IC-5 - 5th International Conference
Materials Challenges for Sustainable Nuclear Fission and Fusion Technologies
ABSTRACTS
Session IC-5.A Structural materials for nuclear fission and fusion application
IC-5.A:IL01 Radiation Stability of Carbide Ceramics: The Roles of Short-Range Order and Interfaces
I. SZLUFARSKA, S. WEI, W.M. QURESHI, University of Wisconsin-Madison, Madison, WI, USA
Future nuclear energy systems require materials that can withstand extreme conditions of irradiation, temperature, and chemical reactivity. Complex carbide ceramics are promising candidates due to their exceptional thermal stability, corrosion resistance, and strength, yet the mechanisms governing their radiation stability remain poorly understood. In this talk, I will highlight recent advances in uncovering how atomic-scale structure, particularly chemical short-range order (CSRO) and interfaces, controls the radiation response of high-entropy carbides (HECs) and SiC. We find that grain boundaries in these materials are inherently off-stoichiometric and that their complex energy landscape drives radiation-induced segregation (RIS). Surprisingly, RIS can improve corrosion resistance in high-temperature aqueous environments. In HECs, we demonstrate for the first time the presence of CSRO and its impact on radiation-induced swelling. Complementary first-principles calculations provide new predictive tools for assessing phase stability and uncover diffusion mechanisms fundamentally distinct from those in high-entropy alloys. Together, these findings offer new scientific insights and practical pathways for designing radiation-tolerant carbides for future nuclear energy systems.
IC-5.A:IL02 Highly Recyclable Low-activation Vanadium Alloys for Fusion Reactors
T. NAGASAKA, National Institute for Fusion Science, Toki, Japan; A. NOMURA, T. SUGAWARA, Institute for Materials Research, Tohoku University, Sendai, Japan; S. SAKURAI, Taiyo Koko Co., Ako, Japan; K.-I. FUKUMOTO, Research Institute of Nuclear Engineering, University of Fukui, Tsuruga, Japan; H. WATANABE, International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University, Oarai, Japan; Y. YAMAUCHI, Faculty of Engineering, Hokkaido University, Sapporo, Japan; K. KATAYAMA, Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, Kasuga, Japan; V. TSISAR, Belgian Nuclear Research Centre SCK•CEN, Boeretang, Belgium
Low-activation vanadium alloys, such as V-4Cr-4Ti, are an alternative to ferritic steels for fusion reactors because of their high-temperature strength and good compatibility with liquid metal lithium, which serves as both a coolant and a tritium breeder material. Their low-activation characteristics are also advantageous, as they allow for earlier materials recycling than steels. Since V and Cr do not create long-lived radioactive isotopes emitting high-energy gamma rays, the cooling period before material recycling is controlled by the levels of Ti and harmful, high-activation impurities, such as Co, Cu, Fe, Nb, Ni, and Mo. Based on recent results, a recycling scenario with a cooling period of only 10 to 20 years would be possible with a minimized Ti concentration and appropriate impurity controls. In this scenario, the vanadium alloys could be reused in the same fusion reactor without any transportation or off-site intermediate storage of radioactive materials, because the 10-20 year cooling is short enough compared to the expected 30-40 year lifespan of fusion reactors. This paper reviews recent progress in the development of V-4Cr-4T and such low-Ti high-purity alloys, including the addition of further Cr and other minor elements to compensate for the loss of Ti.
IC-5.A:IL03 Achieving High Tensile Strength and Ductility in Refractory Alloys by Data-driven Design
H. HUANG, P. SINGH, D.D. JOHNSON, G. OUYANG, L. GAYDOS, T. RIEDEMANN, R.T. OTT, N. ARGIBAY, Ames National Laboratory, IA, USA; D. BENIWAL, P.K. RAY, Indian Institute of Technology Ropar, Rupnagar, PB, India; T. INGALE, V. SONI, R. BANERJEE, T. SCHARF, University of North Texas, Denton, TX, USA; P. LU, F.W. DELRIO, A. KUSTAS, Sandia National Laboratories, Albuquerque, NM, USA; J. SHARON, R. DEACON, RTX Technology Research Center, East Hartford, CT, USA; S.I.A. JALALI, M. PATULLO, S. PARK, K. HEMKER, Johns Hopkins University, Baltimore, MD, USA
Energy efficiency of heat engines (gas and steam turbines) for electricity production and propulsion scale with operating temperature following the Carnot cycle. Commercial Ni- and Co-based superalloys melt near 1500 °C and rapidly lose mechanical strength beyond 1000 °C. Refractory metals melt well above 2000 °C but have inherent manufacturability challenges, like high ductile-to-brittle transition temperatures, that are significant barriers to adoption. Using density-functional theory guided design, we demonstrated tailored local lattice distortions that promote phase-stable, non-equiatomic refractory concentrated solid-solutions with both high ductility and strength. This behavior was exemplified for single-phase, body-centered cubic Nb4Ta4V3Ti, which exhibited castability, excellent room-temperature tensile yield strength (≈1 GPa) and ductility (approaching 20% uniform strain), and exceptional high-temperature tensile strength (500 MPa at 1000 °C). The findings illustrate a path forward for designing materials that hold great potential for next-generation technologies such as Gen-IV fission reactors, first-generation fusion-plasma reactors, and more efficient gas turbines for electricity generation and propulsion.
