Accelerated Materials Evaluation for Nuclear Applications Utilizing Irradiation and Integrated Modeling: Current and Advanced Structural Materials II
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee
Program Organizers: Assel Aitkaliyeva, University of Florida; Peter Hosemann, University of California - Berkeley; Samuel Briggs, Oregon State University; David Frazer, Los Alamos National Laboratory
Thursday 8:30 AM
February 27, 2020
Room: Theater A-8
Location: San Diego Convention Ctr
Session Chair: Peter Hosemann, University of California - Berkeley
8:30 AM
Fabrication and Characterization of Massive Crack-free Delta Phase-zirconium Hydride for High-performance Moderator Application: Xunxiang Hu1; Wei Tang1; Kurt Terrani1; 1Oak Ridge National Laboratory
The successful implementation of ZrHx moderator in advanced reactors requires consistent and affordable production along with hydrogen retention throughout the reactor life. A fully programmable hydriding system with continuous hydrogen partial pressure and flow control to facilitate processing of massive delta-phase zirconium hydride has been developed at Oak Ridge National Laboratory. Characterization of the produced ZrHx includes X-ray powder diffraction to identify the present phases, LECO H and O analysis to quantify hydrogen, and X-ray Computed Tomography to visualize the possible cracks. The results indicate that crack-free ZrH1.6±0.1 with >10 cm3 have been successfully produced. Maintaining constant hydrogen concentration in ZrHx throughout reactor life is necessary to ensure safe, economic, and reliable operation. A possible solution is to develop cladding materials that prevent hydrogen release at elevated temperatures. Thermal stability of the encapsulated ZrHx will also be assessed through characterizing the hydrogen concentration before and after heat treatments.
8:50 AM
Neutron Irradiation Damage and Fission Product Transport in the SiC Layer of TRISO Fuel Particles: Subhashish Meher1; Isabella van Rooyen1; Chao Jiang1; 1Idaho National Laboratory
It has been recently discovered that the precipitation of intragranular palladium-containing fission products in the SiC layer of neutron-irradiated tristructural isotropic (TRISO) fuel particles possibly occurs via a novel dual-step nucleation mechanism. Direct observations of Pd silicide imprinting into morphological templates of α-SiC precipitates in neutron-irradiated SiC layer of TRISO fuel will be discussed, with examples from selected particles from both the Advanced Gas Reactor (AGR)-1 and AGR-2 fuel irradiation experiments. The physical understanding of intragranular fission product precipitation has been studied by both advanced microscopy and first principle calculations. Along with this, the large-scale precipitates of fission products at IPyC layer, and grain boundaries of SiC have been found to be constituted of separate regions of fission products.
9:10 AM
Three-dimensional Analysis of the IPyC/SiC Interface in Irradiated TRISO Fuel Particles: Tyler Gerczak1; Rachel Seibert1; 1Oak Ridge National Laboratory
The silicon carbide (SiC) layer of tristructural-isotropic (TRISO) coated particle is the primary barrier to release of fission products not retained in the fuel kernel. However, diffusion into and reaction with the SiC layer has been observed for select fission products (e.g. palladium and silver). The nature of the inner pyrolytic carbon (IPyC)/SiC interface likely governs the interaction of these fission products with the SiC layer. Three-dimensional reconstruction image analysis of the IPyC/SiC layer from as-fabricated TRISO particles are compared with the IPyC/SiC interface from irradiated particles from the AGR-2 irradiation experiment with varying silver retention behaviors. The analysis provides a quantitative description of the IPyC/SiC interface – IPyC/SiC interface stitching and open porosity – as well as insights into their influence on fission product interactions with the SiC layer.
