Composite Materials for Nuclear Applications: Composite Fuels/Graphite Carbon
Sponsored by: TMS Structural Materials Division, TMS: Composite Materials Committee, TMS: Nuclear Materials Committee
Program Organizers: Anne Campbell, Oak Ridge National Laboratory; Dong Liu, University of Oxford; Rick Ubic, Boise State University; Lauren Garrison, Commonwealth Fusion Systems; Peng Xu, Idaho National Laboratory; Johann (Hans) Riesch, Max Planck Institute for Plasma Physics

Tuesday 2:00 PM
March 16, 2021
Room: RM 52
Location: TMS2021 Virtual

Session Chair: Anne Campbell, Oak Ridge National Laboratory


2:00 PM  Invited
Improved Techniques for Determining Local Thermal Transport in Composite Nuclear Fuels: Scott Middlemas1; Joshua Kane1; Tsvetoslav Pavlov1; Boopathy Kombaiah1; Daniel LaBrier2; Yu-lin Shen3; Isabella Van Rooyen1; 1Idaho National Laboratory; 2Idaho State University; 3University of New Mexico
    Various composite fuel systems are currently being developed for greater accident tolerance in light water reactors, higher efficiency in advanced gas-cooled reactors, and lower enrichment for high-performance research reactors. While the use of dispersed phase/multilayered fuel forms have several advantages over monolithic forms, they also pose additional challenges in characterizing and modeling reactor performance. Accurate knowledge of thermal conductivity in the various layers/phases of the composite is required for predicting fuel temperature profiles. In this presentation, we will discuss various laser based thermoreflectance techniques that have been employed to determine the mesoscale thermal transport of several different types of composite fuels with micron level spatial resolution. Where possible, the local thermal transport measurements and microstructures of post-irradiated specimens are compared with those of pre-irradiated specimens and correlated to show the degradation of thermal transport due to the evolution of fission gas bubbles and other irradiation induced defects.

2:30 PM  Invited
Overview of the Westinghouse Accident Tolerant and High Burnup Fuel Program: Edward Lahoda1; Zeses Karoutas1; Luke Olson1; Luther Hallman1; Kathryn Metzger1; Jorie Walters1; Michael Sivack1; John Lyons1; Luke Czerniak1; Allan Jaworski1; Ben Maier1; Robert Terry1; Zachary McDaniel1; Frank Boylan1; Jeffrey Kobelak1; Michael Shockling1; Magnus Limback1; Antoine Claisse1; Jonathan Wright1; John Ghergurovich1; 1Westinghouse Electric
    Westinghouse is commercializing a portfolio of four EnCore accident tolerant fuel (ATF) designs; chromium-coated zirconium alloy cladding, SiGA silicon carbide (SiC) cladding, ADOPT fuel and UN fuel. This program has been expanded to include development of high burnup, high energy (HBHE) fuel including advanced Zr alloy cladding (AXIOM cladding) and greater than 5% U235 UO2 and ADOPT fuel. Lead test rods have been inserted into the commercial reactor Byron Unit 2 as part of the Spring 2019 and into Doel-4 in summer of 2020 fuel reloads. Cr coated cladding samples are continuing irradiation at the Massachusetts Institute of Technology Reactor. Out-of-reactor testing on UN fuel has revealed a much better response to a potential leaking fuel rod during operation and has replaced U3Si2 as the advanced fuel of choice. Licensing is proceeding in parallel on ATF and HBHE products incorporating the advanced technologies of in-rod sensors and atomic scale modeling.

3:00 PM  
Development of UN/UO2 Composite Fuels for LWR Applications: Peng Xu1; Lingfeng He1; Brian Jaques2; Kumar Sridharan3; Darryl Butt4; 1Idaho National Laboratory; 2Boise State University; 3University of Wisconsin; 4University of Utah
    UN is a promising fuel form with a higher thermal conductivity and density than conventional UO2 that is widely used as the light water reactor (LWR) fuel. The main drawback of UN is inadequate water corrosion resistance as UN reacts with coolant water at normal operating conditions much faster than UO2, resulting in higher contamination of the coolant and increased dose rate for onsite workers. To mitigate the corrosion issue for UN, a natural step forward is to form a UN composite with a uranium bearing phase that is corrosion resistant. In this study, the UN/UO2 composites of various compositions were fabricated and tested for corrosion resistance. The corrosion behavior and mechanisms will be discussed. The microstructure of the composites was characterized using advanced electron microscopy such as SEM and TEM. Selected composites were also irradiated using 2MeV proton beam, and the radiation induced phase and defect evolutions were characterized.

