Accelerated Discovery and Qualification of Nuclear Materials for Energy Applications: Irradiation Effect in Nuclear Fuels and Materials
Sponsored by: TMS Structural Materials Division, TMS Materials Processing and Manufacturing Division, TMS: Integrated Computational Materials Engineering Committee, TMS: Nuclear Materials Committee, TMS: Additive Manufacturing Committee
Program Organizers: Yongfeng Zhang, University of Wisconsin; Adrien Couet, University of Wisconsin-Madison; Michael Tonks, University Of Florida; Jeffery Aguiar, Lockheed Martin; Andrea Jokisaari, Idaho National Laboratory; Karim Ahmed, Texas A&M University

Thursday 2:00 PM
March 18, 2021
Room: RM 48
Location: TMS2021 Virtual

Session Chair: Mohammed Abdoelatef, Texas A&M U


2:00 PM  
Point Defect Capture Characteristics and Stress States of Dislocation Loops in α-zirconium: Jose March-Rico1; Brian Wirth1; 1University of Tennessee Knoxville
     The accuracy of mesoscale computational techniques such as cluster dynamics (CD) is dependent upon mechanistically-informed input from lower-length scale techniques such as molecular dynamics (MD). In this work, we use a modern interatomic potential for hcp α-Zr to assess point defect capture characteristics to interstitial a-loops, vacancy a-loops, and vacancy c-loops through MD simulations. Spontaneous capture distances (defined as the distance at which the binding energy dropped to half of its maximum value) are compared to the distances needed to reduce the binding below the thermal energy kT. Spontaneous capture distances are correlated to the gradient of the potential energy field while long-range thermal drift capture is correlated to the stress field of the loop. Short-range spontaneous capture (~1-10 Å) favors interstitials while long-range thermal drift (>10 Å) favors vacancy capture under certain circumstances.Input parameters for point defect capture are provided for incorporation into CD frameworks.

2:20 PM  
Comparison of Void Swelling in Conventional and Novel HT9 Alloys after High Damage Level Ion Irradiation: Hyosim Kim1; Jonathan Gigax1; Osman Atwani1; Stuart Maloy1; Yongqiang Wang1; Matthew Chancey1; Jon Baldwin1; 1Los Alamos National Laboratory
    HT9 is a conventional cladding and structural material that is used in nuclear reactors, with an extensive history in service due to its superior radiation tolerance up to ~200 dpa. While conventional HT9 alloys, such as the INL and ACO-3 heats, have been studied after both neutron and ion irradiation, novel alloys such as nitrogen modified heats, have not been investigated in such detail. We utilized Fe self-ion irradiation at 450 °C to irradiate conventional (INL, AC03) and novel (Low N, High N) HT9 alloys up to 600 peak dpa. The INL and ACO-3 heats showed a similar swelling response. The low N heat exhibited the largest amount of swelling, about twice more than the conventional heats. However, the high N heat showed the least amount of swelling, with one third the swelling of the conventional heats and an order of magnitude less than the low N heat.

2:40 PM  
Dislocation Loop Formation in Self-ion Irradiated Ultra-high Purity FeCr Alloys: Yao Li1; Yajie Zhao1; Arunodaya Bhattacharya2; Jean Henry3; Steven Zinkle1; 1The University of Tennessee, Knoxville; 2Oak Ridge National Laboratory; 3The French Alternative Energies and Atomic Energy Commission
    The evolution of ½ <111> vs <100> loops in Fe-Cr alloys as a function of Cr concentration, dose, dose rate, and irradiation temperature is important for understanding the basis for radiation resistance. 8 MeV Fe ion irradiations (~0.35 and 3.5 dpa midrange doses) at dose rates of E-5 and E-4 dpa/s were performed on ultra-high purity Fe and Fe-Cr alloys (3-18 wt.%Cr) at 350-450°C. Using conventional diffraction contrast TEM imaging and detailed Burgers vector analysis, we observed loop size reduction as well as density increase with the addition of Cr, an increasing ½ <111> fraction with higher Cr concentration, the fraction of ½ <111> loops dropping from ~20% to ~2% with increasing temperature (in agreement with prior limited studies), nanoscale voids in Fe, and petal shaped loops at 450°C. Analytical scanning TEM (STEM)-high throughput energy dispersive X-ray spectroscopy (EDX) quantified Cr segregation onto dislocation loops for Cr<10 wt.%.

3:00 PM  
Effect of Microstructure and Rolling Treatment on Static Recrystallization Behavior in Monolithic U-10Mo Fuel Foils: William Frazier1; Kyoo Sil Choi1; Lei Li1; Zhiie Xu1; Vineet Joshi1; Ayoub Soulami1; 1Pacific Northwest National Laboratory
    A developed and validated coupling of Kinetic Monte Carlo (KMC) Potts Model and finite element method (FEM) simulations was implemented in order to investigate the impact of microstructural features in as-cast and homogenized monolithic U-10Mo foils on the emergent microstructure after rolling and reheating treatments. Grain size distribution, uranium carbide (UC) size distribution, UC volume fraction, the spatial distribution of UC, and rolling reduction magnitude were considered parameters that could potentially affect recrystallization behavior of the rolled U-10Mo foils. This simulation study established that grain structure and the magnitude of rolling reduction have the strongest influence on recrystallization kinetics and the fabricated grain size distribution. The UC distribution had only a weak impact on the recrystallization kinetics and final microstructures. While particle stimulated nucleation (PSN) occurred in simulation with increasing frequency as grain size increased, its incidence did not appear to considerably affect the recrystallization kinetics or grain size distribution.

