Mechanical and Creep Behavior of Advanced Materials: A SMD Symposium Honoring Prof. K. Linga Murty: Materials for Nuclear Environments
Sponsored by: TMS Structural Materials Division, TMS: Mechanical Behavior of Materials Committee, TMS: Nuclear Materials Committee
Program Organizers: Indrajit Charit, University of Idaho; Yuntian Zhu, North Carolina State University; Stuart Maloy, Los Alamos National Laboratory; Peter Liaw, University of Tennessee - Knoxville

Wednesday 8:30 AM
March 1, 2017
Room: 24A
Location: San Diego Convention Ctr

Session Chair: Jacob Eapen, North Carolina State University; Stuart Maloy, Los Alamos National Laboratory

8:30 AM  Keynote
Enhanced Radiation Tolerance of Single Phase Solid Solution Alloys: Shi Shi1; Mo Rigen1; Shuai Wang1; Ian Robertson1; 1University of Wisconsin-Madison
    In this talk, which we dedicate to Professor Murty, we will show how alloy complexity leads to greater tolerance to irradiation damage but the relationship is multifaceted and does not depend simply on the number of elements in the alloy. Specifically, the evolution of damage in a series of Ni-containing single-phase solid solution alloys under electron and ion irradiation will be presented. The dynamics of the damage evolution as well as the mobility of the damage is captured through the use of in-situ irradiation and characterization in a transmission electron microscope. In addition to the defects produced it will be shown that the local compositional state of the alloys is impacted by the irradiations. These effects will be explained in terms of the modifications of the energy dissipation processes and defect formation and migration energies due to the compositional complexity of the alloys.

9:00 AM  Invited
Deformation and Fracture Behavior of Irradiated and Nonirradiated Austenitic Stainless Steels: Thak Sang Byun1; Maxim Gussev2; Timothy Lach1; 1Pacific Northwest National Laboratory; 2Oak Ridge National Laboratory
    Austenitic stainless steels are the most commonly used structural materials in nuclear reactors and their coolant systems. Understanding of the behavior of austenitic stainless steels in service has always been incomplete because too many deformation and fracture mechanisms can operate in the steels depending on their chemistry, defect structure, and operating environment. Irradiation, in particular, can produce various defect structures and result in a variety of deformation microstructures including tangled dislocations, dislocation pileups, stacking faults, twins, dislocation channels, and martensite particles. This presentation will attempt to integrate the mechanical properties and related microstructural phenomena observed in the steels before and after irradiation, and to elucidate mechanical properties and deformation mechanisms which are dependent on or independent of irradiation. Discussion will particularly focus on how the nearly-invariant material parameters such as stacking fault energy, plastic instability stress, and true fracture stress can be related to the varied microscopic mechanisms.

9:20 AM  Invited
A Rate Theoretic Approach to Modeling Irradiation Creep: Jacob Eapen1; 1NC State University
    The structural components in a nuclear reactor are expected to maintain their mechanical integrity for forty to sixty years. During reactor operation, they are subjected to fast neutron flux and moderate-to-high thermo-mechanical stresses. Aided by the high temperatures, the accumulated stresses from irradiation are relieved through irradiation creep. While stress relaxation is desirable, the dimensional changes due to creep are limited by the operational and safety design constraints. While historical data suggest that irradiation creep behavior is similar to that of thermal creep, it is often handled through ad-hoc assumptions. In this work, we apply the rate theory formalism to published creep data of two nuclear graphite grades. Interestingly, the mean activation energy for steady-state creep deformation under irradiation is observed to be rather low (<0.32 eV) with a mean stress exponent of ~1, indicating a diffusion-dominated mechanism.

9:40 AM  Cancelled
Anisotropic Biaxial Creep of Textured Nb-modified Zircaloy-4 Tubing: Nilesh Kumar1; Kaitlin Grundy1; Boopathy Kombaiah2; Baifeng Luan3; K Murty1; 1NC State University; 2Carnegie Mellon University; 3Chongqing University
    Biaxial creep of Nb-modified Zircaloy-4 (HANA4) tubing is investigated at varied ratios of hoop and axial stresses at a constant temperature of 500 °C using internal pressurization superimposed with axial load while monitoring the hoop and axial strains using non-contact laser telemetric extensometer and linear variable differential transducer, respectively. Steady-state creep rates along the hoop and axial directions were evaluated in the power-law creep regime from which the creep locus was derived at a constant energy of dissipation. The resulting creep locus was compared with that predicted by the anisotropy parameters, R and P in the Hills’ formulation for generalized stress for anisotropic materials. Crystallographic texture of the tubing was characterized using electron backscatter diffraction technique. Research is supported by NSF grant #DMR0968825.

