Materials and Fuels for the Current and Advanced Nuclear Reactors VI: Structural Materials II
Sponsored by: TMS Structural Materials Division, TMS: Corrosion and Environmental Effects Committee, TMS: Nuclear Materials Committee, TMS: Mechanical Behavior of Materials Committee
Program Organizers: Ramprashad Prabhakaran, Pacific Northwest National Laboratory; Dennis Keiser, Idaho National Laboratory; Raul Rebak, GE Global Research; Clarissa Yablinsky, Los Alamos National Laboratory

Wednesday 8:30 AM
March 1, 2017
Room: Cardiff
Location: Marriott Marquis Hotel

Session Chair: Clarissa Yablinsky, Los Alamos National Laboratory; Indrajit Charit, University of Idaho


8:30 AM  
Irradiation-Induced Microstructure of Proton Irradiated Commercial Austenitic Alloys: Miao Song1; Gary Was1; 1University of Michigan
    Proton-irradiated microstructures are reported for several austenitic alloys to a damage level of 5 dpa at 360 C, including 316L stainless steel (SS), 310SS, alloy 800, alloy 690 and alloy 625. The Fe content ranges from 70 wt. % in 316L down to 3.5 wt.% in nickel-base alloy 625. In contrast, the nickel content ranges from 10 wt. % in 316L to 61% in alloy 625. The as-received alloys show clean grain structures with few dislocations and intergranular carbides. After irradiation, dislocation loops were formed in all of the alloys. Nano-voids were observed in 316L SS. In the nickel-rich alloys, voids were not observed but they contained a high density of irradiation-induced precipitates. Although these alloys all have an austenitic structure with major elements Fe-Cr-Ni, the irradiated response of the microstructures showed significant variance with chemical composition. The microstructure evolution will be interpreted in the context of alloy composition.

8:50 AM  
Neutron Irradiation-induced Creep of IG-110 Nuclear Graphite: Anne Campbell1; Eiji Kunimoto2; Yutai Katoh1; 1Oak Ridge National Laboratory; 2Toyo Tanso Co. Ltd.
    Graphite will be heavily used in the Generation-IV High Temperature Gas Reactors (HTGR) and in the Molten Salt Reactor (MSR) concepts. Within each graphite core component (block) there are temperature and neutron flux gradients that will cause the build-up of internal stresses, which can rapidly surpass the tensile strength and cause cracking and failure of the blocks. Meanwhile nuclear graphite also undergoes irradiation-induced creep, which allows for relaxation of these stresses and prevents cracking. But this behavior has not been extensively studied for newer graphite grades, e.g. IG-110, so the creep behavior of IG-110 is not well known. This presentation will discuss the irradiation creep experiments that are being performed at Oak Ridge National Laboratory in the High Flux Isotope Reactor. This program studied the compressive creep behavior at three temperatures and a single applied dead-load compressive stress, and the accompanied changes to the elastic properties and compressive strength.

9:10 AM  
Investigation of Property-Property Correlations for Irradiated Steels: Peter Wells1; Takuya Yamamoto1; Nathan Almirall1; Randy Nanstad2; Timothy Milot1; G. Odette1; 1UC Santa Barbara; 2Oak Ridge National Laboratory
    Irradiation of structural alloys in a nuclear reactor leads to changes in their mechanical properties, including hardening and embrittlement. Ensuring the safe operation of reactors requires an understanding and quantification of these property changes. Ideally, standard tensile or fracture tests would be used to fully quantify a steels response to irradiation. Unfortunately, irradiated test specimens are often very small, requiring alternative testing methods, such as microhardness, to be used. This talk details correlations that have been developed, and refined over time, between various mechanical property changes under irradiation, including tensile yield and microhardness changes, as well as ductile-to-brittle transition temperature shifts. In addition, shear punch testing, which can be used to derive stress-strain curves similar to tensile tests, will also be discussed and correlated with the other test methods.

9:30 AM  
Mitigation of IASCC Susceptibility in a BWR-irradiated 304L Stainless Steel Utilizing Post-irradiation Annealing: Justin Hesterberg1; Zhijie Jiao1; Gary Was1; 1University of Michigan
    Post-irradiation annealing (PIA) of alloy 304L stainless steel irradiated to 5.9 dpa was conducted to investigate the cause of irradiation-assisted stress corrosion cracking (IASCC). The annealing treatments were performed at temperatures in the range 450–600C and for times ranging from 1–20 hours. Based on the residual hardening values, four annealing conditions: 500C: 1 hr and 550C: 1, 5, and 20 hrs, were selected to cover the expected range of IASCC mitigation resulting from PIA treatment. IASCC susceptibility was measured for the as-irradiated and annealed conditions via CERT tests under simulated BWR-NWC conditions (288C, 2000 ppb DO) at a strain rate of 3.0 x 10-7 1/s through incremental straining experiments to provide a more precise determination of the onset of IASCC crack initiation and correlation with dislocation channeling. The resulting IASCC behavior is correlated with annealing condition and dislocation channel emergence.

