Microstructural, Mechanical and Chemical Behavior of Solid Nuclear Fuel and Fuel-cladding Interface: Poster Session
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
Tuesday 5:30 PM
March 21, 2023
Room: Exhibit Hall G
Location: SDCC
Session Chair: Yi Xie, Purdue University
Cancelled
O-13: An Efficient and Oxidation-preventive Method for UN Surrogate Pellets Fabrication: Logan Joyce1; Yi Xie1; 1Purdue University
Uranium mononitride (UN) is a promising fuel material due to the high U loading and thermal conductivity. However, due to the fast oxidation, making a densified UN fuel pellet is challenging. The conventional sintering method requires an extremely low-oxygen atmosphere and uniform powder particles, which needs a long and complex powder processing. To tackle the problems, we developed an efficient and oxidation-preventive method for producing highly densified UN surrogate pellets using the fast sintering system. The fast sintering process facilitates higher surface diffusion, which results in thicker bonds between particles. The pellets were sintered at 1800 °C for 30 seconds, leading to 95 % relative density. The powder particles were in a wide range of size, but the sintered pellets show uniform microstructure with grains consistent in 10 – 25 µm. Therefore, this method can achieve densified and oxidation-free UN surrogate pellets efficiently and reduce powder processing efforts.
O-14: Calculation of Grain Boundary Diffusion Coefficients in Gamma U-Mo Using Atomistic Simulations: ATM Jahid Hasan1; Benjamin Beeler1; 1North Carolina State University
The gamma U-10Mo alloy has been selected for the conversion of US High-Performance Research Reactor (HPRR) fuels from high-enriched uranium to low-enriched uranium. Although UMo alloys have high uranium density and good overall irradiation performance, the irradiation-induced swelling and creep remain important effects that influence the mechanical integrity of the fuel. Fuel performance models need fundamental properties, such as diffusion coefficients, as input to accurately describe the fuel evolution. In this study, we quantify the diffusivity along grain boundaries in UMo fuel. The grain boundary diffusivities of UMo alloys are obtained utilizing molecular dynamics simulations for a temperature range of 600-1200 K. We observe that U diffusivity in the grain boundary is significantly higher than Mo diffusivity, and the total diffusivity decreases with increasing Mo content. The information gathered in this work can inform creep and fission gas swelling models and help understand various phenomena related to UMo fuel performance.
O-15: Experimental Methods for Comprehensive PIE of Test Fuel Rods: Chaitanya Peddeti1; 1UC Berkeley
Post-irradiation examination (PIE) of reactor fuel components is essential to the analysis of reactor operating conditions. However, the various analysis techniques paired with the process of sample preparation is expensive and time intensive. To combat this, we have designed a multifaceted, optical analysis system that can prepare samples for mechanical testing and microscopy, as well as perform various PIE techniques within a hot cell environment. These measurements include surface element analysis, released isotope analysis, and thermal conductivity. The system utilizes laser ablation-inductively coupled plasma mass-spectroscopy (LA-ICP-MS), along with light induced breakdown spectroscopy (LIBS) to perform PIE on spent fuel materials. We expect this system to accelerate nuclear fuel materials research. We have shown the basic setup is viable to perform laser processing on hot samples in a safe manner, as well as the ability to detect gases released from laser processing.