Advanced Characterization and Modeling of Nuclear Fuels: Microstructure, Thermo-physical Properties: Nuclear Fuels Thermo-physical Properties - Experiment
Sponsored by: TMS Structural Materials Division, TMS: Advanced Characterization, Testing, and Simulation Committee, TMS: Energy Committee, TMS: Nanomechanical Materials Behavior Committee, TMS: Nuclear Materials Committee
Program Organizers: David Frazer, Idaho National Laboratory; Fabiola Cappia, Idaho National Laboratory; Tsvetoslav Pavlov, Idaho National Laboratory; Peter Hosemann
Wednesday 8:30 AM
March 2, 2022
Location: Anaheim Convention Center
Session Chair: Tsvetoslav Pavlov, Idaho National Laboratory
8:30 AM Invited
Accelerating Nuclear Fuel Qualification through Multiscale Models: Joshua White1; Tammie Nelson1; David Andersson1; 1Los Alamos National Laboratory
Qualification of nuclear fuels for use in modern nuclear reactors currently requires lengthy, and costly, irradiation test campaigns to provide assurance of the reactor performance under a variety of scenarios. The Nuclear Regulatory Commission is interested in advanced modeling and simulation methods to expedite the qualification of fuels by utilizing improved mechanistic models combined with separate effects experiments to decrease the number of required data sets to commission fuels for use. The use of innovative machine learning (ML) tools coupled with state-of-the-art experimental measurements will be used to benchmark the models as well as reduce the uncertainty, providing confidence in fuel performance in-pile. This talk will focus on UN and UC, two potential high-impact fuels for microreactors and gas-cooled reactors. The developed ML models will be integrated with multiphysics engineering-scale neutronics and reactor simulations that ultimately define safety margins for novel nuclear fuels within specific reactor conditions.
Diffusion Coefficients of Zr- and Cr-based Binary Systems for Simulation of Cr-coated Zircaloy Nuclear Fuel Cladding: Ella Kartika Pek1; Wei Zhong1; Alexander Butler1; Ji-Cheng Zhao1; 1University of Maryland, College Park
Chromium-coated Zircaloy is one of the accident-tolerant fuel (ATF) claddings. A thin Mo, Nb, or Ta interlayer may be placed between Cr and Zircaloy to serve as a diffusion barrier to reduce reaction between Cr and Zircaloy under accident conditions. Diffusion coefficients of Zr-(Cr, Mo, Nb, Ta), Zircaloy-Cr, and Cr-(Mo, Nb Ta) binary systems are essential for accurate prediction of the ATF lifetime. Diffusion coefficients consisting of these binary systems are being made and will be annealed at 6 temperatures. This allows us to obtain diffusion coefficients in order to establish a reliable and comprehensive diffusion coefficient database for the 8 binary systems, to fill a knowledge gap in diffusion data for the Cr-coated Zircaloy ATF.
Thermo-physical Properties of the Ternary (U2Cr)N3 Phase: Yulia Mishchenko1; Denise Adorno Lopes1; Elina Charatsidou1; 1KTH Royal Institute of Technology
Thermal conductivity of the ternary compound (U2Cr)N3 was measured experimentally by the Transient Hot Bridge technique and compared with the results predicted using modeling calculations. The ternary phase (U2Cr)N3 was obtained experimentally for the first time in the composite pellets by combining UN powders and CrN powders and sintering using SPS (Spark Plasma Sintering) method. The crystal structure of the ternary U2CrN3 phase was investigated and the positions of the Cr atoms were obtained. The elastic properties of the (U2Cr)N3 phase were also obtained by modeling calculations. The uranium density of the ternary compound (U2Cr)N3 is higher than that of UO2 making it a promising candidate for the Accident Tolerant Fuel (ATF) concept.
Thermal Transport Behavior of Pristine and Zirconium-doped Alpha-Uranium: Zilong Hua1; David Hurley1; 1Idaho National Laboratory
The U-Zr based metallic fuel is a promising candidate for the next generation of fast reactors. The structure-induced thermal anisotropy of alpha-U has been reported, but accurate knowledge and its dependence on temperature is still lacking. In this talk, recent research results of thermal conductivity in pristine alpha-U are reported. Experimental data in a wide temperature range are compared with DFT and MD simulation to investigate the transport mechanisms. In addition, the influence from different mass percentages of Zr-doping is studied to validate the modeling work that can be used to predict the in-reactor performances with irradiation-induced point defects. These new insights are expected to aid in fuel performance modeling and fuel design.
10:00 AM Break
Bulk Thermal Conductivity Measurement of Fuels and Surrogates: Kunal Mondal1; Scott Middlemass1; Peng Xu1; Isabella van Rooyen1; 1Idaho National Laboratory
Thermal conductivity, a thermal property which characterizes the capacity for a material to conduct heat energy under a specific temperature gradient and these properties can be assessed on the global (bulk) and the local (micro, nano) scales, using several techniques. Variations or anomalies of material properties on the micro/nano scale may disturb the thermal properties of the bulk material, thus having a possible effect on the overall thermomechanical performance of the bulk counterpart. In this study, we are investigating the effect of material variations –packing fraction and uranium content – to the overall thermal properties of fueled TRi-structural ISOtropic (TRISO) compacts (surrogate (ZrO2), and natural Uranium). To measure bulk thermal conductivity of TRISO fuel compacts, a steady-state measurement system based on the guarded-comparative-longitudinal heat flow technique will be designed, built, and analysis will be made for use at high temperatures.
The Influence of Radiation-induced Microstructural Defects on the Optical and Elastic Properties of Ceramic Nuclear Fuels: Amey Khanolkar1; Zilong Hua1; Cody Dennett1; Joshua Ferrigno2; Marat Khafizov2; J. Mann3; David Hurley1; 1Idaho National Laboratory; 2The Ohio State University; 3Air Force Research Laboratory
A fundamental understanding of radiation effects on thermophysical properties is critical for the development of advanced nuclear fuels. In this talk, the impact of early-stage defect accumulation on the optical and elastic properties of single crystal thorium dioxide (ThO2) is explored. Lattice defects are seeded by irradiating the crystals with energetic protons. Changes in the optical properties brought about by point defects and color centers are probed using ellipsometry and photoluminescence spectroscopy. Depth-dependent opto-elastic changes arising from structural damage are measured using time-domain Brillouin scattering. Surface-confined elastic properties are further measured using a laser-induced transient grating method. The all-optical and non-destructive approach shows promise in quantifying the population of lattice defects in the early stages of damage accumulation that cannot be resolved conventional characterization techniques. Measurements of dose-dependent elastic property changes may also provide further insight into lattice swelling and enable validation of fuel performance model predictions.