IC-5.A:IL04 Effect of Helium in Neutron Irradiated RAFM Steels
A. BHATTACHARYA, School of Metallurgy & Materials, University of Birmingham, Edgbaston, UK
Reduced activation ferritic-martensitic (RAFM) steels are the most promising candidates for water cooled, and PbLi fusion first-wall/blanket systems. A major technical gap in understanding fusion-specific radiation damage in RAFM steels is the high helium (He) generation rate, ~10 atomic parts per million (appm) He/dpa, under the fusion neutron spectra. But in absence of a high flux 14 MeV neutron facility, the effect of He on mechanical properties degradation is challenging to estimate with test reactors due to a lack of He production in RAFM steels in a fission neutron spectra. Using new results and from literature, the overarching concern about the insufficient understanding of the effect of He on low-temperature hardening-embrittlement (LTHE), cavity swelling and high temperature He embrittlement (HTHE) will be presented. The presentation will collate results from spallation neutron experiments, neutron irradiated isotopically tailored RAFM steels (>85 dpa, >900 appm He) and ion irradiation to showcase the current understanding of the effect of He. Focus will be given to comparing spallation versus isotopically tailored RAFM steels neutron data where spallation sources significantly over-predict the effect of He on tensile properties. Microstructural understanding of the phenomenon will also be discussed.
IC-5.A:IL05 Reducing the Waste Burden of Nuclear Graphite
L. SNEAD, Stony Brook University and MIT, New York, NY, USA
Waste burden of irradiated graphite (i-graphite), is currently >250,000 tons worldwide, having been the moderator-of-choice for many research and power reactors. Currently, many advanced reactors assume graphite use. In the absence of significant cross-contamination from fuel, 14C, 36Cl, 3H, and 60Co are isotopes driving waste management approach for i-graphite. As 3H and 60Co will substantially decay within decades after shutdown, they are of near-term concern for operation/maintenance, shipping and storage. However, isotopes of 14C and 36Cl drive the long-term disposal management schemes. Development of economically attractive and environmentally sound i-graphite management strategies are paramount to limit environmental impact of waste i-graphite waste and ultimate acceptability of graphite reactors. In this presentation we discuss sources, levels, and efforts to mitigate the issues of 14C and 36Cl in i-graphite. As example, as 14C primarily results from neutron capture of native 13C and neutron capture of 13N which can be present in as-processed graphite, or surface adsorbed, we will present a survey of nitrogen levels of current and historic graphite materials, parsing the sources of the 13N and discussing the implications of fabricating lower N graphite materials.
IC-5.A:L07 Effect of High Temperature He Ion Irradiation on Corrosion Properties of Nickel-based Alloys in Molten Salt Environment
GUANHONG LEI, CHENG LI, HEFEI HUANG, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai, China
GH3535 alloy is the structural material for the Molten Salt Reactor (MSR). It will be subject to the extreme environments, i.e. high temperature, high neutron doses and corrosive coolant during the operation of MSR. Much effort has been made to explore the synergetic effect of irradiation and corrosion on material performance degradation. Helium bubble is a typical irradiation defect that formed due to the transmutation reaction of neutron with boron and nickel in GH3535 alloys in reactor. In molten salt environment, helium bubbles enhanced corrosion of nickel-based alloy was observed. In addition, the growth and coalescence of bubbles may accelerate the formation of cavities and even holes with increasing salt exposure time, which will lead to less ductility and easier rupture of structural alloys, jeopardizing their applications in reactors. Hence, understanding the governing factors of helium bubbles evolution in molten salt environment is crucial for the reliability issues of nuclear reactors.
