Meeting Materials Challenges for the Future of Fusion Energy: Ceramic & Functional Materials I
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee, TMS: Additive Manufacturing Committee, TMS: Mechanical Behavior of Materials Committee
Program Organizers: Tianyi Chen, Oregon State University; Amy Gandy, United Kingdom Atomic Energy Authority; Reuben Holmes, Kyoto Fusioneering; Ian Mccue, Northwestern University; Sneha Prabha Narra, Carnegie Mellon University; Jason Trelewicz, Stony Brook University; Weicheng Zhong, Oak Ridge National Laboratory
Monday 8:30 AM
March 24, 2025
Room: 158
Location: MGM Grand
Session Chair: Taishi Sugiyama, Kyoto Fusioneering; Kun Wang, Alfred University
8:30 AM Keynote
Transmutations in Advanced Nuclear Ceramics Exposed to Fusion Environments: James Wade-Zhu1; Alex Leide1; Hazel Gardner1; Max Rigby-Bell1; Max Emmanuel1; Douglas Andrews1; 1UK Atomic Energy Authority
Commercial fusion reactors will be one of the most extreme environments for materials, operating at high temperatures under stress while simultaneously exposed to 14 MeV neutrons, corrosive chemicals, and gaseous species. Ceramics will play an essential role in these systems, either as superconducting magnets driving the plasma, neutron shielding protecting the magnets, permeation barriers inhibiting the release of tritium, or structural materials supporting the breeder/coolants in the blanket. During service, these ceramics will undergo transmutations that may affect their structural and functional properties.This presentation will cover R&D activities at UKAEA on advanced ceramics intended for fusion applications. Predictions of the transmutation elements and their quantities in these ceramics will be provided. In addition, details will be given on the various methods being employed to simulate transmutation effects in these advanced materials as well as the test techniques being developed to enable us to compare changes in their performance.
9:00 AM Invited
Functional Coating Development for Fusion Reactors: Takumi Chikada1; 1Shizuoka University
Commercial fusion blankets must ensure high thermal efficiency by high-temperature operation, for instance, using liquid metals or molten salts as tritium breeders. In these blanket concepts, tritium permeation through structural materials, magneto-hydrodynamic pressure drop of liquid metal flow, and corrosion are critical challenges. Functional coating is one of the few promising solutions to mitigate tritium permeation, generation of eddy current in liquid metals and metal components, and corrosion simultaneously. In the last two decades, investigations on functional coatings have progressed experimentally and theoretically with the achievement of relevant properties. The functional coating study proceeds to the next phase: high-performance with highly reliable coating structure. In this presentation, recent achievements and remaining challenges of functional coatings in terms of plant-scale fabrication, hydrogen isotope permeation, liquid metal corrosion, and irradiation effects, are summarized and discussed toward the actual use of the coatings in fusion reactors.
9:30 AM
Tungsten Boride Shielding Material for Fusion Reactors: David Jarvis1; Rosanna van den Blik-Jarvis1; Rosie Mellor1; Max Rigby-Bell2; 1VSCA; 2UKAEA
Nuclear fusion promises to unlock clean energy with near-zero CO2 emissions for future generations. Whilst numerous fusion reactors are presently being designed, developed and built, the successful implementation of this technology relies upon shielding which can protect the underlying hardware from high-energy irradiation. New industrial materials are urgently required that can withstand severe bombardment by neutrons and gamma radiation. VSCA, in collaboration with UKAEA, has developed and tested novel tungsten boride composites that can tolerate realistic radiation doses and ion bombardment. Kilogram blocks of this fully dense shielding material are now being produced at semi-industrial scale by VSCA. The nanostructured tungsten boride composite and its unique material properties will be presented in this paper for the first time.
9:50 AM
Micromechanical Investigation of WC for Shielding Applications in Compact Fusion Devices: Max Chester Jude Emmanuel1; Max Rigby-Bell1; James Wade-Zhu1; 1UK Atomic Energy Authority
Spherical tokamaks offer significant advantages in magnetic confinement, making it easier to achieve fusion by enabling higher plasma pressures. However, the compact nature of these devices means there is limited space for shielding to protect the central column from intense neutron and gamma radiation. Thus, new, high-efficiency neutron absorbing materials are required for these applications. Cemented tungsten carbide (WC) has a high neutron attenuation across broad energy spectrum and is a good candidate for shielding in fusion. However, studies examining the effect of irradiation damage on the material properties of WC is limited. In this work, we have conducted self-ion irradiations on WC-FeCr to doses of 0.13, 1.3 and 13 dpa across an elevated temperature range. Ambient and elevated temperature nanoindentation have been used to measure hardness and modulus alongside single cantilever bending tests to measure fracture toughness. These properties are compared to understand irradiation-induced changes in performance.
