Mechanical Behavior of Nuclear Reactor Materials and Components III: Ferritic Alloys I
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

Monday 8:30 AM
March 20, 2023
Room: 28D
Location: SDCC

Session Chair: Kayla Yano, PNNL; Eda Aydogan, METU


8:30 AM  Invited
Microstructure-aware Predictions of the Creep Response of Metals Subjected to Nuclear Environments: Laurent Capolungo1; Arul Kumar1; Aritra chakraborty1; Aaron Kohnert1; Andrea Rovinelli1; Ricardo Lebensohn1; 1Los Alamos National Laboratory
    Structural metals used for nuclear energy applications (Gen II to Gen IV) are subjected to extreme environments in the form of irradiation, temperature, stresses and corrosion. The challenge lies in anticipating, in the context of a data scarcity, how the initial state of the material (i.e., microstructure, initial dislocation content) will condition its mechanical response. This problem is complex due to the dynamic changes to the microstructure during service (e.g., irradiation induced defects, precipitation, etc.). This work will present a multi-scale modeling framework, connecting discrete dislocation dynamics simulations to a chemo-mechanical crystal plasticity-based polycrystal model (EVPFFT), to quantify the effects of microstructure on the creep performance (i.e., primary, secondary and tertiary) of metals in a wide spectrum of conditions. Focus will be placed on demonstrating the important role of diffusion mediated processes to creep. Applications will be brought on the case of HT9, Gr91 and other alloys.

9:00 AM  
Examining Microstructures and Mechanical Properties of Neutron and Ion Irradiated T91, HT9 and 800H Alloys: Pengcheng Zhu1; Shradha Agarwal1; Steven Zinkle1; 1University of Tennessee, Knoxville
    9-12% Cr ferritic-martensitic steel (T91, HT9) and austenitic alloy 800H are promising candidates for the structural components in Generation IV fission reactors. Alloys T91, HT9 and 800H were irradiated with neutrons (BOR60 reactor) and dual-ions (9MeV Fe3+ and 3.24MeV He2+) from 376 to 520 ℃ with damage from 16.6 to 72 dpa, to quantify the possibility of using ion irradiation to simulate neutron irradiation in microstructures and mechanical properties. Nanoindentation testing was performed to obtain the bulk hardness of the dual-ion irradiated samples due to limited damage volumes. For the neutron irradiated samples, both nanoindentation and Vickers hardness testing were conducted. An accurate way to extract bulk hardness from nanoindentation data will be used based on our previous study on FeCr alloys. The results of TEM characterization of the bubbles and dislocation loops will be provided to compare the predicted (dispersed barrier hardening) and measured strength of the irradiated specimens.

9:20 AM  
Irradiation and Nanomechanical Performance of Additively Manufactured, In Situ Tempered Grade 91 Steel: Calvin Lear1; Emily Proehl2; Todd Steckley1; Matthew Chancey1; Hyosim Kim1; Yongqiang Wang1; Tuhin Mukherjee3; Jeff Bickel4; Tarasankar DebRoy3; Peter Hosemann4; Thomas Lienert5; Stuart Maloy6; 1Los Alamos National Laboratory; 2University of Tennessee, Knoxville; 3Pennsylvania State University; 4University of California, Berkeley; 5Optomec, Inc; 6Pacific Northwest National Laboratory
    Laser-directed energy deposition (L-DED) offers significant savings in tooling and material. However, Grade 91 steel contains mostly fine martensite when produced using L-DED. This renders a promising reactor material – resistant to radiation and high-temperature creep in the wrought form – extremely brittle. In place of a traditional tempering heat treatment, which would limit part size and increase expense, the principles of multi-pass welding were used to temper L-DED build layers in situ. To understand the impact of L-DED and in situ tempering (i.e., changes in feature size and phase) on the radiation tolerance of Grade 91 steel, wrought material and L-DED samples built with and without the in situ “pre-heat” were irradiated with self-ions (300 °C, 60 dpa, 5e-3 dpa/s) and protons (300 °C, 0.3 dpa, 1e-5 dpa/s). The accumulation of radiation-induced defects in these samples and the resulting changes to their nanomechanical behavior are considered.

