Mechanical Behavior of Nuclear Reactor Materials and Components III: Zr Alloys and Beyond
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee, TMS: Mechanical Behavior of Materials Committee
Program Organizers: Assel Aitkaliyeva, University of Florida; Clarissa Yablinsky, Los Alamos National Laboratory; Osman Anderoglu, University of New Mexico; Eda Aydogan, Middle East Technical University; Kayla Yano, Pacific Northwest National Laboratory; Caleb Massey, Oak Ridge National Laboratory; Djamel Kaoumi, North Carolina State University
Wednesday 2:00 PM
March 22, 2023
Room: 28D
Location: SDCC
Session Chair: Caleb Massey, Oak Ridge National Laboratory; Djamel Kauomi, North Carolina State University
2:00 PM Invited
The Role of Stress-State on the Failure Mechanism, Strain to Failure and Fatigue Resistance of Zircaloy-4: Brian Cockeram1; Kwai Chan2; Bruce Kammenzind1; 1Nnl Fluor Marine Propulsion; 2Southwest Research Institute
The strain to failure for Zircaloy-4 under monotonic loading and cyclic loading is affected by stress-state, hydrogen, and temperature. Stress-state, which can be considered a measure of constraint, was quantified using triaxiality, which is defined as the ratio of hydrostatic stress to von Mises stress. The increase in triaxiality from 0.33 for smooth tensile specimens to 0.9 for notched specimens, along with the presence of hydrides, produces a decrease in both the strain to failure and fatigue resistance. An increase in strain to failure and fatigue resistance is observed at lower values of zero triaxiality under shear loading with larger increases observed in compression. Hydrogen is shown to have a relatively small effect on the results for specimens with low triaxiality values. In-situ and ex-situ testing of notched Zircaloy-4 specimens with varying levels of hydrogen and stress state was performed, and these results are used to show this behavior is the result of the change in the mechanisms for deformation and fracture.
2:30 PM
Impact of Thermal Treatment and Irradiation on Mechanical Behavior of Cold Spray Cr Coatings on Zr-alloy Cladding: Tyler Dabney1; Hwasung Yeom1; Nan Li2; Ben Eftink2; Kumar Sridharan1; 1University of Wisconsin-Madison; 2Los Alamos National Laboratory
Chromium-coated zirconium-alloy cladding is being considered for implementation in light water reactors for its superior oxidation resistance during high-temperature accident scenarios and improved corrosion kinetics during normal operating conditions. From reactor insertion through end-of-life, the coated cladding will undergo thermal- and irradiation-induced microstructural changes, and temperatures exceeding 1000 °C in the event of an off-normal condition. These environmental effects will cause changes to the mechanical behavior of the cladding due to stress relief, recrystallization and grain growth, irradiation-induced defect formation, or high-temperature diffusion leading to intermetallic compound formation at the interface. This study investigates the mechanical behavior of the coating and coating-substrate interface system after annealing at multiple temperatures as well as after ion irradiation, using a picoindenter housed in an SEM and the miniature cantilever techniques. This study is supported by post-test characterization of the deformed microstructure using SEM, TEM, and EBSD.
2:50 PM
Hydride Reorientation Behavior in ZIRLO Using Ring Compression Tests: Soyoung Kang1; Arthur Motta1; Maxim Gussev2; Michael Billone3; 1Pennsylvania State University; 2Oak Ridge National Laboratory; 3Argonne National Laboratory
The hydrogen picked up during corrosion of nuclear fuel cladding can precipitate as zirconium hydrides when the terminal solid solubility is exceeded. When cooling is performed under sufficient stress, zirconium hydrides can macroscopically re-orient along the cladding radial direction. This hydride reorientation severely degrades cladding ductility. The purpose of this study is to observe the effect of alloy microstructure (i.e., grain shape and grain size) on hydride reorientation especially comparing ZIRLO and Low Tin ZIRLO. These materials have similar crystallographic texture and chemistry, but different microstructures due to the different heat treatments used. Ring compression tests (RCT) were used to apply stress to make reoriented hydrides. Finite element modeling (FEM) was performed to analyze the stress state in the ring samples. From FEM and microstructural characterization, it was possible to determine the threshold stress of hydride reorientation for both materials and the effects of microstructure on hydride reorientation are shown.