Session IC-5.B Materials for first wall components of nuclear fusion systems
IC-5.B:IL08 Synergistic High-heat Flux and Particle Loads on Plasma-facing Components
M. WIRTZ, M. GAGO, D. DOROW-GERSPACH, A. KRETER, G. PINTSUK, J. TWEER, B. UNTERBERG, C. LINSMEIER, Forschungszentrum Jülich, Institute of Fusion Energy & Nuclear Waste Management (IFN-1), Jülich, Germany
The plasma facing wall of future thermonuclear fusion reactors with magnetic confinement such as ITER or DEMO has to withstand harsh loading scenarios. These processes are associated with quasi-stationary thermal loads up to 20 MW/m2 combined with extremely strong thermal transients up to the GW/m2-range during Edge Localized Modes (ELMs). In addition, irradiation effects resulting from the plasma species and the 14 MeV fusion neutrons have strong impact on the integrity of the wall armour materials. Therefore, synergistic effects resulting from simultaneous thermal and plasma wall loads are evaluated in complex experiments. Under reactor-relevant conditions, thermally induced defects such as cracking and melting of the plasma facing material (PFM), thermal fatigue damage of the joints between the plasma facing material (PFM) and the heat sink of a plasma facing component (PFC), hydrogen-induced blistering and helium-generated formation of nano-sized tendrils at the plasma facing surface leading to a reduction of the thermal conductivity and embrittlement are the most serious damaging mechanisms. Today tungsten is considered as the most promising and reliable material for high heat flux components in future fusion reactors.
IC-5.B:IL09 The Impact of Boron in Tungsten Plasma Facing Walls: Sputter Yields, Near-surface Morphology, and Fuel Retention
E. PITTHAN1, D. GAUTAM1, T.T. TRAN1, M. FELLINGER2, B. BURAZOR DOMAZET2, R. GURSCHL2, J. BRÖTZNER2, R.A. WILHELM2, F. AUMAYR2, H. RIEDL3, A. CLEMENT4, N.F. MOFRAD4, A. SAND4, M. RUBEL1,5, D. PRIMETZHOFER1, 1Uppsala University, Uppsala, Sweden; 2TU Wien, Institute of Applied Physics, Fusion@ÖAW, Vienna, Austria; 3TU Wien, Institute of Materials Science and Technology, Vienna, Austria; 4Department of Applied Physics, Aalto University, Aalto, Espoo, Finland; 5KTH Royal Institute of Technology, Stockholm, Sweden
Tungsten (W) is the main candidate for plasma-facing materials in tokamaks, as it offers, among other properties, low sputter yield and low retention of hydrogen isotopes. However, W does not provide sufficient mid-Z gettering properties, which are necessary to reduce the presence of impurities that degrade the plasma. ITER plans to use boronization as a wall conditioning process that coats the walls with boron (B) to reduce the partial pressure of oxygen and water. This procedure may lead to formation of W–B compounds through redeposition. Such modifications can significantly affect sputtering and fuel retention. This talk will present recent efforts to understand the behavior of these materials and their consequences in fusion devices. W-B layers of multiple compositions were prepared, mimicking expected layers. Experiments investigated sputter yields, surface morphology evolution (surface enrichment, crack formation) after ion irradiation, and hydrogen isotope retention at different temperatures. Finally, detailed atomic-scale modeling, including sputtering simulations based on molecular dynamics as a function of surface composition, were conducted and show good agreement with experimental trends.
IC-5.B:IL10 Development of SiCf/SiC Composites Blanket
T. SUGIYAMA1, D. YAGI1, S. OGAWA1, P. BARRON1, R. HOLMES1, Y. KANEKO1, JUNYEAB LEE2, F. SHINODA2, T. HINOKI2, 1Kyoto Fusioneering Ltd., Kyoto, Japan; 2Kyoto University, Kyoto, Japan
SiCf/SiC composites are candidate materials for advanced fusion concepts due to their strong mechanical performance after high temperature irradiation, corrosion resistance, hydrogen permeation resistance, adequate thermal properties, and sputtering resistance. They are being considered as materials for the first wall, blanket, piping, flanges, and heat exchanger components and systems. Progress and results from Kyoto Fusioneering’s materials research will be presented, covering both the development of our fusion-grade SiCf/SiC and our work using it to produce a prototype blanket module. This module, which utilises liquid lithium-lead as coolant, breeder and neutron multiplier, aims to achieve simplicity and high conversion efficiency, as informed by KF’s blanket design team. Liquid phase sintering is the process for manufacturing KF’s SiCf/SiC, and various shapes manufactured using liquid phase sintering will be presented. Recent developments in using SiCf/SiC for components other than the blanket are extensive. These advances will be discussed in the context of their potential to advance fusion plant design, as well as how KF’s “UNITY” experimental facilities are well-placed to test these new components.
