Ceramic Materials for Nuclear Energy Research and Applications: Processing and Evaluation of Alternative Fuels and Materials
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee, TMS: Mechanical Behavior of Materials Committee, TMS: Energy Committee
Program Organizers: Walter Luscher, Pacific Northwest National Laboratory; Xian-Ming Bai, Virginia Polytechnic Institute and State University; Lingfeng He, North Carolina State University; Sudipta Biswas, Idaho National Laboratory; Simon Middleburgh, Bangor University
Tuesday 2:30 PM
March 21, 2023
Room: 28B
Location: SDCC
Session Chair: Simon Middleburgh, Bangor University
2:30 PM Invited
Zirconia-coated Uranic Fuel Particles Processing and In Situ Sintering Characterisation: Phylis Makurunje1; Gareth Stephens1; Simon Middleburgh1; 1Nuclear Futures Institute
Gel casting was studied as an encapsulation-aiding process for the manufacture of uranic coated fuel particles (CFP) for gas and light water reactors. Encapsulation involved the envelopment of colloidal droplets within a zirconium oxynitrate precursor solution during gel casting. The enveloping zirconium oxynitrate precursor phase was to result in a zirconium oxide coating on uranium dioxide kernels. The surfactants-conditioned uranium dioxide colloid was modified by an alginate cross-linking matrix prior to encapsulation, to facilitate solidification by ion exchange when contacted with calcium chloride. Pourbaix diagrams were employed to analyse zirconium oxynitrate solution stability with varying pH and complemented by experimental zeta potential studies. The kernels' sintering behaviour and thermal expansion were analysed by an optical dilatometer system up to 1600 °C in argon and characterised by XRD, Raman spectroscopy and SEM. Encapsulated wet granulation is a promising method for manufacturing coated fuel particles.
3:00 PM Invited
Oxidation Behavior and Mechanisms of the SiC Coating in TRISO Fuel Particles: Haiming Wen1; Adam Bratten1; Visharad Jalan1; 1Missouri University of Science and Technology
While high-temperature gas reactors use helium as a coolant, in some accident scenarios significant amounts of air or water vapor can be introduced into the coolant and reactor core. It is important to understand the oxidation behavior and mechanisms of TRISO particles (especially the SiC coating layer) under these conditions. In this study, surrogate TRISO particles were subjected to oxidation in oxygen or water vapor containing environments at different temperatures with different partial pressures of oxidants. The microstructures of the SiC coating and the oxide layer after oxidation were carefully characterized via different advanced techniques. The oxidation mechanisms were ascertained in relation to the oxidation conditions and microstructures of the SiC. Passive oxidation occurred at high oxygen partial pressure. At low partial pressure of oxygen, the oxidation mechanism was determined to be a mixture of passive and active oxidation; nanocrystalline grain size promotes activation oxidation, followed by redeposition of SiO2.
3:30 PM
Phase Equilibria and Thermodynamics of Tri-carbide Fuels for Nuclear Thermal Propulsion: Ronald Booth1; Juliano Schrone Pinto1; Erofili Kardoulaki2; Ken Mcclellan2; Jhonathan Rosales3; Theodore Besmann1; 1University of South Carolina; 2Los Alamos National Laboratory; 3NASA
Nuclear thermal propulsion (NTP) is an exceptional alternative to chemical propulsion for space travel due to its increased power density providing twice the specific impulse (the ratio of thrust to propellent mass flow rate) of chemical propulsion. Uranium monocarbide is the nuclear material of most interest for NTP due to its high thermal conductivity and high uranium atom density. However, NTP requires a nuclear reactor with high temperature stability to over 2700oC, where UC has a low melting point of below that at 2420oC. To address this issue there is interest in using a tri-carbide nuclear fuel of UC-ZrC-TaC. In the current work this system has been thermodynamically modelled to predict phase equilibria and behavior to determine temperature stability over 2700oC.
