Microstructural, Mechanical and Chemical Behavior of Solid Nuclear Fuel and Fuel-cladding Interface: Fuel-Cladding Interaction and Fission Products Retention
Sponsored by: TMS Structural Materials Division, TMS: Advanced Characterization, Testing, and Simulation Committee, TMS: Nuclear Materials Committee
Program Organizers: Yi Xie, Purdue University; Miaomiao Jin, Pennsylvania State University; Jason Harp, Oak Ridge National Laboratory; Fabiola Cappia, Idaho National Laboratory; Jennifer Watkins, Idaho National Laboratory; Michael Tonks, University of Florida
Wednesday 8:30 AM
March 22, 2023
Room: 26B
Location: SDCC
Session Chair: Yi Xie, Purdue University
8:30 AM Introductory Comments
8:35 AM Invited
Thermochemical Investigation of Advanced Reactor Fuels and Fuel-clad Chemical Interaction: Elizabeth Sooby1; 1University of Texas at San Antonio
Both the evolution of commercial reactor fuel forms and development of the next generations of reactors have sparked a demand to further understand and predict the behavior of advanced reactor fuel forms. Commercial fuel vendors are considering higher uranium density fuels to enhance fuel economy and better accommodate accident tolerant cladding, and the leading designs for small modular reactors are looking toward particle fuel architectures. A number of data gaps and/or inconsistencies exist for these lesser known fuel concepts. Of particular interest are the as fabricated microstructures of these novel fuels, the microstructural evolution during thermal synthesis and processing, and particularly fuel-clad and fission product-clad chemical interaction for novel fuel forms. Presented will be the experimental investigation of high uranium density fuel fabrication, including as-fabricated and thermally annealed microstructures. Additionally, limited experimental assessments of fuel and fission product interactions with silicon carbide.
9:00 AM Invited
Fuel Performance Analysis of an Annular Type Metallic U-10Zr Fuel: Di Yun1; Shilun Zheng1; Zhengyu Qian1; Linna Feng1; 1Xi'An Jiaotong University
Annular type metallic fuels have been proposed for ultra-high burn-up fast reactor applications. Traditional fission gas behaviors models rely on gas bubble swelling mechanisms which is sensitive to the hydrostatic stress induced by the early contact between the fuel and the cladding. It is perceived that such high sensitivity may not be suitable for the analysis of annular type metallic fuels. As a result, a mechanistic void swelling model has been applied to a fuel performance analysis of an annular metallic U-10Zr fuel. The characteristic performance calculation results are compared to experimental data from the AFC irradiation tests.
9:25 AM
Microstructural Characterization of the SiO2-SiC Interface of Oxidized TRISO Particles: Katherine Montoya1; Rachel Seibert2; Tyler Gerczak2; Elizabeth Sooby2; 1University of Texas at San Antonio; 2Oak Ridge National Laboratory
The SiC layer is the main structural component and barrier to fission product release in the TRISO particle configuration and damage to this layer can compromise the integrity of the fuel form. During an accident scenario, oxidants can be introduced to the core and expose the fuel compacts to an oxidizing atmosphere. Investigation of the stability of TRISO fuel under accident scenarios will aid in furthering the advancement of this fuel type. The compromised compact could allow for the oxidants to be introduced to the TRISO particles, specifically the SiC layer. Presented is a mapping of microstructures associated with the oxidized SiC layer from various thermochemical conditions. Characterization techniques will be used to investigate the microstructure, crystallinity, and composition of the SiO2-SiC interface. Comparisons are planned between the interface structure of irradiated and unirradiated particles. This insight is expected to provide context on the influence of irradiation on oxidation response.