IC-5.B:IL11 Nuclear Aspects of Plasma-wall Interaction Processes
C. LINSMEIER1, R. RAYAPROLU1, J.W. COENEN1, A. LAU1, Y. MAO1, A. LITNOVSKY1, B. UNTERBERG1, M. WIRTZ1, G. PINTSUK1, J. RIESCH2, 1Forschungszentrum Jülich GmbH, Institute of Fusion Energy and Nuclear Waste Management – Plasma Physics, Jülich, Germany; 2Max-Planck-Institut für Plasmaphysik, München, Germany
Neutrons from the D-T fusion reaction impinge in a fusion reactor at the plasma-facing materials. Due to collisions, defects are created in the material, described by “displacements per atom”. Transmutation reactions induced by the neutrons lead to change in the chemical composition of the materials by creating new elements. Both effects lead to severe changes of the physical and mechanical properties of first wall materials. In this contribution, neutron effects are described, and their consequences will be discussed. Furthermore, material solutions are presented which allow to compensate for the detrimental effects of neutrons. In particular, solutions against embrittlement of tungsten materials will be demonstrated, improving the application of tungsten materials in a fusion environment. Finally, in this presentation requirements for testing of plasma-facing materials including neutron effects will be presented. This includes irradiation concepts in lieu of a fusion reactor as source for 14 MeV neutrons.
IC-5.B:IL12 Perspectives and Challenges of Tungsten and Tungsten Alloys for Fusion Plasma-facing Applications
XUNXIANG HU, Sichuan University, Chengdu, China; L. SNEAD, Stony Brook University USA; B.D. WIRTH, University of Tennessee, Knoxville, USA; Y. KATOH, Oak Ridge National Laboratory, USA
Tungsten (W) and its alloys are among the most promising candidates for plasma-facing materials (PFMs) in future fusion reactors due to their exceptional melting point, thermal conductivity, and sputtering resistance. However, their deployment remains challenged by intrinsic brittleness, radiation-induced degradation, and complex interactions with the fusion plasma environment. This presentation provides a comprehensive overview of the current state and future directions of W-based materials for fusion applications. We first examine the fundamental thermo-mechanical properties of tungsten and its alloys, emphasizing microstructure engineering strategies that link structural features to performance optimization. We then summarize recent advances in understanding the effects of neutron irradiation, hydrogen isotope permeation and retention, and other fusion-relevant conditions. Remaining challenges, including the synergistic impact of these factors and the need for enhanced radiation tolerance, are discussed. Finally, we outline key research and development priorities required to enable the reliable deployment of tungsten-based PFMs in next-generation fusion devices.
IC-5.B:L14 Healing of Microcracked Tungsten by Electroplating and Hot Isostatic Pressing
H. NOTO, J. SHEN, S. TAKAYAMA, H. UEHARA, T. MUROGA, National Institute for Fusion Science, Toki, Japan; H. WANG, Y. NORIKAWA, T. NOHIRA, Institute of Advanced Energy, Kyoto University, Gokasho, Japan
Plasma-facing materials in fusion reactors are subjected to high heat and particle loads, resulting in various types of damage. Mechanical damage begins as microcracks and eventually progresses to extended cracks, necessitating replacement and disposal. Healing the damage at a minor stage for continuous use would reduce the amount of radioactive waste. This study aims to determine the feasibility of healing microcracks induced by pulsed heat loads in tungsten through hot isostatic pressing (HIP). In particular, this attempt to heal open cracks by pre-forming a tungsten plating layer on the surface is unprecedented. ITER-grade tungsten irradiated with an Nd:YAG laser at an intensity of ~100 J/cm2 for 8 ns produced ablation craters, which contained a high-density microcrack on the surface. By HIP-bonding this sample to a copper sheet at 1000°C and 200 MPa, it was confirmed that the width of the microcracks was reduced significantly. Regarding tungsten plating, a molten CsF-CsCl-WO3 system was used, and it was found that an extremely smooth and dense β-W plating film could be obtained on a Cu substrate at 550°C. The results of an experiment integrating these technologies to heal microcracks in tungsten through electroplating and HIP at temperatures exceeding 1000°C will be reported