3:20 PM  
Uranium Nitride Advanced Fuel: An Evaluation of the Oxidation Resistance of Coated and Doped Grains: Yulia Mishchenko1; Denise Adorno Lopes2; Kyle Johnson3; Janne Wallenius1; 1KTH; 2Westinghouse Electric Company; 3Studsvik Nuclear AB
     Uranium nitride is a high-density fissile fuel under consideration as a promising fuel for light water reactor (LWR) systems. However, its resistance to corrosion in contact with steam and air is significantly inferior to UO2 fuel and therefore must be improved. In this study, the oxidation behaviour of the composite UN-AlN, UN-CrN and UN-AlN-CrN pellets was investigated and compared with the pure UN and UO2 pellets. These dopants were selected based on ab-initio modelling calculations.UN powders were mixed with various amounts of AlN and CrN powders and sintered into pellets of high density using the SPS method. The pellets were sectioned into small cubes and were subjected to a thermal transient up to 800°C under inert, air and steam atmospheres in an STA-EGA (TGA-DSC-Gas-MS) system. The work performed here are intended to provide the basis for future improvement in water tolerance of advanced UN fuels for LWR applications.

3:40 PM  
Fabrication, Characterisation and Oxidation Resistance of an Innovative Composite Fuel: UN Microspheres Embedded in UO2 Matrix: Diogo Costa1; Marcus Hedberg2; Simon Middleburgh3; Janne Wallenius4; Pär Olsson4; Denise Lopes5; 1KTH Royal Institute of Technology, Westinghouse Electric Sweden AB; 2Chalmers University of Technology; 3Bangor University; 4KTH Royal Institute of Technology; 5Westinghouse Electric Sweden AB
    UN-UO2 composite fuels are being considered as an accident tolerant fuel option for LWR. Mixtures of UN microspheres (10-50wt%) and UO2.13 powder were sintered by spark plasma sintering and characterised by XRD, DSC, and SEM-EDS/WDS/EBSD. The UN and UO2 interaction is driven by O diffusion into the UN phase and N diffusion in the opposite direction, forming a long-range solid solution (UO2-xNx) in the UO2 matrix. On cooling, the N solubility in UO2-xNx decreases and causes N redistribution and precipitation as α-U2N3 phase along and inside the UO2 grains. Additionally, TG/DSC investigations in air (30-700°C) show that the UO2 increases the oxidation onset temperature (OOT) from 260(3)°C (pure UN pellet) to 320(4)°C (10wt% UN-UO2 composite fuel). Composites with 30wt% and 50wt% UN present similar OOT (~285°C), which is similar to pure UO2 pellet (~300°C). Furthermore, the weight variations and the maximum reaction temperatures are reported and discussed.

4:00 PM  Invited
Use of Carbon Fibre-reinforced Carbon in Wendelstein 7-X: Jean Boscary1; Henri Greuner1; Boris Mendelevitch1; Gunnar Ehrke1; Patrick Junghanns1; Reinhold Stadler1; 1Max-Planck-Institut für Plasmaphysik
     Wendelstein 7-X (W7-X) is an optimized stellarator which started operation in 2015 at Greifswald, Germany. The objective is to demonstrate steady state operation of fusion-relevant hydrogen and deuterium plasmas. The divertor function is to control the plasma power and particle exhaust and is subject to heat loads by convection or thermal radiation from the plasma. The 20m² W7-X divertor is made of 890 target elements (TEs) designed to remove a stationary heat flux of 10 MW/m² and cooled with pressurized water. The TEs are made of the Cu alloy CuCrZr heat sink armored with about 16,000 carbon fibre reinforced carbon (CFC) NB31 flat tiles as plasma facing material. The developed joining technology between tiles and heat sink is a Cu bi-layer technology. The manufacturing route of CFC and target elements will be presented.

4:30 PM  
Sub-critical Crack Initiation, Coalescence and Propagation in Nuclear Graphite Studied by High-speed Pink Beam Synchrotron Tomography: Thomas Zillhardt1; Dong Liu2; James Marrow1; 1University of Oxford; 2University of Bristol
    The United Kingdom has fourteen active Advanced Gas-Cooled moderated by Gilsocarbon graphite, a polygranular quasi-brittle that also acts as a neutron reflector and as a structural component. The development of keyway root cracking in graphite is a potential limiting factor for the lifetime of the AGRs. Gilsocarbon graphite is a heterogeneous material that contains many defects at different length scales which are distributed in both the matrix and the filler particles. To understand how keyway root cracking may be affected by microstructural differences, we have studied how damage occurs and is accommodated at the microstructural level, prior to fracture, and observed the initiation, nucleation, coalescence and propagation of sub-critical cracks, and we have also made observations of microstructural damage leading to brittle fracture. This has been made possible with the use of advanced characterization techniques at the PSICHE beamline, where we have carried out in-situ high-speed continuous X-Ray tomography.