Tackling Structural Materials Challenges for Advanced Nuclear Reactors: Mechanical Behaviors
Sponsored by: TMS Corrosion and Environmental Effects Committee, TMS Nuclear Materials Committee, TMS: Advanced Characterization, Testing, and Simulation Committee
Program Organizers: Miaomiao Jin, Pennsylvania State University; Xing Wang, Pennsylvania State University; Karim Ahmed, Texas A&M University; Jeremy Bischoff, Framatome; Adrien Couet, University of Wisconsin-Madison; Kevin Field, University of Michigan; Lingfeng He, North Carolina State University; Raul Rebak, GE Global Research
Wednesday 2:00 PM
October 12, 2022
Room: 330
Location: David L. Lawrence Convention Center
Session Chair: Xing Wang, Pennsylvania State University
2:00 PM
Deformation Behaviour of Ion-irradiated FeCr – A Nanoindentation Study: Kay Song1; Hongbing Yu2; Phani Karamched1; Kenichiro Mizohata3; David Armstrong1; Felix Hofmann1; 1University Of Oxford; 2Canadian Nuclear Laboratories; 3University of Helsinki
Understanding the mechanisms of plasticity in structural steels is essential for the operation of next-generation fusion reactors. The study of ion-irradiated FeCr alloys is key to gaining a mechanistic understanding of irradiation damage in steels. This work on the deformation behaviour of FeCr, focusses on distinguishing between the nucleation of dislocations to initiate plasticity, from their propagation through the material. The effect of Cr content and irradiation dose on these dislocation mechanisms were examined through nanoindentation with a range of tip radii. Our results suggest that both Cr and irradiation-induced defects reduce the barrier for dislocation nucleation. However, pre-existing defects appear to be the dominant source. Yield strength, an indicator of dislocation mobility, increased with irradiation damage and Cr content, while work hardening capacity decreased. The synergistic effects of Cr and irradiation damage in FeCr appear to be more important for the propagation of dislocations than for their nucleation.
2:20 PM Invited
Mechanical Behavior of Additively Manufactured Steels with Monotonic and Graded Microstructures: Thak Sang Byun1; Maxim Gussev1; Timothy Lach1; 1Oak Ridge National Laboratory
Mechanical behaviors of additively manufactured (AM) single-phase austenitic stainless steel (316L) and compositionally (microstructurally) graded alloy (CGA) were investigated in various conditions using in-situ SEM, ex-situ tensile testing and TEM. Uniaxial tensile testing before and after irradiation was performed for AM 316L to assess its performance as a reactor core structural material. The AM 316L stainless steel, regardless of post-build treatment, showed higher strength and comparable ductility compared to the wrought 316L stainless steel. It also demonstrated high resistance to the neutron irradiation at 300 and 600°C. For the CGA build, the electron back scattered diffraction (EBSD) maps display a clear microstructural transition from an austenite dominant structure, to a complex composite structure containing ferrite, martensite, and austenite, and then to a fully ferritic structure. Deformation mechanism was highly dependent on the local composition. In particular, the Ni/Mn-rich austenite showed a complex deformation-induced martensitic transformation involving epsilon martensite.
2:50 PM Invited
Computer Modeling of Oxidation-induced Grain Boundary Embrittlement in Nickel: Ziqi Xiao1; Xian-Ming Bai1; 1Virginia Polytechnic Institute and State University
Nickel (Ni) based alloys are not only widely used structural materials for current light water reactors, but also are important candidate structural materials for future advanced reactors. However, Ni based alloys are susceptible to stress corrosion cracking (SCC) in corrosive environment. During SCC, the preferential corrosion at grain boundaries (GBs) reduces their strength and thus leads to brittle intergranular fracture under tensile loads. In this work, density functional theory (DFT) modeling is used to study how the GB oxidation level, the oxygen incorporation type at GB, and GB character influence the degradation of GB strength. The atomistic modeling results are then used to inform finite-element-method (FEM) based cohesive zone modeling to study how the oxidized GBs impact the intergranular fracture propagation under tensile loads in a polycrystalline Ni.
3:20 PM Break
3:40 PM
Investigation of Fracture Behavior of Nuclear Graphite NBG-18 Using In-situ Mechanical Testing Coupled with Micro-CT: Gongyuan Liu1; Yichun Tang1; Jing Du1; Aman Haque1; 1Penn State University
This paper presents the results of an experimental investigation on fracture behaviors of NBG-18 nuclear graphite through three-point bend tests. Four single-edge notched beam (SENB) specimens were prepared according to the suggested dimensions in ASTM D7779-20. Three-point bending tests on three SENB specimens were performed using a mechanical tester. The fracture toughness was measured to be 1.186 - 1.240. An incremental three-point bending test was conducted on the remaining specimen using a mechanical tester coupled with micro-CT. Oriented lenticular thermal cracks and round gas entrapment pores in the reconstructed tomographic cross-sectional images are used to identify the filler and binder phases. racks were observed to originate primarily from larger open pore structures on the edge of the notch. Then, the cracks propagated along the paths of the larger open pores. Crack bridging by uncracked ligament, crack deflection, and crack blunting were observed as toughening mechanisms.
4:00 PM Invited
Multi-scale Modeling of the Mechanical Response of Structural Metals Subjected to Thermo-mechanical Loads and Irradiation: the Role of Microstructure.: Laurent Capolungo1; Arul Kumar1; Andrea Rovinelli1; Ricardo Lebensohn1; 1Los Alamos National Laboratory
Structural metals considered for use in advanced nuclear reactors will be subjected to extreme environments leading to rapid and profound changes in the microstructure, and potentially of the composition, of the metal. Mechanistic models can be utilized to extrapolate the overall mechanical behavior of the material and work in unison with experiments to fully comprehend materials performance. The work to be presented introduces an advanced polycrystal modeling framework, applied both to the case of ferritic and austenitic steels, that can simultaneously predict the overall tensile and creep response (primary secondar and tertiary) of representative materials volumes. Uniquely the constitutive model captures the absolute and relative contributions of a vast array of deformation mechanism (e.g. dislocation glide, dislocation climb, point defect diffusion mediated plasticity) as a function of the fingerprint of the microstructure -such is defined by the dislocation content, arrangement, solute, non-equilibrium point defect and precipitate content-.
4:30 PM Invited
Microplasticity of Irradiated Inhomogeneous Alloys: Anter El-Azab1; Yash Pachaury1; 1Purdue University
We report on a preliminary effort to understand interaction between alloy microplasticity and compositional inhomogeneity from a dislocation dynamics perspective. We tackle this problem by a multiscale simulation in three steps: (1) Analysis of the 3D composition morphology in alloys with tendency to undergo spinodal instability both thermally and under irradiation, with bcc FeCrAl alloys as a model system, (2) atomistic simulation of the dislocation mobility as a function of the local composition, and (3) using dislocation dynamics simulations to understand the impact of composition inhomogeneity and coherency strain on microplasticity. In this presentation, we give a quantitative analysis of the dislocation microstructure differences among the inhomogeneous and homogeneous alloys and the dependence of alloy strength on the wavelength and amplitude of the alloy composition fluctuations.