Radiation Effects in Metals and Ceramics: Irradiation of Fe-based Systems
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee
Program Organizers: Djamel Kaoumi, North Carolina State University; Thak Sang Byun, Oak Ridge National Laboratory; Dane Morgan, University of Wisconsin-Madison; Maria Okuniewski, Purdue University; Mahmood Mamivand; Geoffrey Beausoleil, Idaho National Laboratory; Philip Edmondson, The University Of Manchester; Khalid Hattar, University of Tennessee Knoxville; Aurelie Gentils, Université Paris-Saclay; Joel Ribis, Cea

Tuesday 8:30 AM
February 25, 2020
Room: Theater A-7
Location: San Diego Convention Ctr

Session Chair: Joel Ribis, Commissariat a l'Energie Atomic CEA; Phil Edmonson, Oak Ridge National Laboratory


8:30 AM  Invited
Separate Effects to Integral Effects - All Things Radiation Effects in FeCrAl: Kevin Field1; Maxim Gussev1; Xiang Chen1; Caleb Massey1; Dalong Zhang2; Samuel Briggs3; Janelle Wharry4; Kurt Terrani1; 1Oak Ridge National Laboratory; 2Pacific Northwest National Laboratory; 3Oregon State University; 4Purdue University
    FeCrAl alloys are a class of high-chromium ferritic alloys that are being studied for use in nuclear power applications. FeCrAl alloys possess many favorable properties including environmental degradation resistance (e.g. corrosion/oxidation) which lend them towards cladding and core internal applications. These applications require the FeCrAl alloys to also possess radiation tolerance. Over the past half-decade, a systematic study on the radiation tolerance of FeCrAl alloys has been completed. These experiments include simple, separate effects tests determining trends in irradiation dose, irradiation temperature, starting composition, etc. on both the microstructure evolution and mechanical responses. More recently, a series of tests have also been initiated looking at correlated phenomena such as irradiation creep, fuel-cladding interaction, and irradiation-assisted corrosion testing. This talk will summarize the critical findings from both the separate and integral radiation effects testing and provide an assessment on the current state of knowledge on the radiation tolerance of FeCrAl alloys.

9:00 AM  
Effects of Radiation Parameters on Defect Evolution in FeCrAl Alloys under Single-, Dual- and Triple-Ion Beam Irradiation: Pengyuan Xiu1; Li Jiang1; Chao Ye1; Lumin Wang1; 1Department of Nuclear Engineering and Radiological Sciences, University of Michigan
    Two FeCrAl alloys with different minor additives were irradiated with 6 MeV Au ions at a constant dose rate of 4x10-3 dpa/s to doses of 10, 25, 50 and 100 dpa at 400℃. Both alloys were also irradiated to 100 dpa at 475℃ and 575℃ to study the temperature effects, and with two more dose rates of 4x10-4 and 2x10-2 dpa/s to study the dose rate effect. High density of dislocation loops with Burgers vectors of 1/2<111> and <100> were observed after all irradiations. The formation of <100> loops seems to favors high temperature, low dose rate, and no effect of dose is observed. Although no voids were observed after the single Au ion beam irradiation, a low density of bubbles/voids exists after dual beam using H+Fe or He+Fe ions, or triple beam irradiation using H+He+Fe ions with appmHe/dpa and appmH/dpa to be 9 and 70 to 50 dpa.

9:20 AM  
Irradiation Enhanced Alpha Prime Precipitation in 2nd Gen. FeCrAl Alloys After Neutron Irradiation to 7 dpa: Caleb Massey1; Kevin Field1; Philip Edmondson1; Steven Zinkle2; 1Oak Ridge National Laboratory; 2University of Tennessee
    In the search for advanced materials that can help increase coping time or reduce the severity of nuclear accident scenarios, FeCrAl alloys have shown promise as a potential drop-in replacement for existing Zr-based cladding. However, the radiation tolerance of 2nd Generation Fe-(10-16Cr)-(5-7Al)-2Mo (wt.%) has yet to be investigated to higher doses of neutron irradiation. Recently irradiated FeCrAl alloys in the composition range above have been irradiated in the High Flux Isotope Reactor to 7 displacements per atom (dpa) at 330C. The post-irradiation precipitation state of these alloys have been evaluated using the atom probe tomography technique. Since the precipitation of the Cr-rich alpha prime phase is a key factor in the irradiation hardening of the FeCrAl alloy system in this dose/temperature regime, the dispersion and composition characteristics of this Cr-rich phase is evaluated as a function of alloy composition and irradiation dose.

9:40 AM  
Data-driven Discrete Dislocation Dynamics Modeling of Yielding Behavior of Irradiated FeCrAl Steel: Yash Pachaury1; Sanjoy Mazumder1; George Warren1; Giacomo Po2; Janelle Wharry1; Anter El-Azab1; 1Purdue University; 2University of Miami
    This work entails the development of a novel data-driven 3D discrete dislocation dynamics simulation methodology capable of predicting mechanical behavior of FeCrAl alloy, C35M (nominally Fe-13%Cr-5%Al), due to periodic composition fluctuations induced by irradiation. As known in the literature, irradiation of multi-component alloys results in spinodal like instabilities manifested in the form localized enrichment and depletion of certain species throughout the material due to the inverse Kirkendall effect, which alters the yielding behavior of alloys. Composition fluctuations in irradiated FeCrAl alloy were modeled using statistical methods. Fixed by molecular dynamics, a composition dependent mobility of dislocations within the alloy was utilized into a discrete dislocation dynamics model to predict the yielding behavior of the irradiated alloy. The stress-strain response of the alloys prior to and post irradiation were analyzed and compared. The results elucidate the importance of composition effects in the plastic behavior of irradiated alloys.