IC-5.B:L16 Development of Silicon Carbide Fiber Reinforced Tungsten and Molybdenum Matrix Composites for Nuclear Fusion
TATSUYA HINOKI, T. KUZUYA, Y. DU, Kyoto University, Kyoto, Japan
Tungsten (W) is a primary candidate material for divertor and first wall. However, mechanical properties of W degrade at high temperature due to recrystallization above 1000 °C. Neutron irradiation also affects mechanical properties significantly. Silicon carbide (SiC) has very close coefficient of thermal expansion with W. SiC fibers can be used at above 1000 °C under neutron irradiation without significant degradation of mechanical properties. Molybdenum (Mo) is also a high-Z material and one of candidates of plasma facing materials. The objective of this work is to develop the W and the Mo materials with ductile behavior under high temperature neutron irradiation environment by reinforcement of SiC fibers. Hi-Nicalon type-S SiC fibers, W and Mo powders and thin foils were sintered together at 1500~1900 ºC with 20 MPa pressure by hot press. Mechanical properties were evaluated by tensile tests. Microstructure was evaluated by SEM and EPMA. Thermal diffusivity was measured by light flash apparatus. Heat capacity was measured by DSC. The composites were fabricated at above recrystallization temperature of W and Mo. The composites showed pseudo-ductile fracture behavior at ambient temperature.
This work was supported by JSPS KAKENHI Grant Number JP23K20837.
IC-5.B:L17 Mechanical Properties of a Pure Tungsten – Before and After Recrystallization
CHANGHEUI JANG, T. AN, L. YI, Korea Advanced Institute of Science and Technology, Daejeon, Rep. of Korea; B.S. KONG, Korea Institute of Nuclear Safety, Daejeon, Rep. of Korea; H.C. KIM, Korea Institute of Fusion Energy, Daejeon, Rep. of Korea
Tungsten is considered as a candidate materials for plasma-facing components of fusion reactors. However, when exposed to high thermal loadings during operation, cold-worked tungsten would be recrystallized. Therefore, the effect of recrystallization on the mechanical properties of a pure tungsten should be evaluated. We conducted fracture and fatigue crack growth tests for cold-worked and recrystallized tungsten using a small-sized disk-type compact tension specimen with a sharp pre-crack. The fatigue pre-cracking was performed in high temperature argon gas environment to overcome inherent brittleness and poor oxidation resistance,. Fracture toughness tests were performed in static and quasi-dynamic conditions, in air and argon environments at up to 700°C. Fatigue crack growth tests were performed at 500°C to 700°C in argon environment. Fracture and fatigue properties were significantly degraded after recrystallization. The relevant deformation mechanism depending on test temperature and recrystallization were discussed in view of fracture and fatigue resistance and microstructure.
IC-5.B:IL18 Designing Microstructurally Tailored Tungsten Alloys with Recrystallization Resistance for Fusion Divertors
J.R. TRELEWICZ, Stony Brook University, Oak Ridge National Laboratory, USA
Of the many materials challenges for fusion reactors, the susceptibility of tungsten (W) to recrystallization continues to plague the development of stable divertor armor. Additionally, the need for geometrically complex components has driven interest in laser additively manufactured (AM) W, which suffers from solidification cracking. In this presentation, two computationally designed plasma facing material (PFM) technologies are introduced for addressing the intrinsic limitations of W – grain boundary stabilized ultrafine grained (UFG) W-Ti-Cr alloys and laser AM W-Ti-Fe alloys. In the first system, compositional complexities that stabilize fine-grained microstructures are identified from alloy design maps informed by computational thermodynamics. A new series of ternary UFG W alloys are synthesized with thermal stability demonstrated above common recrystallization temperatures for W. Second, a CALPHAD alloy design strategy to fabricate laser AM W alloys free of cracking and with improved fracture properties is presented. Microstructures are contrasted with pure AM W and discussed in the context of the alloying additions and processing-related defects. Collectively, we show that these materials provide a basis for enhancing PFM performance relative to pure W.