3:10 PM
Anisotropic Compressive Strength of Single Crystal Zirconium Pillars and the Effects of Irradiation Hardening and Temperature Through Micro-Pillar Mechanical Testing: Matthew deJong1; Philip Alarcón-Furman1; Ryan Schoell2; Djamel Kaoumi1; 1North Carolina State University; 2Sandia National Laboratories
Zirconium based alloys are widely used in the nuclear industry, due to their low capture cross section for thermal neutrons, and relatively good corrosion resistance. However, as a result of its HCP crystal structure, Zirconium has anisotropic mechanical properties and response to irradiation. In this study, using a single zirconium crystal of known crystallographic direction, micro-pillars were processed with focused ion beam in different orientations corresponding to different compression directions. The pillars were irradiated to 5 dpa with 1 MeV Kr ions, and subjected to in situ TEM compression testing at room temperature and 300C to determine the effects of anisotropy on the temperature and irradiation response of zirconium in terms of the mechanical properties. The suitability of the in-situ micro-mechanical testing to study mechanical property anisotropy is demonstrated and discussed.
3:30 PM Break
3:50 PM
Mechanical Behavior of Bare and Cr Coated Zirconium Claddings During Burst Testing via In-situ Strain Measurements: Samuel Bell1; Mackenzie Ridley2; Kenneth Kane3; Ben Garrison2; Tim Graening2; Nathan Capps2; 1University of Tennessee Knoxville; 2Oak Ridge National Laboratory; 3John Hopkins University - Applied Physics Laboratory
Cr coated zirconium alloys are a leading concept to replace the incumbent bare zirconium fuel claddings in light-water reactors. Cr coatings have demonstrated greater high temperature steam oxidation resistance, as well the potential for improved mechanical response during accident scenarios. Before this concept can be widely deployed, a better understanding of coated cladding behavior during accident scenarios is necessary. Recent efforts to improve burst testing, an established method of assessing cladding in simulated accident conditions, have integrated digital image correlation techniques to measure in-situ strain during rapid heating of pressurized cladding materials until rupture. In this work, digital image correlation techniques were applied to monitor both Cr coated and bare Zircaloy-4 cladding segments during burst testing. In-situ strain measurements and mechanical behavior of bare and Cr coated Zircaloy-4 will be compared.
4:10 PM
Cladding Coating Integrity Quantified by Ring Pull and Local Strain Analysis: Peter Beck1; Mathew Hayne1; Emily Proehl1; Samuel Briggs2; Julie Tucker2; Tarik Saleh1; Benjamin Eftink1; 1Los Alamos National Laboratory; 2Oregon State University
Evaluating the impact of hoop direction strains on thin-walled tubes is a critical step before employing potential nuclear reactor fuel cladding concepts. This is particularly important for Accident Tolerant Fuel (ATF) claddings with coatings to ensure adherence of the coating during service. In this work, we will present developments extending the strain analysis of ring pull testing, a simple to perform hoop direction mechanical test that uses small segments of cladding tube. A novel, LANL developed, Python-based analysis package is combined with digital image correlation and used for calculating the strain field on the rings to correlate local material response to strain. This is applied to measuring the mechanical integrity of Cr coatings on Zircaloy-4 tube. This coating integrity and response to strain will be related to the microstructure at the coating/clad interface.
4:30 PM
Structure-property Evolution of PM-HIP Fabricated Ni-Alloys 625 and 690 Neutron Irradiated to 1 and 3dpa: Caleb Clement1; Yu Lu2; Sheng Cheng2; Megha Dubey2; Sowmya Panuganti1; Yangyang Zhao1; Katelyn Wheeler3; Donna Guillen3; David Gandy4; Janelle Wharry1; 1Purdue University; 2Boise State University/ Center for Advanced Energy Studies; 3Idaho National Laboratory; 4Electric Power Research Institute
The objective of this talk is to understand irradiation dose effects on the microstructure and mechanical properties of Alloy 625 (Ni-23Cr-8Mo) and Alloy 690 (Ni-31Cr-10Fe) fabricated using powder metallurgy with hot isostatic pressing (PM-HIP). PM-HIP presents an attractive alternative to traditional manufacturing methods for nuclear applications; however its irradiation performance is not yet understood. Alloys 625 and 690 fabricated with PM-HIP and by forging were neutron irradiated to 1 and 3 displacements per atom (dpa) at 400ºC in the Advanced Test Reactor. Post-irradiation transmission electron microscopy and atom probe tomography reveal relatively consistent dislocation loop and nanoprecipitate evolution across the PM-HIP and forged materials, with the size of these defects scaling with dose. Nanoindentation and tensile testing reveal that Orowan strengthening can accurately capture structure-property relationships of both alloys at both doses. These results further our understanding of irradiation effects on PM-HIP alloys towards nuclear code-qualification.