Mechanical Behavior of Nuclear Reactor Materials and Components III: Ferritic Alloys III
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
Tuesday 8:00 AM
March 21, 2023
Room: 28D
Location: SDCC
Session Chair: Caleb Massey, Oak Ridge National Laboratory; Eda Aydogan, METU
8:00 AM Invited
A New Microcrack Healing Mechanism in an Annealed 14YWT Nanostructured Ferritic Alloy: Md Ershadul Alam1; Soupitak Pal1; Nicholas Cunningham1; G. R. Odette1; 1University of California, Santa Barbara
Pre-existing microcracks form in a deformation processed 14YWT nanostructured ferritic alloy by well understood mechanisms, and that may hinder fabrication of defect-free components. Examined crack healing by short time, high temperature anneals and identified a new healing mechanism that involves the redistribution and coarsening of an initially small TiCON inclusions to coarser particles on crack faces, which locally displace matrix atoms to efficiently fill and heal the cracks. Crack healing improved the total room temperature tensile elongation from ≈ 0% for as-processed to ≈ 21% for the 1300ºC/5h annealed condition in tests with loading normal to the crack plane. Remarkably, this very efficient mechanism even healed macroscopic through-thickness fatigue cracks. This concept can be applicable to processing and self-healing of other materials at long time service conditions.
8:30 AM
Preliminary Studies on Creep Behavior of Commercial FeCrAl Alloy (APMT): Hamdy Abouelella1; Benjamin Beeler1; Jacob Eapen1; Korukonda Murty1; 1North Carolina State University
FeCrAl alloys are being considered accident-tolerant fuel-cladding for the current light water reactors (LWRs). The chemical passivity of FeCrAl in a water-steam environment allows a larger time window for accident response. In this work, the high-temperature creep behavior of a commercial FeCrAl alloy (Kanthal APMT) is investigated. Uniaxial tensile creep tests are conducted at four different temperatures (575℃, 600℃,625℃,650℃) and over a range of stresses from 50 MPa to 400 MPa. Transitions in creep mechanisms are noted with a stress exponent of ~3 at low stresses, which increases to ~8 at intermediate stresses, and finally to ~14 at high stresses. While precipitation hardening is believed to be the rate-controlling mechanism in the high-stress regime, dislocation glide controls the creep at the lower stresses.
8:50 AM
Investigating Environmentally-Assisted Cracking in 316 Stainless Steel U-Bend Specimens Exposed to Liquid Sodium: Dustin Mangus1; Xavier Quintana1; Guillaume Mignot1; Wade Marcum1; Julie Tucker1; Samuel Briggs1; 1Oregon State University
The combination of mechanical and environmental stressors tends to accelerate degradation of structural materials in nuclear reactors and other extreme environments, which can be life-limiting in these systems. Despite this, these phenomena are not well-studied for advanced reactor systems, such as sodium fast reactors. Oregon State University has recently developed the Glovebox for Experimental Liquid Sodium (GELS) facility to simulate prototypical SFR environments to enable for a wide array of materials testing. In initial tests, 316 stainless steel U-bend specimens have been immersed in an oxygen-controlled sodium environment to assess the effects corrosion and liquid metal embrittlement on the strained samples. The capabilities of OSU’s GELS facility and details of efforts to develop sodium prewetting capabilities will be highlighted. In addition, post-mortem characterization of as-received and immersed U-bend samples will be analyzed to evaluate the microstructural evolution and performance of strained austenitic stainless steels in SFR environments.
9:10 AM
Effect of Irradiation on the Tensile Strength of Select Layers and Layer Interfaces of TRISO-coated Nuclear Fuel Particles: Tanner Mauseth1; Mary Lou Dunzik-Gougar1; Fei Teng2; Subhashish Meher3; 1Idaho State University; 2Idaho National Laboratory ; 3Idaho National Laboratory
Tri-structural isotropic (TRISO) coated nuclear fuel particles are of interest for new reactor designs; however, knowledge gaps exist about mechanical properties of particle coating layers. To fill this gap, micro-scale tensile strength testing of layers is being conducted. Samples from particles are prepared via FIB micromachining and tested in-situ SEM. Of particular interest are buffer carbon and IPyC layers and their interface. To date, all interface samples fractured inside the buffer layer. Buffer and IPyC layer testing confirms the buffer is weaker. Interface samples testing indicates they have greater average tensile strength than buffer samples. Some interlayer samples displayed stress-strain behavior similar to IPyC. This behavior suggests the IPyC layer may play a larger role in buffer-IPyC interface properties than previously suspected. Surrogate fuel layer samples have been tested and pre-irradiated fueled and irradiated fueled samples are currently under investigation. Complete results will be available at the symposium.