3:50 PM
Thermomechanical Characterization of Advanced Reactor Alloys and Composites Exposed to High-temperature Gas Environments: William Searight1; Leigh Winfrey2; 1Pennsylvania State University; 2SUNY Maritime College
Advanced reactors have much more demanding operating environments than current commercial power reactors in terms of high temperature/heat flux, radiation flux and corrosion. To enable commercial development of these reactors, structural materials must be developed to survive these environments for extended operation without maintenance. In this work, samples of TZM alloy and Highly Oriented Pyrolytic Graphite (HOPG) have been selected for their potential of long-term endurance in high-temperature and corrosive environments. These samples were held in a backfilled tube furnace at 1240 °C for 40 hours to identify thermomechanical property degradation in gas mixtures of argon, hydrogen, and helium. Surface composition and microstructure were analyzed using Scanning Electron Microscopy (SEM) techniques, XRD and optical profilometry and thermomechanical properties using Thermogravimetric Analysis (TGA) and microindentation. Time of Flight Secondary Ion Mass Spectroscopy (TOF SIMS) was used to map hydrogen content of TZM post-exposure.
4:10 PM Break
4:30 PM Invited
Improving Uranium Mononitride Behaviour using Uranium Diboride Addition: Joel Turner1; Tim Abram1; Qusai Mistarihi1; James Buckley1; 1University of Manchester
Uranium nitride remains a promising fuel material for commercial reactor systems, due to its high thermal conductivity and uranium loading. Compared to uranium dioxide, UN can have reduced fuel centreline temperatures and lower enrichment costs, enabling more expensive cladding solutions to be utilised and introducing further accident tolerance. A key challenge for UN deployment in light water reactors is the UN-steam reaction, in which UN rapidly reacts to form oxide powder below expected operating temperatures. A potential solution to this may be the addition of uranium diboride, which has been seen to increase the onset temperature of the reaction, while also potentially improving thermal conductivity and providing a required burnable poison. Data on UN-UB2 composites will be reported, including synthesis and fabrication, pellet characterisation, oxidation performance and thermal properties.
5:00 PM
Silica Formation on SiC Following Steam Attack: Dina Elgewaily1; Jacob Eapen1; 1North Carolina State University
Silicon carbide composites (SiC/SiC) are being considered as accident-tolerant fuel (ATF) cladding of light water reactors due to their low chemical reactivity, which allows a longer time window for accident response. Under high temperature and pressure, SiC reacts with steam to form silica and other oxides together with several volatile gases. In this investigation, Raman spectroscopy is used to identify the elemental and compositional nature of the formed oxide layer after a steam attack. At 1200°C and at pressures ranging from 0.45 to 1.38 MPa, the formation of α-cristobalite phase of SiO2 is established. At lower pressures, the dominant oxide appears to be non crystalline although trace amounts of cristobalite can be observed. A sharp increase in the integrated Raman intensity ratio from 0.1 MPa to 0.45 MPa indicates the onset of accelerated crystallization.
5:20 PM
Exploring Irradiation-induced Phase Evolution in WC: Charles Hirst1; Diana Shklover1; Paola Amadeo1; Scott Middlemas2; Samuel Humphry-Baker3; Michael Short1; 1Massachusetts Institute of Technology; 2Idaho National Laboratory; 3Imperial College London
Understanding defect production and phase evolution mechanisms in irradiated ceramics is critical for their deployment in both advanced fission and fusion reactors. WC is one such candidate material, proposed for fusion magnet shielding, and as a result must demonstrate its ability to withstand irradiation without degradation. X-ray Diffraction (XRD) of ion-irradiated WC reveals the surprising presence of the high-temperature cubic phase at room temperature. Nanoindentation measures a corresponding decrease in hardness and reduced modulus, and Transient Grating Spectroscopy (TGS) also reports a decrease in thermal diffusivity. These experiments are interpreted to uncover the mechanisms behind the degradation of nuclear ceramics in-service and will inform choices about the design of future fission and fusion reactors.
5:40 PM
Radiation Studies on the TiZrNbHfTa High Entropy Alloy and Its Hydrides: Christopher Moore1; Alberto Fraile1; Caitlin Taylor2; Michael Rushton1; Simon Middleburgh1; 1Bangor University; 2Los Alamos National Laboratory
High entropy alloys have been of growing interest for structural, aerospace, and nuclear applications ever since they were first conceptualised in 2004. A few of the documented properties that highlight them as potential candidates for these industries include high strength, corrosion resistance and potential radiation damage resistance. The TiZrNbHfTa high entropy alloy has been the subject of multiple reports exploring high temperature structural applications, although a surprising lack of previous literature on radiation damage and recombination effects is noted. To account for this, computational methods have been implemented to provide a mechanistic understanding of how this system can be expected to perform during, and how it may recover from, radiation damage events. The results of this analysis will act as a basis on which an experimental irradiation study will be planned, thus working to validate the computational model, and establishing the potential of this system as a nuclear material.