9:45 AM
Advanced Characterization of Fuel-cladding Chemical Interaction in HT9 Clad U-Mo-Ti-Zr Metallic Fuel Irradiated in Advanced Test Reactor: Yachun Wang1; Jatuporn Burns1; Mukesh Bachhav1; Tiankai Yao1; Luca Capriotti1; 1Idaho National Laboratory
Abstract: Fuel cladding chemical interaction (FCCI) is recognized as an important fuel performance factor to support high burnup metallic fuel. Idaho National Laboratory’s and Department of Energy Advanced Fuel Campaign program experience on U-Zr fuel R&D focus on developing a mechanistic understanding of FCCI behavior and advanced fuel designs to mitigate and control FCCI. This study leveraged the state-of-the-art materials characterization methods, such as scanning electron microscopy (SEM), transmission electron microscopy (TEM), and atom probe tomography (APT), to better characterize the FCCI region formed in HT9 cladding by interaction with U-5Mo-4.3Ti-0.7Zr (wt.%) fuel during irradiation in the Advanced Test Reactor (ATR) to 2.2 % FIMA at peak inner cladding temperature (PICT) up to 650°C. Results are discussed in terms of the characteristics of FCCI and the effects from fuel composition and irradiation conditions on FCCI.
10:05 AM Break
10:20 AM
High Resolution Microscopic Studies on HT-9 Cladding from U-10Zr Fuel Irradiated at Fast Flux Test Facility: Mukesh Bachhav1; Tiankai Yao1; Luca Capriotti1; Jason Harp2; Maria Okuniewski3; Jonova Thomas4; Yachun Wang1; 1Idaho National Laboratory; 2ORNL; 3Purdue University; 4ANL
The performance of metallic fuel in a fast reactor is known to be depended on several factors including fuel composition, microstructural changes in fuels, fission products distribution, and fuel-cladding chemical interactions (FCCI). Study on evolution of microstructural changes in FCCI layer is highly desirable for a greater phenomenological understanding of FCCI and fuel constituent redistribution. To that end, we investigated HT-9 cladding on U-10Zr fuel subjected to a local burnup of 5.7 atom percent and a local inner cladding temperature of 615 °C. Atom Probe Tomography (APT) and Transmission Electron Microscopy (TEM) is employed to investigate the distribution of fission products and microstructural changes in HT-9. Specimens for TEM and APT were prepared from different part of FCCI to elucidate chemical distribution of fission products and microchemical changes in HT-9.
10:40 AM
Analysis of Secondary Phase Formation at U-10Mo Fuel/Cladding Interfaces During Manufacturing: Adam Koziol1; Miao Song2; Kayla Yano2; Alan Schemer-Kohrn2; Ayoub Soulami2; Vineet Joshi2; Samuel Briggs1; Elizabeth Kautz2; 1Oregon State University; 2PNNL
The monolithic Uranium 10 wt.% Molybdenum (U-10Mo) fuel system is a promising high assay low enriched uranium (HALEU) fuel system for the replacement of highly enriched uranium (HEU) fuels in high performance research reactors. During manufacturing of U-10Mo HALEU fuels, thermomechanical processes induce phase transformations and secondary phase formation at the fuel/cladding (i.e., U-10Mo/Al6061) interface. High resolution techniques (APT, TEM) are implemented to examine the microstructural evolution at the fuel/cladding and fuel/diffusion barrier interfaces for varying hot isostatic press (HIP) processing parameters.Modifying thermomechanical process parameters drastically changed the interaction layer thickness, the distribution of minor alloying elements, and the extent of phase transformation. TEM and APT identified various phases including α-U, γ -UMo, USi, U2 MoSi2C(.33), Mo-enriched phase, as well as various regions of minor alloying element enrichment. The distribution of minor alloying elements and observed phases is compared to interaction layer development.
11:00 AM
Interfacial Microstructure Evolution in Al6061-Al6061 HIP Bonded Samples for Cladding Applications on U-10Mo Monolithic Fuel: Rajib Kalsar1; Miao Song1; Cody Miller2; Nicole Overman1; Kenneth Johnson1; Timothy Roosendaal1; Curt Lavender1; Vineet Joshi1; 1Pacific Northwest National Laboratory; 2Los Alamos National Laboratory
Low-enriched uranium (LEU) alloyed with 10 wt% molybdenum (U-10Mo) has been identified as a promising alternative to high-enriched uranium (HEU) for the United States High Performance Research Reactors (USHPRR). The nominal configuration of the U-10Mo plate-type fuel is a metallic U-10Mo fuel foil, the thickness of which varies from 0.6 mm (0.025") to 0.2 mm (0.0085") depending on the reactor; a 25 μm thick Zr interlayer–diffusion barrier on either side, and an outer cladding of 6061 aluminum. The aluminum is clad on the co-rolled fuel using the HIP process. The impact of different HIP parameters on microstructure as well as bond strength was investigated. Interface microstructures were characterized using SEM, SEM-EBSD and TEM to quantify the oxide layer thickness, Mg2Si precipitate fraction and recrystallization across the interface. A correlation between interface microstructure and bond strength was also established.
