Materials in Nuclear Energy Systems (MiNES) 2021: Fundamental Irradiation Damage- Session IV
Program Organizers: Todd Allen, University of Michigan; Clarissa Yablinsky, Los Alamos National Laboratory; Anne Campbell, Oak Ridge National Laboratory
Wednesday 8:00 AM
November 10, 2021
Room: Urban
Location: Omni William Penn Hotel
Session Chair: Enrique Martinez, Clemson University
8:00 AM Cancelled
Low Temperature Hardening-embrittlement Phenomenon IN 9-14% Chromium Based Ferritic-martensitic and Oxide Dispersion Strengthened Steels: Arunodaya Bhattacharya1; Samara Levine2; Yutai Katoh1; Steven Zinkle3; 1Oak Ridge National Laboratory; 2University of Tennessee, Knoxville; 3Oak Ridge National Laboratory; University of Tennessee, Knoxville
High chromium ferritic-martensitic (FM) steels and ODS steels are promising candidates for the fusion first-wall, and claddings/ducts of sodium-cooled fast reactors [1]. However, a narrow operating temperature window, ~330-550°C, is typically envisaged for FM steel thick-wall structures. The upper temperature limit, which may increase to ~650°C with ODS, is due to poor creep strength. The low temperature hardening-embrittlement (LTHE) phenomenon causing fracture toughness loss imposes the lower temperature limit [1]. Currently, LTHE is not well-understood. Here, a phenomenological overview of LTHE is presented to better understand the operating limits for FM steels, using which some alloy design guidance is presented. In addition, the susceptibility of ODS steels to hardening-embrittlement is presented based on compelling new neutron irradiated data from ORNL. Research sponsored the U.S. Department of Energy, Office of Fusion Energy Sciences under contract DE-AC05-00OR22725 with UT-Battelle LLC. [1] S. J. Zinkle et al., Nucl. Fusion. 57(9) (2017) 092005
8:40 AM Cancelled
Decoupling Thermal and Irradiation Effects on Clustering and Chemical Redistribution in 14YWT ODS: Amrita Sen1; Mukesh Bachhav2; Janelle Wharry1; 1Purdue University; 2Idaho National Laboratory
The objective of this study is to understand the roles of temperature and irradiation on nanocluster stability and grain boundary chemistry in 14YWT oxide dispersion strengthened (ODS) steel. ODS steel is being considered for advanced nuclear reactor structural and cladding material. Studies have attempted to understand nanocluster irradiation behavior, but there remains little consensus on the influence of competing factors of thermal diffusion and irradiation disordering. Moreover, synergistic chemical evolution of grain boundaries and nanoclusters has not yet been considered. Here, we conduct separate annealing and irradiation experiments on 14YWT at 400°C and 500°C to 100 dpa using 4.5-5.0 MeV Fe++. We characterize the nanocluster and grain boundary microchemistry using atom probe tomography (APT). Results suggest nanocluster morphology is driven by irradiation, but a combination of irradiation and thermal diffusion produces grain boundary segregation of clustered species. These mechanisms and their influence on properties will be discussed.
9:00 AM
Dose and Temperature Effect on Dispersoids in Neutron Irradiated Oxide Dispersion Strengthened (ODS) Alloys: Samara Levine1; Arunodaya Bhattacharya2; Jonathan Poplawsky2; Andrew Lupini2; David Hoelzer2; Yutai Katoh2; Steven Zinkle1; 1University of Tennessee; 2Oak Ridge National Laboratory
Although ODS alloys are among the leading candidates for fusion first-wall/blanket (FW/B) structures, the effect of fusion-relevant irradiation on these advanced materials is still not well-understood. Here, three ODS steel variants (12%Cr 12YWT, 14%Cr MA957, and 20%Cr-5.5%Al PM2000) were neutron irradiated in the high-flux isotope reactor (HFIR) at 300-500°C for ~3.5-80 dpa. Previously we revealed at low temperatures and high doses, transformation of the nano-dispersoids included cavitation, amorphization, and formation of internal “cherry-pit” structures. Here we expand upon the effects of neutron dose and temperature on nano-dispersoid stability using analytical scanning transmission electron microscopy (STEM) and atom probe tomography (APT). Possible mechanisms are presented to explain structure-chemistry evolution of the nano-dispersoids. Research sponsored by the U.S. Department of Energy, Office of Fusion Energy Sciences, under contract DE-AC05-00OR22725 with UT-Battelle, LLC. APT/STEM was conducted at CNMS, which is a DOE Office of Science User Facility. Work supported under UTK’s GATE Fellowship.
9:20 AM
The Subtle Effects of Nitrogen on Radiation Effects in Tempered Martensitic Steels: Stuart Maloy1; B Eftink1; H Kim1; C Rietema2; E Aydogan3; H Vo4; 1Los Alamos National Laboratory; 2Colorado School of Mines; 3Middle East Technical University; 4University of California, Berkeley
The Nuclear Technology R&D program is investigating options to transmute minor actinides. To achieve this goal, new fuels and cladding materials must be developed and tested to high burnup levels (e.g. >20%) requiring cladding to withstand very high doses (greater than 200 dpa) while in contact with the coolant and the fuel. Tempered Martensitic Alloys are the leading candidates for these extreme service conditions. Recent irradiations on tempered martensitic alloys show that slight variations in the composition of one tempered martensitic alloy, HT9, can improve resistance to low temperature embrittlement. This material maintained 5% uniform elongation after irradiation to 6 dpa at 290C while all other alloys exhibited less than 2% uniform elongation. To investigate the reasons behind these significant improvements, controlled alloys were produced while systematically varying the nitrogen concentration between 10 and 500 wppm nitrogen. Ion irradiations performed on these alloys showed that model alloys with higher nitrogen show a higher loop density while heats of HT9 with controlled nitrogen content show lower overall void swelling and a higher density of fine G-phase precipitates.These results will be summarized along with their correlations with radiation effects in tempered martensitic steels.
9:40 AM
Defect Cluster Configurations and Mobilities in α-zirconium: Implications for Breakaway Irradiation Growth: Jose March-Rico1; Brian Wirth1; 1University of Tennessee, Knoxville
A key component necessary for the predictive modeling of irradiation growth strains is a detailed understanding of defect cluster configurations and their expected modes of transport. In this work, we use a modern interatomic potential published in 2020 (the BMD19 potential) to analyze the preferred structures and mobilities of SIA and vacancy clusters. We find that small SIA clusters form configurations that are contained entirely within a single basal plane and, consequentially, migrate exclusively in 2-D within the basal plane. Large clusters form perfect dislocation loops and migrate rapidly in 1-D. Conversely, small vacancy clusters migrate either quasi-isotropically or with a preference for migration along the c-axis; this is in stark contrast to the considerable anisotropy of the monovacancy. Therefore, the inherent difference in the anisotropy of diffusion of defect clusters, rather than point defects, will be a critical component for the accurate modeling of microstructural evolution in irradiated zirconium.
10:00 AM Break