9:30 AM
Fabrication and Charaterization of High Burnup Nuclear Fuel Surrogate for the Anlysis of Fuel Fragmentation Phenomenon: Jae Joon Kim1; Ho Jin Ryu1; 1KAIST
To identify the microstructural cause of nuclear fuel fragmentation phenomenon in LOCA conditions, CeO2, a surrogate of the UO2, was implanted with xenon and krypton ions. Fine grain-sized CeO2 with a porosity of 10 % and a grain size of about 550 nm, and normal-grain-sized CeO2 with a porosity of 4 % and a grain size of 7 m were used for the irradiation test for the rim part and center part of the nuclear fuel, respectively. The irradiated CeO2 were annealed at temperatures between 300 °C and 1200 °C, and the microstructure of each specimen was analyzed by TEM. Each specimen was then rapidly heated again at a rate of several tens °C / s for the LOCA situation simulation, and then the behavior of the implanted gas bubble was also investigated by TEM. The results of this experiment show how fission gas bubbles behave in the LOCA situation.
9:50 AM
Microstructural Changes and Corrosion of Proton-pre-irradiated Hastelloy N in FLiNaK Molten Salt: Andres Morell-Pacheco1; Lingfeng He2; Ruchi Gakhar2; Yachun Wang3; Adam Gabriel1; Lin Shao1; 1Texas A&M University; 2Idaho National Laboratory; 3Rensselaer Polytechnic Institute
UNS N10003 (Hastelloy N) is a leading candidate for use as a structural material in molten salt reactors (MSRs). A complete understanding of the evolution of alloy UNS N10003 under diverse conditions, needed for commercial implementation of MSRs, has yet to be realized. Our work aims at understanding the coupling effect of radiation damage on the alloy’s corrosion sensitivity in fluoride salts. UNS N10003 was irradiated using 2.5 MeV protons to 0.25, 0.5, and 1.0 local dpa and subsequently exposed to molten lithium-sodium-potassium eutectic fluoride (FLiNaK) salt at 700°C for 500 hours within carbon crucibles in an argon atmosphere. Cross-sectional microstructural changes were characterized using focused ion beam (FIB) technique and scanning transmission electron microscopy (S/TEM) coupled with energy dispersive X-ray spectroscopy (EDS), electron energy loss spectroscopy (EELS) and precession electron diffraction (PED) capability.
10:10 AM Break
10:25 AM
On the Role of Heterogeneity in Concentrated Solid‒solution Alloys in Enhancing their Irradiation Resistance: Shijun Zhao1; 1City University of Hong Kong
Concentrated solid‒solution alloys (CSAs) demonstrate excellent mechanical properties and promising irradiation resistance depending on their compositions. Existing experimental and simulation results indicate that their heterogeneous structures induced by the random arrangement of different elements are one of the most important reasons responsible for their improved properties. Nevertheless, the nature of this heterogeneity remains unclear. Here we scrutinize the role of heterogeneity played in damage evolution in different aspects through atomistic simulations. Different effects induced by atomic-level heterogeneity are considered, including lattice misfit, thermodynamical mixing, point defect energetics, point defect diffusion, and dislocation properties. Our results reveals that the defect evolution in CSAs only has weak relations with most of these parameters, suggesting the complexity of defect dynamics in these complexed alloys. However, our results indicate that defect properties may be the most important factors in influencing the irradiation performance of CSAs.
10:45 AM
Diffusion of Fission Products in Virgin Nuclear Graphite: Kevin Graydon1; Mikhail Klimov1; Edward Dein1; Kevin Coffey1; Yongho Sohn1; 1University of Central Florida
The diffusivities of fission products in various grades of nuclear graphite are needed to aid in the design and licensing requirements for the next generation of nuclear reactors. Discussed are the methodologies and the resultant findings of the diffusivities and transport mechanisms of fission products, Ruthenium and Silver, through a graphite collection. The graphite selection encompasses AXF-5Q, ZXF-5Q, IG-110, NBG-18, PCEA, and HOPG. By utilizing the thin film approach and secondary ion mass spectroscopy (SIMS), the diffusivities of the fission products can be experimentally determined. An initial value of the diffusivity of Ruthenium in AXF-5Q at 600°C has been found to be 1.46E-16 ± 6.12E-17 (95%) cm2/s. Based on the preliminary data from the SIMS depth profile, more than one mechanism of diffusion may play a role in Ruthenium transport through the graphite. Diffusion of each fission product in each graphite will be analyzed and compared.
