors are seeking to employ advanced coolant media, such as liquid metals, molten salts, or pressurized gases. However, these fluids present corrosion and failure mechanisms unique to each system. Experimental techniques must be developed to understand and predict the performance of structural materials throughout the lifetime of reactor operation. This work describes how loading and diagnostic methods commonly employed in aqueous environments have been adapted to characterize environmentally assisted cracking in advanced reactor coolant environments. An internally pressurized metal bellows has been used to apply a load to a compact tension (CT) specimen without the use of a pull rod or load cell. Additionally, the reversed direct current potential drop (DCPD) technique is applied to measure crack propagation in prolonged sample immersions in these advanced fluids. Initial results and suggestions for employing these techniques in various extreme energy environments will be discussed.
9:20 AM
Characterization of Stress and Environment Dependent Fracture Mechanisms of SiC/SiC CMCs: Morgan Price1; Clifton Bumgardner1; Frederick Heim1; David Roache1; Xiaodong Li1; 1University of Virginia
Silicon carbide fiber reinforced silicon carbide matrix (SiC/SiC CMCs) composites, which are used in nuclear reactor applications for superior high temperature and corrosion resistant properties, exhibit degraded mechanical strength and embrittlement under sustained or slowly ramping stress levels. The morphology and distribution of microcracks was evaluated during tension and expanding plug tests with variable strain rates and levels of environmental exposure (up to 1200 °C in inert and oxidizing environments). All tests were conducted with in-situ monitoring of local deformation using stereoscopic digital image correlation and acoustic emissions to characterize microcrack growth. Experimental crack growth mechanisms, including crack opening displacement with stress level, were used to validate a finite element model to predict the fracture energy release. Thus, this study aims to elucidate the coupled stress-based and environmental-based mechanisms driving crack initiation and propagation within SiC/SiC composites.