9:40 AM  
Fracture Toughness of Highly Irradiated RPV Steels: Mikhail Sokolov1; Xian Chen1; Takuya Yamamoto2; Robert Odette2; Randy Nanstad3; 1ORNL; 2UCSB; 3R&S Consulting
    This paper describes fracture toughness of 3 radiation-sensitive RPV steels. Steels were irradiated up to ~ 1.38x1020 n/cm2 at 290oC. ATR-2 experiment also included Disk Compact Tension (DCT) specimens to explore possible changes in the ASTM E1921 Master Curve shape and reference temperature shifts (ΔTo) of highly sensitive irradiated steels, which experience very large hardening, up to more than 400 MPa. The ATR-2 data were also used to establish the corresponding irradiated ΔTo vs. Δσy relation. This unique data set suggests that KJc(T) for all three steels consistent with the Master Curve shape over the temperature range of cleavage fracture. The ΔTo/Δσy ratio varied between ~0.6 to 0.84 averaging ~0.71, which is very close to previously observed values. However, this experiment has also highlighted that the shape of the Master Curve may less suitable for highly embrittled steels with low upper ductile fracture toughness.

10:00 AM Break

10:20 AM  
Microstructure and Mechanical Properties of Neutron Irradiated Tantalum-alloyed Ferritic Martensitic Steels: Weicheng Zhong1; Lizhen Tan1; Thak Sang Byun1; Ying Yang1; 1Oak Ridge National Laboratory
    Castable nanostructured alloys (CNAs) are being developed in ORNL as American reduced activation ferritic martensitic steels for the first wall/blanket applications in fusion reactors. One of the early generation (tantalum-alloyed) CNAs has been irradiated in High Flux Isotope Reactor at temperatures up to ~650℃ to damage doses of ~0.5 - 9 dpa. Vickers microhardness and tensile tests were performed at room temperature, and they reveal consistent hardening or softening depending on the irradiation temperature. Irradiation induced hardening was observed at the irradiation temperature of 400℃, and softening at >490℃. Microstructures in irradiated samples were characterized to understand the post-irradiation mechanical behaviors. SEM-EBSD is used to trace the grain structure evolution, and TEM is used to investigate the irradiation induced defects and precipitates stability. The microstructural changes will be correlated to understand the tensile behaviors after irradiation.

10:40 AM  
Impact of Electrolytic Hydrogen Charging on Fatigue Crack Propagation in Reactor Steels: Melissa Weihrauch1; Maulik Patel1; Eann Patterson1; 1University of Liverpool
    In pressurised water reactors, coolant pipes are exposed to hydrogen and irradiation, both of which cause embrittlement. Thus, investigations into their synergistic effects are required to support advances in design concepts. 316LN stainless steel samples were electrolytically charged and the release rate of hydrogen from samples was monitored. Subsequently, compact tension specimens, intended for fatigue testing, were hydrogen-charged for 48 hours and neutron damage was emulated via irradiation with 30 MeV Ni ions. Specimens were fatigue tested in either virgin, hydrogen precharged, irradiated and hydrogen precharged + irradiated states. The fatigue crack growth and plastic zone size were monitored using thermoelastic stress analysis (TSA). TSA is a full field measurement technique which can uniquely measure the extent of plasticity and the level of elastic stresses in specimens. A reduction of fatigue life in irradiated specimens was observed. Furthermore, hydrogen precharging prior to irradiation adversely effected fatigue life.

11:00 AM  
Estimating the Strengthening Parameters for Irradiated Alloys using Atomic Scale: Osetsky Yury1; German Samolyuk1; 1Oak Ridge National Laboratory
     Models based on the dispersed barrier hardening (DBH) approach are often used for interpreting experimental results and estimating the mechanical response of irradiated materials. However, the accuracy of DBH models is rather low due to weak physical approaches and phenomenological fitting in estimating the obstacle strength parameter. We suggest a new approach that utilizes the accumulated over decades results of atomic scale modeling of dislocation-obstacle interactions, much of which is directly related to radiation induced defects. In this work, we developed a simple approach that combines early parametric continuum modeling with current atomic scale dislocation dynamics. This approach allows estimating the size and temperature dependent values of obstacle strength parameters, using the available atomistic modeling data. Examples are presented for cases in two materials of practical significance e.g. ferritic and tungsten-rhenium alloys. The implication of these results to the understanding of experimental observations is discussed.