Composite Materials for Nuclear Applications II: Graphite/Carbon Composites
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee, TMS: Composite Materials Committee, TMS: Mechanical Behavior of Materials Committee, TMS: Advanced Characterization, Testing, and Simulation 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 Riesch, Max-Planck-Insitut Fuer Plasmaphysik
Tuesday 2:30 PM
March 21, 2023
Room: 24B
Location: SDCC
Session Chair: Dong Liu, University of Bristol; Rick Ubic, Boise State University
2:30 PM Invited
Nuclear Graphite as a Core Composite Material: Willliam Windes1; 1Idaho National Laboratory
Commercial grade graphite used for core components in nearly all High Temperature Reactor (HTR) graphite-core designs is composed from a complex composite mixture of carbonized and anisotropic graphene-like phases, with a healthy amount of porosity thrown into the microstructure. Yet the (bulk) material properties display a near isotropic response even though graphite is composed from graphene-like structures which respond anisotropically. On one hand, the atomic length-scale irradiation response is nearly completely anisotropic with atomic displacement imposing shrinkage in the a-axis crystallographic direction and expansion within the c-axis direction. On the other hand, the irradiation macroscopic (bulk) response is near-isotropic. This discussion will review this unique irradiation material behavior, discuss crystallographic versus macroscale response, and speculate on how the microstructure may be influencing the overall isotropic behavior. New information and the latest thoughts on how this complex mixture of disparate phases can account for the odd behavior will be discussed.
3:00 PM
Oxidation Effects on the Microstructure of Nuclear Graphite: David Arregui-Mena1; Phillip Edmondson1; James Spicer2; Cristian Contescu1; Paul Mummery3; Lee Margetts3; Nidia Gallego1; 1Oak Ridge National Laboratory; 2Johns Hopkins University; 3The University of Manchester
Large amounts of graphite are required to form the core components of Very High Temperature Reactors (VHTR) and Molten Salt Reactors (MSR) that are being planned. These components must withstand high and strong temperature gradients, neutron irradiation, and oxidation effects. Limited research has been conducted on the effects of oxidation on the microstructure of nuclear graphite especially for modern graphite grades. This research focuses on oxidation experiments combined with x-ray computed tomography to characterize the effects of oxidation on the microstructures of different types of graphite. This type of analysis helps determine the extent of damage generated as a result of accidental ingress of air into a reactor core or by chronic oxidation. The microstructural characterization aspects of this research were complemented with image-based models (IBMs) and random field theory to predict the effects of oxidation in nuclear graphite on its elastic properties.
3:20 PM
Ruthenium and Silver Diffusion in Nuclear Graphite: Dina Elgewaily1; Jacob Eapen1; 1North Carolina State University
Nuclear graphite is used for neutron moderation as well as a fission product (FP) barrier in high-temperature gas-cooled reactors. A firm knowledge of FP diffusion mechanisms in graphite under normal and accident operational conditions is therefore important for reactor design. In this work, the diffusional characteristics of ruthenium and silver are investigated for five graphitic grades – AXF-5Q, ZXF-5Q, IG-110, NBG-18, and PCEA – for temperatures ranging from 500 to 900°C. Interestingly, two diffusional regimes, a fast region, and a slow region, can be identified. Surface characterization using electron microscopy reveals a transition from a spatially uniform surface coverage to a clustering behavior; this transition occurs between 700 to 800°C, approximately, for all the graphite grades. Concurrent Raman spectroscopy further corroborates this transitional behavior.
3:40 PM Invited
Role and Structure of HTGR Matrix Material: Tyler Gerczak1; Anne Campbell1; Grant Helmreich1; William Cureton1; Elizabeth Sooby2; Ryan Latta3; Gerald Jellison1; John Hunn1; 1Oak Ridge National Laboratory; 2University of Texas San Antonio; 3Kairos Power
Typical fuel forms for high temperature gas-cooled reactors consist of tristructural isotopic (TRISO) coated particles dispersed within a composite matrix of graphite flake and partially carbonized resin binder. In general, the final fuel form is either a right cylindrical compact or spherical pebble. The matrix material provides multiple functions such as neutron moderation, control of inter-particle spacing, structural support, and mitigation of fission product release. The matrix has generally received less attention relative to the TRISO fuel particle. Here, a discussion on the matrix structure and its relationship to the overcoating and compacting process will be presented. Additional insight on the influence of microstructure on the matrix’s thermophysical properties and their impact on fuel performance during normal and transient operating conditions will be explored.
4:10 PM Break
4:30 PM
Irradiation Effects in the Composite Phases of Graphite and Carbon-Based Materials: Anne Campbell1; Jose Arregui-Mena1; 1Oak Ridge National Laboratory
Graphite and carbon-based materials have been long used in the nuclear industry, from the first man-made critical pile to Gen-IV fission reactors and in fusion systems. Graphite inherently is a composite material because it contains three distinct phases: filler, binder, and pores, while carbon-composites contain similar phases: carbon fiber, binder/matrix, etc., making these arguably the most widely used composite nuclear material. Understanding the behaviors of these different phases, when exposed to the extreme environments in fission and fusion systems, including different forms of energetic particles and ionizing radiation and a range of temperatures, is critical to understand and predict the lifetime of components made from these materials. This talk will focus on what the current knowledge state about irradiation effects in the different phases present in graphite and carbon composites, where the knowledge is lacking, and future work that would be required to fill in the gaps.
4:50 PM
ENHANCED Shield: A Critical Materials Technology Enabling Compact Superconducting Tokamaks: David Sprouster1; B Cheng1; J Trelewicz1; G Khose2; E Peterson2; S Zinkle3; Lance Snead1; 1Stony Brook University; 2Massachusetts Institute of Technology; 3University of Tennessee Knoxville
With significant improvement in High Temperature Superconductors (HTS), a number of projects are adopting HTS for power systems. Compact HTS tokamaks offer advantages including lower plant costs and lower cost of electricity. As compact reactors have less space for shielding, HTS degradation is potentially design limiting. Shielding must mitigate threats to the superconducting coils. Unfortunately, there are currently no hi-performance shielding materials to enable the potential performance enhancement offered by HTS. We present an advanced manufacturing method to fabricate a new class of shields that are high-performance, high-operating temperature, and simultaneously neutron absorbing. The designs consist of an entrained metal-hydride phase within a radiation stable ceramic host. We present fabrication, characterization, and thermophysical data for a series of down selected composites inspired by future fusion designs and performance metrics. This work was performed under the auspices of the ARPA-E Galvanizing Advances in Market-Aligned Fusion for an Overabundance of Watts program.