9:30 AM Break
9:50 AM
Fatigue Assessment of Metastable Austenitic AISI 347 Pipe Components for Nuclear Engineering: Kai Donnerbauer1; Tobias Bill2; Peter Starke2; Ruth Acosta3; Christian Boller3; Jens Arndt4; Klaus Heckmann4; Jürgen Sievers4; Tim Schopf5; Frank Walther1; 1TU Dortmund, Chair of Materials Test Engineering (WPT); 2University of Applied Sciences Kaiserslautern, Department of Materials Science and Materials Testing (WWHK); 3Saarland University, Chair of Nondestructive Testing and Quality Assurance (LZfPQ); 4Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH; 5University of Stuttgart, Materials Testing Institute (MPA)
In German nuclear power plants AISI 347 steel is frequently used for pipe components. The transfer of results from fatigue tests to component service life is always accompanied by an uncertainty. Therefore, pipe components manufactured from a pipe originally intended for use in a nuclear power plant, were characterized using a setup which allows pressures up to 80 bar and temperature cycling between 20 and 300 °C. These investigations were supported by instrumented fatigue tests on conventional specimens from same pipe and bars of same material. During all types of tests various non destructive testing methods were applied, to understand which are suitable to indicate and separate damage mechanisms in metastable austenitic steels. It could be shown, that the testing strategy performed and measurement techniques applied allow reliable calculations of (remaining) fatigue life for AISI 347 steel.
10:10 AM
ODS Cu Materials for Fusion Application Produced by Mechanical Alloying: Carsten Bonnekoh1; Andrei Galatanu2; David Bürger3; Thomas Gietzelt1; Michael Rieth1; 1Karlsruhe Institute of Technology; 2National Institute of Materials Physics; 3Ruhr-University Bochum
Under the exposure to fusion neutrons, the current baseline copper alloy for the divertor heatsink (CuCrZr) suffers a loss of microstructural stability drastically lowering the bearable material temperature. Applying oxide-dispersion strengthened (ODS) Cu alloys is a promising attempt to overcome this dilemma. For this study, ODS Cu alloys were industrially produced using mechanical alloying (MA), which significantly differentiates these alloys from ODS Cu materials prepared by the internal oxidation method (better known under the trademark Glidcop). Here, we show the influence of alloy composition and alterations in parameters of MA on yield strength and toughness. Improvements in yield strength are considered with regard to changes in thermal conductivity. Preliminary results indicate that well-designed solid solution hardening in addition to ODS hardening is beneficial for the aggregated material performance. All results are discussed in the context of data for Cu1Cr0.1Zr and Glidcop Al-60 also acquired within this study for optimum comparability.
10:30 AM
Additively Manufactured Digital Image Correlation for Nuclear Materials: Kaelee Novich1; Timothy Phero1; Sarah Cole1; Michael McMurtrey2; David Estrada1; Brian Jaques1; 1Boise State University; 2Idaho National Laboratory
Real-time monitoring of materials in harsh environments is a crucial technique towards reducing innovation time in nuclear systems. Non-contact sensors, such as digital image correlation (DIC), offer a less destructive way to measure deformation of materials compared to alternative methods of in-situ strain determination, such as weldable strain gauges. However, DIC depends on high contrast surfaces, which often relies on the implementation of artificial patterns. Traditional splatter techniques have limitations, including poor surface adhesion and poor reproducibility. In this work, we have implemented additive manufacturing solutions to avoid such limitations. Accordingly, aerosol jet printing (AJP) was used to print small scale periodic patterns of silver on stainless steel and aluminum tensile specimens. DIC was employed to monitor strain (up to 1100 µε) during temperature cycling from 23-600 °C. The results validated the use of AJP to better control pattern parameters for small fields of view at high temperatures.