9:50 AM  
Role of Localized Deformation and Grain Boundary Plane Orientation on Crack Initiation in Irradiated Stainless Steels: Drew Johnson1; Bryan Kuhr2; Diana Farkas2; Gary Was1; 1University of Michigan; 2Virginia Tech
    The localization of strain into discrete dislocation channels is a potentially key factor in the crack initiation process in irradiated austenitic stainless steels. High local build-up of stress generated by dislocations pinned at dislocation channel-grain boundary (DC-GB) intersections is believed to assist in nucleating intergranular stress corrosion cracks. High Angular Resolution Electron Backscatter Diffraction (HREBSD) coupled with 3-dimensional measurements of the grain boundary orientation have been used to determine the component of stress normal to the grain boundary plane at specific DC-GB interaction sites. Stress distributions have been correlated with various material and orientation parameters to help isolate the factors important for the elevated stress values. Slow strain rate tests in a high temperature water environment provided Irradiation Assisted Stress Corrosion Crack (IASCC) initiation data, which was then compared with 3D stress measurements and molecular dynamics simulations to gain deeper insight into the mechanism responsible for IASCC.

10:10 AM Break

10:30 AM  
The Effect of Low-fluence Neutron Irradiation on Cast Austenitic Stainless Steels: Siwei Chen1; Yuichi Miyahara1; Akiyoshi Nomoto1; Kenji Nishida1; 1Central Research institute of Electric Power Industry
    Neutron irradiation was performed to three sorts of cast austenitic stainless steels with and without thermal aging at 290 C to a fluence of 4.31018 n/cm2. Microstructural evolution and hardness change were investigated by atom probe tomography and nano-indentation, respectively. No obvious microstructural and hardness changes of austenite phase were detected. On the other hand, the changes in microstructure and hardness of ferrite phase were observed in irradiated materials with and without thermal aging. For materials without thermal aging, ferrite hardness increased after the low-fluence neutron irradiation. In case of thermally-aged materials, the Mo added material showed an increase of hardness after irradiation; however, the hardness change of material without Mo was limited. In this study, we discussed the microstructural evolution and corresponding hardness change after low-fluence neutron irradiation.

10:50 AM  
Effects of Thermal Aging and Neutron Irradiation on Cast Austenitic Stainless Steels: Wei-Ying Chen1; Yiren Chen1; Chi Xu2; Zhangbo Li2; Yong Yang2; Nicholaos Demas1; 1Argonne National Laboratory; 2University of Florida
    Cast austenitic stainless steels (CASS) have a dual-phase microstructure of delta ferrite and austenite, and are used widely in the cooling system of light water reactors. Due to both thermal aging and irradiation embrittlement, the long-term performance of CASS is of concern. In this study, the microstructure and mechanical property of thermally-aged and neutron-irradiated specimens were characterized with transmission electron microscope and nanoindentation. After neutron irradiation of 0.08 dpa at ~320C, a few dislocation loops and Si/Ni-enriched clusters were observed in austenite. The ferrite, however, exhibited a structure of α-α′ separation, G-phase precipitates (3-8 nm, 1021-1022 m-3) and very few dislocation loops. After either thermal aging or neutron irradiation, the hardness of ferrite increased by about 40%, while the hardness of austenite remained unchanged. Both hardness and TEM indicated that the decrease in the fracture toughness of CASS under current irradiation condition was mainly due to the embrittlement of ferrite.

11:10 AM  
Utilizing In-situ Microtensile Testing to Evaluate Mechanical Property Changes Due to Ion-beam Irradiation: Hi Vo1; Stuart Maloy2; Peter Hosemann1; 1University of California, Berkeley; 2Los Alamos National Laboratory
    Understanding changes in mechanical properties and failure mechanisms as a function of irradiation damage is crucial in extending reactor lifetime and designing new generation reactors. Testing neutron irradiated materials can be challenging because of the long irradiation time needed to achieve high dose. Therefore, ion beam irradiation has been used as a surrogate for neutron irradiation. However, obtaining mechanical data on shallow ion beam irradiation depths is difficult. As a result, small scale mechanical testing has recently been employed to study ion beam-irradiated materials and quantify the mechanical property changes. In this study, in situ microtensile testing was used to investigate the effect of proton irradiation on 304 SS and HT-9. Increase in yield strength, decrease in total elongation, and brittle fracture were observed in 304 SS after ion irradiation. In addition, the effect of nitrogen interstitials on the ductility of irradiated HT-9 was also investigated.

11:30 AM  
In-situ High Energy X-ray Characterization of Neutron Irradiated HT-UPS Stainless Steel under Tensile Deformation: Chi Xu1; Xuan Zhang2; Meimei Li2; Jun-Sang Park2; Peter Kenesei2; Jonathan Almer2; Yong Yang3; 1Argonne National Laboratory / University of Florida; 2Argonne National Laboratory; 3University of Florida
    To understand the neutron irradiation effect on the deformation behavior of a High Temperature Ultra-fine Precipitate Strengthened (HT-UPS) Stainless Steel, in-situ tensile tests with high-energy X-ray measurements were conducted on neutron irradiated (400C/3dpa) and un-irradiated samples at room temperature using the in-situ Radiated Materials (iRadMat) thermo-mechanical characterization apparatus at the Advanced Photon Source. Simultaneous Wide Angle X-ray Scattering (WAXS) and Small Angle X-ray Scattering (SAXS) measurements were performed during tensile deformation using 123 keV X-ray beams. The changes of X-ray peak positions and peak broadening were analyzed and correlated with the deformation process. TEM was conducted post-mortem to characterize the microstructure of as-irradiated and deformed regions to complement X-ray measurements.