10:00 AM Break

10:20 AM  Keynote
The Enhanced Radiation-resistance of Ultrafine-grained Metals Produced by SPD Processing: Ruslan Valiev1; Nariman Enikeev1; Marina Abramova2; Bertrand Radiguet3; Auriane Etienne3; Xavier Sauvage3; 1Laboratory for Mechanics of Bulk Nanomaterials, Saint Petersburg State University; 2Ufa State Aviation Technical University; 3Université et INSA de Rouen
    Ultrafine-grained (UFG) materials produced by severe plastic deformation (SPD) have a specific structure with extremely small grain size (tens to hundreds of nanometers) and dramatically increased volume fraction of grain boundaries and other nanostructured features. Grain boundaries are effective sinks for radiation-induced point defects, and there is reason to expect that the radiation resistance of the material can be significantly increased as a result of nanostructuring. The aggregate of high mechanical properties of NSM in conjunction with increased radiation stability can make them very attractive for use in advanced energy technologies, so that research in radiation stability of nanostructured materials is of high scientific and practical value. This report presents the results of studies and the development of enhanced radiation resistance of nanostructured model metals and austenitic stainless steels in terms of systematic and complex studies of changes in the structure and properties compared with the coarse-grained counterparts.

10:50 AM  Keynote
High Temperature Behavior of Zirconium Alloys in Air: Brian Jaques1; Jordan Vandegrift2; Patrick Price1; Jatuporn Burns1; Isabella van Rooyen3; Darryl Butt4; 1Micron School of Materials Science and Engineering, Boise State University; Center for Advanced Energy Studies; 2Micron School of Materials Science and Engineering, Boise State University; 3Idaho National Laboratory; 4Micron School of Materials Science and Engineering, Boise State University; Center for Advanced Energy Studies; University of Utah
    Zirconium alloys are used as structural and cladding materials in the nuclear industry. Under transient conditions, these materials will be exposed to air atmospheres at high temperatures. It is the purpose of this research to investigate the effects of transients on phase transformations, microstructure, and oxidation of Zircaloy 3 and Zircaloy 4 alloys. Phase transformations were studied by differential scanning calorimetry and dilatometry. Effects of transient conditions on the microstructure and composition were assessed by analysis of quenched specimens with electron microscopy and energy dispersive spectroscopy. Corrosion behavior was characterized with a combination of thermogravimetric techniques and oxide imaging. The performance of selected zirconium alloys under transient conditions has been compared.

11:20 AM  Invited
Synergistic Effects of Neutron Irradiation and Interstitial Nitrogen on Strain Aging in Ferritic Steels: Nilesh Kumar1; Ahmad Alsabbagh1; C. Seok2; K Murty1; 1NC State University; 2SungKyunKwan University
    Ferritic steels that are generally used in pressure vessels and various reactor support structures in light water reactors exhibit dynamic strain aging (DSA) resulting in increased work-hardening accompanied by ductility loss. While there is a possibility of adding this embrittlement known as blue brittleness to the well-known radiation embrittlement, it has been amply demonstrated that radiation exposure leads to decreased concentrations of interstitial impurity atoms in solution. Thus the critical temperature for DSA increases with increased neutron fluence very similar to the increase observed in dry hydrogen treated mild steel samples with decreased concentration of nitrogen in solution with increased treatment time. We summarize here the mechanical and fracture studies made on three different materials: a mild steel and two ferritic steels (A533B and A516 Grade70). In addition, effects of interstitial nitrogen are evaluated by heat treating to different times in dry hydrogen atmosphere.

11:40 AM  
Study of High Temperature Deformation Behavior of Graded Transition Joints (GTJs) (Relevance to Nuclear Power Plant Components): Mohan Subramanian1; Sudarsanam Babu1; Jonathan Galler2; John DuPont2; Xinghua Yu3; Zhili Feng3; 1University of Tennessee; 2Lehigh University; 3Oak Ridge National Laboratory
    Bimetallic weld made between 2.25Cr-1Mo steel (lower bundle component) and Alloy 800H (upper bundle component) using Inconel filler metal in the steam generator of Nuclear power plants fail prematurely in service. Simulated creep tests of this dissimilar weld configuration show that the failure occurs along the 2.25Cr-1Mo steel HAZ close to the weld interface due to the steep change in chemical composition and microstructure along the interface during service. Current research is focused on developing functionally Graded Transition Joints (GTJs), made by gradually changing the chemical composition in layers using Dual wire Gas Tungsten Arc Welding (DWGTAW) process. To study the creep deformation behavior of inhomogeneous GTJs, measuring overall creep strain using conventional extensometer will not be an appropriate methodology. Hence, Digital Image Correlation (DIC) was used to study the localized deformation response of GTJs.