10:10 AM Break
10:30 AM Invited
Current R&Ds on Advanced Breeding Functional Materials for JA DEMO Activities: Jae-Hwan Kim1; Taehyun Hwang1; Yutaka Sugimoto1; Suguru Nakano1; Hiroyasu Tanigawa1; 1National Institutes for Quantum Science and Technology
R&Ds on fabrication process of breeding functional materials, tritium breeders and neutron multipliers for JA DEMO applications have been done so far. Through many trials and errors, Li added Li2TiO3 with Li2ZrO3 solid solution (LTZO) as 1 mm-pebble have been successfully fabricated by an emulsion process. Additionally, we optimized the fabrication process to create not only 1 mm-pebbles but also small-sized pebbles with 0.3 mm in diameter. Regarding R&Ds on beryllides, several characterizations of beryllide blocks, including mechanical properties, reactivity to water vapor, swelling properties, hydrogen retention, compatibility with other materials, have been carried out, indicating that beryllide blocks can be adopted. In this study, current R&Ds on advanced breeding functional materials in JA DEMO activities will be given. Additionally, the refinement and recycling process of these materials established by the QST will be addressed by applying the innovative refining process with chemical reaction and microwave heating.
11:00 AM Cancelled
Processing and Irradiation Damage in Novel Tritium Breeding Ceramics with High Lithium Content: David Armstrong1; 1University of Oxford
In this work we will demonstrate the synthesis of high phase purity octolithium compounds, of general formula Li8XO6 where X can be Pb, Zr, Sn, Ge. The complex multi stage reaction is characterised using in situ XRD. This shows significant differences in the four manufacture routes and allows for optimisation of the processing conditions. The materials are then characterised using SEM, EDX and nanoindentation. The mechanical properties measured show excellent agreement with DFT predictions of elastic modulus. Alpha particle irradiation was carried out to simulate neutron damage and transmutation. This has been characterised using Ramen spectroscopy and nanoindentation, showing significant microstructure degradation and embrittlement.
11:20 AM
Innovative Lithium-Based Tritium Breeder Material with Promising Microstructure: Saurabh Sharma1; Chase Taylor2; Dong Zhao1; Kevin Yan1; Jie Lian1; 1Rensselaer Polytechnic Institute; 2Idaho National Laboratory
Multifunctional ceramic breeder materials are highly desirable for deuterium-tritium fusion with high efficiency in breeding tritium through neutron irradiation of lithium-containing blankets. Li2TiO3 displays unique attributes as a potential ceramic breeder material due to its favorable tritium release properties, excellent thermal properties, and low activation characteristics. An optimized microstructure with three-dimensional interconnected pore structure is required for rapid transport of the tritium for effective fuel cycle, which unenviably results in the degradation of their thermal-mechanical properties. Here, Li2TiO3 ceramic pellets with controlled porosities of 14% and 20% are manufactured by volume-controlled spark plasma sintering with an optimized 3D interconnected pore structure, which could be beneficial to facilitate easy removal of transmuted gases. These samples also display simultaneously enhanced thermal-mechanical properties, superior to current state-of-the-art materials. The meticulously designed monophasic Li2TiO3 cylindrical pellets feature an optimized microstructure, porosity, and high thermo-mechanical properties, establishing its potential as a promising tritium breeder material.
11:40 AM
Phase Field Fracture Modeling to Investigate the Integrity of Lithium Aluminate Pellets Used for Tritium Breeding: Kranthi Balusu1; Andrew Casella1; Ayoub Soulami1; 1Pacific Northwest National Lab
Tritium, the principal fuel for future fusion reactors, might rely on lithium aluminate (LiAlO₂) for breeding. Although these ceramic breeding pellets are already deployed in fission reactors, their integrity under more severe conditions must be determined to safely maximize tritium yield. In the reactor, defect-containing pellets undergo extensive microstructural changes. Therefore, a model to predict microstructure-informed fracture is required.In this study, we develop a modeling strategy using the phase field method for fracture within MOOSE software’s finite element framework. The strategy involves three steps: microstructure generation, model parameter determination, and prediction validation. Random microstructures for a representative volume element were based on characterization data of pellets before and after reactor deployment. Numerical model parameters were determined through convergence testing, and material parameters were calibrated using compressive tests on pre-irradiated pellets of different porosities and grain sizes. The simulation capability was validated with compressive strength tests on post-irradiated pellets.