Mechanical Behavior of Nuclear Reactor Components: Microstructure Effects
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

Tuesday 2:00 PM
March 16, 2021
Room: RM 50
Location: TMS2021 Virtual


2:00 PM  
He Ion Irradiation Response of a Gradient T91 Steel: Zhongxia Shang1; Jie Ding1; Cuncai Fan2; Di Chen3; Jin Li1; Yifan Zhang1; Yongqiang Wang4; Haiyan Wang1; Xinghang Zhang1; 1Purdue University; 2Oak Ridge National Laboratory; 3University of Houston; 4Los Alamos National Laboratory
    Metallic materials with a gradient microstructure usually exhibit excellent mechanical properties. However, the radiation response of gradient structural materials is less well understood. Here, room-temperature He ion irradiation up to ~ 4.5 dpa with ~ 10 at.% He injection was performed on a gradient T91 steel processed by surface severe plastic deformation. In comparison to the coarse-grained ferritic T91, the gradient T91 with the nanocrystalline layers shows improved radiation tolerance in terms of less bubble swelling and radiation hardening. Additionally, bubble distribution along grain boundaries depends on misorientation angle suggesting a strong capacity of non-equilibrium grain boundaries in storing He atoms. The present study provides insight into the design of radiation tolerant gradient steels for nuclear industry applications.

2:20 PM  
High Temperature Strength of Additively Manufactured Gr91 Steel: Benjamin Eftink1; Daniel Vega2; Osman El Atwani1; David Sprouster3; Carl Cady1; Mohamad Al-Sheikhly4; Thomas Lienert5; Stuart Maloy1; 1Los Alamos National Laboratory; 2DOE; 3Stony Brook University; 4University of Maryland; 5T.J. Lienert Consulting, LLC
    As-deposited additively manufactured Grade 91 steel had excellent mechanical properties with greater strength than wrought Grade 91 steel at room temperature, 300 and 600 C showing excellent promise for nuclear applications. Retention of strength at 300 and 600 C for the as-deposited additively manufactured material was attributed to transitional carbides in lower bainitic microstructural regions. As-deposited additively manufactured Grade 91 steel had a microstructure of lower bainitic regions surrounded by martensite. This is significantly different from the typical tempered martensitic microstructure of conventionally produced Grade 91 steel. In this talk the relationship between microstructure and mechanical properties will be discussed for the Grade 91 steel processed by additive and conventional methods.

2:40 PM  Invited
Wear Behavior of Incoloy™ 800HT and Inconel™ 617 for High-Temperature Gas-cooled Reactor (HTGR) Applications: Valentin Pauly1; Joseph Kern1; Malcolm Clark1; David Grierson1; Kumar Sridharan1; 1University of Wisconsin-Madison
    Incoloy™ 800HT and Inconel™ 617 are metallic materials that have been selected for structural applications in high-temperature gas-cooled reactors (HTGRs) based on their ASME code certification. Wear of rubbing components in HTGRs such as valves, valve seats, and valve shafts at high temperatures is of concern. Impurities (e.g., H2O and CH4) in the helium coolant induce corrosion reactions at high temperatures, and tribological performance will be strongly affected by the corrosion product layers that form on the components’ surfaces. Here we investigate the tribological behavior of Incoloy™ 800HT and Inconel™ 617 in the temperature range of 650°C to 900°C. We find that the presence of small concentrations of oxygen-bearing impurities are beneficial for wear resistance as they promote the formation of a compacted, hard, low-wear glaze-oxide layer. Surface modification treatments, such as aluminization, are also investigated to promote the formation of a glaze layer and improve tribological performance.

3:10 PM  
Modeling the Effect of Helium Bubbles, Rigid Inclusions, and Grain Boundaries on Crack Initiation in Nickel: Tung Yan Liu1; Michael Demkowicz1; 1Texas A&M University
    In this work, we present molecular dynamics simulations of plastic deformation of nickel containing nano-scale helium bubbles, rigid inclusions, and grain boundaries. Our goal is to study how these microstructure features affect crack initiation. The models are built based on experimental characterization of defect structures in Ni-base alloy components irradiated during service in pressurized heavy water reactors. In particular, we examine dislocation nucleation and motion, deformation-induced changes in bubble shape, and slip band formation. We discuss the implications of our work for understanding the mechanism of deformation and fracture in irradiated materials, which aid lifetime predictions of nuclear reactor components.

3:30 PM  
Quantifying Zirconium Embrittlement Due to Hydride Microstructure Using Image Analysis: Pierre-Clement Simon1; Cailon Frank1; Long-Qing Chen1; Mark Daymond2; Michael Tonks3; Arthur Motta1; 1The Pennsylvania State University; 2Queen's University; 3University of Florida
    A fraction of the hydrogen produced by waterside corrosion of the zirconium nuclear fuel cladding material in light water reactors is picked up by the cladding, and precipitates into brittle hydrides once the solid solubility limit is reached. Embrittlement of the zirconium due to both circumferential and radial hydrides through the material thickness depends on the exact microstructure and connectivity of the hydride particle network. However, quantifying hydride microstructure is challenging, and several of the metrics currently being used have significant shortcomings. A new metric, the Radial Hydride Continuous Path (RHCP), has been developed using image analysis and implemented using a genetic algorithm to quantify hydride microstructures in relation to their potential effect on crack propagation through the cladding thickness. The RHCP is compared against existing metrics, demonstrating a more precise assessment of the effect of hydrides on cladding integrity.

3:50 PM  
In-situ Observations of the Failure Mechanisms of Hydrided Zircaloy-4 under Different Stress-States: Brian Cockeram1; Kwai Chan2; 1Naval Nuclear Laboratory-Bettis Laboratory; 2Southwest Research Institute
    In-situ testing of notched Zircaloy-4 specimens with varying levels of hydrogen was performed, and the deformation mechanisms were characterized. High-resolution Electron Backscattered Diffraction (EBSD) was used to quantify the distribution of residual stresses near the hydrides. The effect of local texture on the hydride distribution and fracture mechanism was observed. Stress-state is also shown to have a significant effect on the fracture mechanism. A process of void nucleation, growth, and coalescence has been observed for the non-hydrided condition. For hydrided materials, a similar mechanism is observed in the regions between the hydrides. Cracking of the hydrides may occur at local regions of high strain when the macroscopic deformation is elastic. The fracture of hydrides is strongly dependent on the residual stresses, local texture, and stress state. A micromechanical model is developed for hydrided Zircaloy-4 that accounts for the effects of stress-state, residual stress from hydride formation, local texture, and temperature.