Ceramic Materials for Nuclear Energy Research and Applications: Advanced Ceramics Concepts
Sponsored by: TMS Extraction and Processing Division, TMS Structural Materials Division, TMS Light Metals Division, TMS: Advanced Characterization, Testing, and Simulation Committee, TMS: Energy Committee, TMS: Nuclear Materials Committee
Program Organizers: Xian-Ming Bai, Virginia Polytechnic Institute and State University; Yongfeng Zhang, University of Wisconsin; Larry Aagesen, Idaho National Laboratory; Vincenzo Rondinella, Jrc-Ec
Thursday 2:00 PM
March 18, 2021
Room: RM 51
Location: TMS2021 Virtual
Session Chair: Haiming Wen, Missouri University of Science and Technology; Xunxiang Hu, Oak Ridge National Laboratory
2:00 PM Invited
Development of Yttrium Hydride for High Temperature Moderator Application: Xunxiang Hu1; Kurt Terrani1; 1Oak Ridge National Laboratory
The use of metal hydrides, especially the zirconium hydride (ZrHx), as high-performance moderators in advanced reactors has long-term precedent. However, the application of ZrHx in nuclear system required careful management of the moderator temperature to avoid hydrogen desorption at elevated temperatures. Yttrium hydride (YHx) is more attractive for high temperature moderator application, attributed to its superior thermal stability. The successful deployment of high temperature YHx moderator in advanced reactors requires development of a consistent and affordable production pathway along with the establishment of a complete database of the thermomechanical and neutronics properties and irradiation response. In this presentation, we will demonstrate the successful fabrication of crack-free bulk YHx and discuss the detailed characterization of the as-fabricated YHx, including hydrogen concentration determination, crystal structure, thermal stability, thermophysical and mechanical properties, and thermal neutron scattering properties. In addition, the impact of neutron irradiation on the thermomechanical properties will also be discussed.
2:30 PM
Ionization Effects on Damage Accumulation Behavior in SiC: Lauren Nuckols1; Miguel Crespillo1; Yanwen Zhang2; William Weber1; 1University of Tennessee Knoxville; 2Oak Ridge National Laboratory
Characterization of ion-irradiation effects in materials is needed to develop performance models for materials and devices in reactor environments. In SiC, ion-induced ionization facilitates defect annealing along the ion path and competes with damaging processes. Ionization effects are present in SiC under all intermediate to high energy ion irradiations and fusion conditions. Here, the sensitivity of disordering processes to changes in total ionization energy deposition is examined for a range of ions in 4H- and 3C-SiC. Damage accumulation was examined using Rutherford backscattering spectrometry in channeling geometry. Ion beam induced luminescence was used in situ to characterize defect formation and evolution from low damage energy ions. There is a direct relationship between incident ion atomic number and its disordering sensitivity to ionization changes. The harder recoil spectrum induced by heavier ions are less coupled to ionization induced annealing and produce more thermally stable defects compared to lighter mass ions.
2:50 PM
Microstructural Characterization of Radiation Effects in 3D printed SiC: Timothy Lach1; Takaaki Koyanagi1; Chad Parish1; Thak Sang Byun1; Kurt Terrani1; 1Oak Ridge National Laboratory
Most of the reactor core components for the Transformational Challenge Reactor (TCR) will be built through additive manufacturing (AM). The performance of the SiC fuel matrix is particularly important because it must demonstrate sound structural stability and acceptable heat transfer properties, as it serves as the fuel particle matrix, an additional barrier to fission product release, and a heat transfer medium. A recently developed binderjet printing process combined with chemical vapor infiltration (CVI) is being leveraged to produce the SiC fuel matrix for the TCR core. However, detailed microstructural evaluation of this 3D printed SiC and its stability under irradiation conditions is required for reactor design and material qualification purposes. High-resolution electron microscopy is being used to evaluate the microstructural evolution of 3D printed SiC after ion irradiation and neutron irradiation. Detailed comparisons will be made with reference chemical vapor deposition (CVD) SiC.
3:10 PM Invited
Microstructure and Chemical States of Fission Products in Irradiated AGR-1 and AGR-2 TRISO Particle UCO Fuel Kernels: Yong Yang1; Isabella van Rooyen2; Zhenyu Fu1; Boopathy Kombaiah2; 1University of Florida; 2Idaho National Laboratory
Over the past three years, extensive microstructural characterizations were conducted on irradiated AGR-1 and AGR-2 TIRSO fuel UCO kernels. It was found that the irradiated fuel kernel mainly consists of UO2 and UC, while the UO2 is the dominating fuel phase. UC phase contains a significant amount of fission products including Zr, Nb, Mo, Ru, Tc and Rh. The post irradiation safety testing significantly promotes the precipitations of U2Ru(Rh)C2 and UMo(Tc)C2 phases. Fission gas bubbles were only observed in UC or UMoC2 phases. Results from EDS analysis and APT study show that the lanthanides, e.g., Nd, are more likely to enrich within the UO2 phase. Comparison between the AGR-1 and AGR-2 fuel particle kernels shows the differences in irradiated microstructures and chemical states of fission products may result from a combination of irradiation temperature, fuel geometry, and chemical composition. However, irradiation temperature probably plays a more deterministic role.
3:40 PM Invited
Oxidation Behavior of TRISO Fuel Materials: Haiming Wen1; Adam Bratten1; Visharad Jalan1; 1Missouri University of Science and Technology
While high-temperature gas reactors use pure helium as a reactor coolant, in some accident scenarios significant amounts of moisture or air can be introduced into the helium coolant and reactor core. It is important to understand the oxidation and degradation processes exhibited by both TRISO particles (particularly the SiC layer) and matrix graphite under these conditions. In this study, matrix graphite and surrogate TRISO particles (with ZrO2 kernel) were subjected to oxidation by air and/or moisture under conditions relevant to the air/moisture ingress accidents, using thermogravimetric analysis (TGA) – differential scanning calorimetry (DSC) or custom-built tube furnace setup. The kinetic parameters of oxidation were measured through the oxidation studies at different temperatures with different partial pressure of oxygen or water vapor. The microstructures of the materials before and after oxidation were carefully characterized. The oxidation mechanisms were ascertained in relation to the oxidation conditions and microstructures of the materials.
4:10 PM
Evolution of Ion Irradiated Nitride Ceramics Properties for Coated Particle Fuel Systems: Adrien Terricabras1; Alicia Raftery2; Andrew Nelson2; Steven Zinkle1; 1University of Tennessee; 2Oak Ridge National Laboratory
Novel coated particle fuel systems are of particular interest within the research community lately with possible applications ranging from space reactors to Small Modular Reactors and more. Using a uranium oxycarbide or uranium nitride core with alternative coatings such as ZrC or ZrN, they are designed to alleviate current TRISO shortcomings such as loss of integrity of the SiC layer and fission product attacks. Ion irradiation was performed on polycrystalline Si3N4 and ZrN using 15 MeV Ni5+ ions. Midrange doses varied from 1 to 50 dpa and temperatures from 300 to 700 °C. Thermal conductivity of the ion irradiated region was measured using thermal wave methods. Volumetric lattice swelling of the ceramics was determined by grazing incidence X-ray diffraction. Defect evolution was tracked using Transmission Electron Microscopy, while nanoindentation was performed to quantify the ceramics’ mechanical properties evolution. The nitrides behavior will be compared with SiC and other ceramics.
4:30 PM
On the Role of Neutron Irradiation Damages on Fission Products 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.