11:05 AM
Irradiation Behavior of Mechanically Processed Zr-Nb Multilayers at Very High Doses: Madhavan Radhakrishnan1; Daniel Savage2; Marko Knezevic2; John Watt3; Yongqiang Wang3; Katherine Jungjohann4; Nathan Mara5; Osman Anderoglu1; 1University of New Mexico; 2University of New Hampshire; 3Los Alamos National Laboratory; 4Sandia National Laboratory; 5University of Minnesota
Past studies have showed that PVD-processed multilayers provide enhanced radiation damage resistance owing to large fraction of interfaces. This work investigates the radiation resistance of bulk nanolayered zirconium-niobium composites exposed to very high doses. Three Zr/Nb multilayers with individual layer thicknesses of 15 nm, 45 nm, 90 nm were synthesized by accumulative roll-bonding process. Micro-Vickers hardness values follow a Hall-Petch strengthening trend with layer thickness. The multilayers were subjected to a self-ion irradiation with 7 MeV Zr2+ ion beam at 500°C. Irradiation caused a maximum dose of ~82 dpa at a depth of 1.5 μm. Cross-sectional TEM examination indicates that, in all multilayers, heavy dose ion-irradiation has induced a heterogeneous fragmentation of Zr, Nb layers and a chemically homogeneous mixed layer beneath the irradiated surface. The extent of sub-surface microstructural changes in multilayers correlates with the SRIM profile. Here, we report the correlation between layer thicknesses and evolution of radiation microstructures and mechanical property during irradiation.
11:25 AM
High-Throughput Synthesis and Ion Irradiation of High-Entropy Alloys using Additive Manufacturing: Michael Moorehead1; Michael Niezgoda1; Calvin Parkin1; Chuan Zhang2; Phalgun Nelaturu1; Mohamed Elbakhshwan1; Kumar Sridharan1; Dan Thoma1; Adrien Couet1; 1University of Wisconsin - Madison; 2Computherm LLC
Development of materials for nuclear application often entails a rigorous series of testing and characterization which can encumber the pursuit of potentially novel materials. Such materials include high-entropy alloys (HEAs): multicomponent alloys which often exhibit desirable mechanical and radiation-tolerant properties. While several HEAs have shown excellent phase stability and irradiation resistance, the origins of these properties are still unclear – further challenging nuclear HEA development. In light of this, high-throughput (i) synthesis, (ii) ion irradiation, and (iii) characterization techniques have been employed to accelerate HEA development. Additive manufacturing was used to synthesize 25-coupon arrays of different alloys from the Cr-Fe-Mn-Ni and Mo-Nb-Ti-V composition spaces, from binary to quaternary alloys, using elemental powders. Sample arrays were irradiated with self-ions to 10 dpa and characterized using nano-indentation, XRD and profilometry measurements. Phase evolution has been supported by CALPHAD modeling and compositional trends in radiation response have been compared to theory and literature.
11:45 AM
Kinetic Study on the Evolution of Nano-ceramic Coatings Under Heavy Ions Irradiation: Matteo Vanazzi1; Luca Ceseracciu2; Gaelle Gutierrez3; Celine Cabet3; Marco G. Beghi4; Fabio Di Fonzo1; 1Center for Nano Science and Technology (CNST) - IIT; 2IIT; 3CEA; 4Politecnico di Milano
Heavy ions in the KeV/MeV range are quite appropriate to simulate neutrons. Previously, we have reported on the evolution of amorphous alumina coatings under ions irradiation (up to 450 dpa) showing a general radiation-induced crystallization trend. Here, we concentrate on the low dpa, to evaluate the first stages of irradiation and to obtain neutrons-compatible data. Irradiations are performed at different temperatures, to decouple the thermal contribution from the radiation-induced effects. The evolution seems strictly temperature-dependent, with no structural changes at the lower temperature. At the higher levels, results show the formation of different crystals, depending on test conditions. A kinetic model for the grain growth is proposed, based on the experimental data. Mechanically, an evident size-effect is manifested. The formation of nano-crystalline domains increase rapidly the hardness, in accordance with the Hall-Petch model. Nevertheless, the hardening does not cause any brittleness, but it improves the mechanical properties of the material.