10:00 AM Break

10:20 AM  
Radiation Response of Grade 92 Ferritic-martensitic Steel Irradiated up to 14.63 dpa at ~700°C: Weicheng Zhong1; Lizhen Tan1; 1Oak Ridge National Laboratory
    Ferritic-martensitic steel Grade 92 is a candidate structural material for Gen-IV nuclear reactors. In contrast to the irradiation studies generally conducted at 300–600°C, this study reports the evolution of microstructures and mechanical properties of Grade 92 irradiated up to 14.63 displacements per atom (dpa) at ~700°C in the High Flux Isotope Reactor of Oak Ridge National Laboratory. Room temperature tensile test and Vickers hardness measurement were performed to evaluate mechanical property evolution after 0.46, 7.44, and 14.63 dpa irradiations. Microstructural characterization using transmission electron microscopy, coupled with energy dispersive x-ray spectroscopy and electron backscatter diffraction, manifested the primary evolution of grain structure and precipitates. The virgin and 700°C-aged Grade 92 samples were evaluated in parallel as references for the irradiated samples to better understand the irradiation effect on the microstructure and mechanical property evolution at ~700°C.

10:40 AM  
Comparison of the Irradiated Microstructure Formed in 800H After Neutron Irradiation and Dual Beam Ion Irradiation: Christopher Ulmer1; Arthur Motta1; 1The Pennsylvania State University
    Ion irradiation is often used as a substitute for in-reactor testing when studying radiation effects in materials. However, the radiation damage process is affected by several independent variables including irradiation dose rate, irradiation temperature, and the production of helium by nuclear reactions. In this study, samples of 800H alloy were irradiated in the BOR-60 reactor to 17 dpa at nominally 376 °C, while complementary samples were irradiated to similar doses with dual beam 5 MeV Fe and energy-degraded 1.95 MeV He ions at the Michigan Ion Beam Laboratory. The helium co-injection ranged from 0 to 276 appm. A temperature shift was used to compensate for the higher dose rate, and ion irradiation temperatures ranged from 430 to 500 °C. The irradiated microstructure, including faulted dislocation loops, cavities, precipitates, and grain boundary segregation, was characterized with transmission electron microscopy. The microstructures formed under dual ion beam and neutron irradiations are compared.

11:00 AM  
Microstructural Investigation of Flux Effect on Neutron-irradiated RPV Steels: Auriane Etienne1; Andreas Ulbricht2; Bertrand Radiguet1; 1University Of Rouen; 2Helmholtz-Zentrum Dresden-Rossendorf
     The extension of the lifetime of pressurized water reactors up to 60 years or more requires to anticipate the evolution of the microstructure at the origin of irradiation hardening and embrittlement. A way to reach high neutron fluences corresponding to these durations is the use of material test reactors, with high neutron flux. In this case, it is important to understand the effect of neutron flux on microstructural evolution. In this work, two reactor pressure vessel base metals (low Cu, low Mn, medium Ni) and two welds (low Cu, medium Mn, high Ni) were neutron irradiated at the same fluence but under low and high fluxes. Irradiated microstructures were investigated by atom probe tomography. The flux effect on solute clustering and segregation will be discussed

11:20 AM  
Influence of Alloying Elements on Microstructure Evolution in 21Cr32Ni Model Alloy Microstructure after In-situ Ion Irradiation: Muhammet Ayanoglu1; Christopher Ulmer1; Arthur Motta1; 1The Pennsylvania State University
     The development of new generation advanced reactors relies on understanding how radiation affects the microstructure of candidate alloys in service. The austenitic alloy 800H (21w/t% Cr and 32w/t% Ni) has been selected as one of the primary candidate alloys for use in advanced nuclear systems due to its high-temperature corrosion and creep resistance. However, its behavior under irradiation needs further investigation to be able to certify its operation in Gen IV reactors.In this study, the effect of dose and irradiation temperature on microstructure evolution of compositionally simple austenitic 21Cr32Ni model alloy has been investigated in-situ at Intermediate Voltage Electron Microscope (IVEM) at Argonne National Laboratory up to ~2 dpa at irradiation temperatures ranging from -223oC to 440oC. The radiation average diameter, size distribution and density of irradiation-induced dislocation loops, and clusters were experimentally determined as a function of irradiation dose and temperature using transmission electron microscopy, and compared with those obtained in a commercial alloy 800H irradiated under similar conditions.

11:40 AM  
Response of Solidification Cellular Structures in Additively Manufactured 316 Stainless Steel to Heavy Ion Irradiation: an In situ Study: Zhongxia Shang1; Cuncai Fan1; Sichuang Xue1; Jie Ding1; Jin Li1; Thomas Voisin2; Yinmin Wang2; Haiyan Wang1; Xinghang Zhang1; 1Purdue University; 2Lawrence Livermore National Laboratory
    In-core or cladding structural materials exposed to heavy ion irradiation often suffer serious irradiation-induced damages. Introducing defect sinks can effectively mitigate irradiation-induced degradation in materials. Here, we investigated the radiation response of additively manufactured 316 austenitic stainless steel with high-density solidification cellular structures via in situ Kr++ irradiation at 400˚C to 5dpa. The study shows that the cellular walls with trapped dislocations can serve as effective defect sinks, thus reduce dislocation loop density compared with the conventional coarse-grained counterparts. This study provides a positive step for the potential applications of radiation-resistant, additively manufactured steels in advanced nuclear reactors.