3:20 PM  
Properties of a Helium Ion Beam Degrader for Implanting SSJ2 Tensile Specimens at the LBL 88-Inch Cyclotron: Sarah Stevenson1; Adi Ben-Artzy1; Lee Bernstien2; Peter Hosemann1; 1University of California, Berkeley; 2LBL
    A Helium ion implantation system was developed to evaluate radiation damage of bulk scale HT-9 nuclear application steel. A novel high-energy binary degrader system was designed to provide uniform Helium ion implantation of 5 separate 100 μm thin SSJ2 tensile test specimens at once. The binary degrader is designed to provide 165 discrete energy levels, stopping the Helium ions throughout the HT-9 sample depth. It was shown using SRIM simulations that the degrader will provide the discrete energy Helium ion implantation levels as expected. The LBL 88-Inch Cyclotron will provide a 26 MeV and 20 μA total current incident beam, and will achieve 0.5 at% Helium in each sample in about 3 days of constant beam run. The samples holder is designed to be water cooled and instrumented in order to avoid annealing of the irradiation defects.

3:40 PM  
Proton Irradiation Induced Microstructural Evolution in Compositionally Graded Type 316L Stainless Steel: Xiang Liu1; Jingfan Yang2; Miao Song3; Xiaoyuan Lou2; Yongfeng Zhang4; Lingfeng He1; Daniel Schwen1; 1Idaho National Laboratory; 2Auburn University; 3University of Michigan; 4University of Wisconsin-Madison
    Hf was introduced into type 316L as a minor element that can potentially mitigate materials degradation due to radiation and/or corrosion. Compositionally graded type 316L stainless steel was fabricated through additive manufacturing. The specimen has 6 different levels of Hf concentration (0, 0.2 wt.%, 0.4 wt.%, 0.6 wt.%, 0.8 wt.%, and 1.0 wt.%) to demonstrate the viability of accelerated irradiation testing of gradient alloys. In this study, the graded specimen was proton irradiated up to 2.5 dpa at 360 ºC. The microstructural evolution behavior at different regions in the gradient sample was investigated using transmission electron microscopy and atom probe tomography techniques. Radiation-induced dislocation loops, voids, and segregation behavior in six different regions were compared to understand the effects of Hf concentration and grain size on the radiation resistance.

4:00 PM  
Sink Strength Effect on Bubble Formation in Helium-implanted Nanostructured Ferritic Alloys: Yan-Ru Lin1; Zhanfeng Yan2; David Hoelzer3; Lizhen Tan3; Steven Zinkle1; 1University of Tennessee; 2Peking University; 3Oak Ridge National Laboratory
    Helium in irradiated materials can cause unwanted material degradations. These effects may be mitigated by increasing the number of He trapping sites to control the bubble size or to shield He from the grain boundaries. In this study, 275keV He ions were implanted into 14YWT, 9YWT, CNA3, and CNA1 ferritic/martensitic nanostructured alloys at 500°C and 700°C. The bubble size increased and the density decreased with increasing temperature for all alloys. However, the variation was much more moderate for the 9YWTV and 14YWT alloy that had higher nanocluster density. The two CNA alloys exhibited good He trapping at 500°C but relatively inefficient trapping at 700°C. The 14YWT alloy appears to sequester the helium into smaller bubbles more effectively than other alloys. This may be attributed to the high sink strength associated with the high nanocluster density, or the difference of the He trapping ability (at different temperatures) for different nanoclusters.

4:20 PM  
Synergistic Irradiation and Ageing Effect on Microstructure and Mechanical Properties of Grade 92 at ~700C: Weicheng Zhong1; Lizhen Tan1; 1Oak Ridge National Laboratory
    Advanced nuclear reactors are designed to operate at higher temperatures for greater thermal efficiency, which result in the evaluation necessity of nuclear materials at higher temperatures. Current knowledge on high Cr ferritic martensitic (FM) steels are limited at 300-600℃. To understand the applicability of FM steels in higher temperature reactor designs, Grade 92 steel samples were irradiated in the High Flux Isotope Reactor at ~700℃ up to 14.6 dpa. The goal of this work is to evaluate the response of Grade 92 at such a high temperature irradiation environment. Microstructural and tensile performance were evaluated on the neutron-irradiated Grade 92 and compared with the counterparts thermally aged for a comparable time. Microstructure evolution, including grain structure, precipitates, dislocation, and cavities, were correlated with the mechanical response. The systematic study on the irradiated and aged samples revealed the synergistic effects of irradiation and thermal ageing on Grade 92 at 700℃.

4:40 PM  
Dislocation Loop Characterization Using STEM-Contrast Techniques in an Irradiated FCC Alloy: Pengyuan Xiu1; Lumin Wang1; Kevin Field1; 1University of Michigan
    Rel-Rod transmission-electron-microscopy (RR-TEM) has historically been used to characterize face centered cubic (FCC) materials after irradiation. In this study, we propose that scanning TEM (STEM) is a more convenient and robust method for imaging perfect and faulted types of dislocation loops in FCC materials. Three major zone axes [100], [110], [111] were used to perform on-zone STEM imaging for an irradiated model NiFe-20Cr alloy. With the aid of simulated dislocation loop morphology maps, all loop variant types were accurately identified by using the projected loop morphology on the [100] zone axis – other zone axes had ambiguous morphologies. Compared to on-zone/two-beam TEM bright-field and RR-TEM dark-field imaging, the on-zone [100] STEM imaging is the preferred method. With optimization of imaging parameters such as the collection angle, the on-zone [100] STEM is proven to robustly detect and identify dislocation loops without ambiguity and high visual contrast in irradiated FCC alloys.