11:20 AM
Numerical Modeling of AA6061 Cladding Diffusion Bonding for the U-10Mo Monolithic Fuel: Yucheng Fu1; Taylor Mason1; Rajib Kalsar1; Zhijie Xu1; Kriston Brooks1; Ayoub Soulami1; Vineet Joshi1; 1Pacific Northwest National Laboratory
The low enriched U-10Mo alloy is a promising fuel candidate for United States high performance research reactors. The fuel is designed to be encapsulated in the aluminum alloy 6061 (AA6061) cladding, which is diffusion bonded using the hot isostatic pressing (HIP). The diffusion bonding promotes a homogeneous AA6061/AA6061 bonding interface, preventing fuel corrosion and fission product release. To optimize the bonding quality and reduce experimental cost, an integrated FEM-Diffusion model was developed to model the aluminum cladding diffusion bonding, which involves the interface void closures and the formation of Mg2Al2O5 due to the Mg/Al diffusion. Combined with sensitivity analysis, it was found that temperature was the most dominant factor compared to the HIP time and applied pressure. With HIP temperature over 400 ºC, the interface voids will fully collapse, and the aluminum cladding can reach a desirable bonding strength.
11:40 AM
Irradiation Performance of Densely Packed UN TRISO Fuel in a 3D-Printed SiC Matrix: Christian Petrie1; Kory Linton1; Gokul Vasudevamurthy1; Danny Schappel1; Rachel Seibert1; Nicolas Woolstenhulme2; David Carpenter3; Andrew Nelson1; Kurt Terrani4; 1Oak Ridge National Laboratory; 2Idaho National Laboratory; 3Massachusetts Institute of Technology; 4Ultra Safe Nuclear Corporation
The Transformational Challenge Reactor fuel includes UN TRISO particles densely packed within a complex, 3D-printed SiC matrix that is intended to provide an additional barrier to fission product release. This work summarizes steady state and transient irradiations to evaluate fission product retention in compacts with varying fuel packing fractions. Most tests generally matched expectations, with no observable fission gas release. Particle failures were not observed during transient testing until the energy deposition far exceeded values expected in gas-cooled reactors, consistent with thermomechanical simulations. Fission gas release was observed during one set of low burnup steady state irradiations. Release was a result of matrix cracks that propagated through the particle coating layers in the regions with high matrix density. These failures are explained based on simulated thermal stresses and the dependence of the stress and crack propagation on the spatial distribution of the SiC matrix density.
12:00 PM
Atomistic Simulations of Silicon Carbide Layer in Tristructural Isotropic Fuel Particles: Cong Dai1; Michael Welland1; 1Canadian Nuclear Laboratories
The use of Tristructural Isotropic (TRISO) fuels in advanced reactors greatly inhibits the release of fission gas from fuel as long as the SiC layer remains intact. SiC is a stable and effective coating layer in TRISO fuel particles. Failure mechanisms of the SiC coating layer include metallic fission product degradation, CO corrosion, pressure vessel failure, and oxidation. Experiments found CO corrosion along grain boundaries (GBs) in SiC layer of the TRISO fuel particle. Crystal orientations of adjacent grains and GB plane determine the type of GBs, and they may exhibit different resistance to CO corrosion. Stress-corrosion cracking would occur along low-resistance GBs in SiC layer, which may cause the release of radioactive fission products from the fuel kernel. This work simulate experimentally observed GBs in SiC. Various GB properties are calculated, which can be useful to determine which type of GBs has the highest resistance to CO corrosion.