IC-5.B:IL19 Materials Research for Plasma-facing Materials and Components at SWIP
JUAN DU, Southwestern Institute of Physics, Chengdu, China
Nuclear fusion energy is the ultimate solution to humanity's energy crisis. Achieving its practical engineering application at an early stage is crucial for ensuring national security and economic development. From an engineering perspective, one of the most critical challenges in realizing nuclear fusion energy lies in the development of fusion materials. Plasma-facing materials are among the most challenging materials required for a fusion reactor. Southwestern Institute of Physics (SWIP) was China’s first nuclear fusion research institute and has been an important contributor to China's participation in the ITER project. For many years, SWIP has focused on the development of plasma-facing materials, including advanced tungsten alloys produced via industrial processing routes, W-Y₂O₃ and W-K doped tungsten materials with excellent low-temperature ductility and high recrystallization temperature, as well as fully dense tungsten fiber-reinforced tungsten composites exhibiting room-temperature ductility and superior thermal fatigue resistance. SWIP has also been actively involved in developing plasma-facing components, such as a semi-prototype of the ITER enhanced first wall, as well as the advanced divertor for the China Fusion Engineering Test Reactor (CFETR). This presentation will provide an overview of recent progress in plasma-facing materials and components at SWIP, along with the research strategy for the next-step fusion reactor at the institute.
Session IC-5.C Functional materials
IC-5.C:IL20 Overview on First Mirror Testing for Plasma Diagnosis and Plasma Heating in Fusion Reactors
M. RUBEL1,2, P. PETERSSON2, L. DITTRICH2, E. PITTHAN1, A. WIDDOWSON3, E. FORTUNA-ZALESNA4, 1Uppsala University, Uppsala, Sweden; 2KTH Royal Institute of Technology, Stockholm, Sweden; 3Culham Science Centre for Fusion Energy, Abingdon, UK; 4Warsaw University of Technology, Warsaw, Poland
Optical systems for plasma diagnosis in a D-T operated fusion reactor will be based on metallic so-called first mirrors. Tests in current fusion devices (JET) show that co-deposition processes are responsible for degradation of optical performance, especially of mirrors facing the divertor plasma. Main chamber mirrors are modified by co-implantation of neutral species. In a D-T reactor factors associated with boron and tungsten deposition, implantation of 4He and charge exchange hydrogen isotopes (H,D,T), and neutron-induced effects are to be considered. Their impact must be studied in under laboratory conditions also involving irradiations with a range of accelerators. Molybdenum poly- and mono-crystalline mirrors were irradiated with 30 keV 98Mo+ ions to simulate neutron impact, dose up to 4.5·1015 cm-2. 1H and 4He irradiation at 2 keV simulates gaseous transmutation products and the impact of plasma species. The irradiation of mirrors with thin B and W deposits was also done. The irradiation decreased specular reflectivity. The effects caused by 1H and 4He are stronger than those by 98Mo. The changes are small compared to those caused by thin B and W deposits. Low amounts of 4He were retained: 7% of the dose. Self-irradiation with 98Mo and with 4He lead to blister formation.
IC-5.C:IL21 Compound Stoichiometry Effects in Tungsten Boride Radiation Shielding
S. HUMPHRY-BAKER, T. ZAGYVA, Imperial College London, London, UK
The tungsten borides are highly effective radiation shielding materials for high energy neutrons. However they are particularly susceptible to helium damage when boron 10 isotope captures a neutron; and to point defect swelling at low temperatures. We have irradiated various tungsten boride compounds (W2B, WB, and WB2) with 2 MeV helium and 15 MeV tungsten ions in the range 25-700°C. Samples were characterised with electron back-scatter diffraction and grazing-incidence X-ray diffraction. For the WB compound, the defect storage and lattice swelling was generally lower than the other compounds. It was also more resistant to amorphisation: e.g., its critical amorphisation temperature ~25°C for tungsten self-ions, vs. ~300°C for WB2. Our study suggests that although WB has a moderate radiation attenuation performance it may show optimal neutron irradiation resistance.
IC-5.C:IL22 Accident-tolerant Hybrid Ceramics for Fusion Breeding Blanket
KEISUKE MUKAI1,2, TAEHYUN HWANG3, DAIGO KANAMORI2,3, MINORU KUSABA1,4, JAE-HWAN KIM3, 1National Institute for Fusion Science, Toki, Gifu, Japan; 2The Graduate University for Advanced Studies, Sokendai, Toki, Japan; 3Fusion Energy Directorate, National Institutes for Quantum Science and Technology, Obuchi, Rokkasho, Aomori, Japan; 4The Institute of Statistical Mathematics, Research Organization of Information and Systems, Tachikawa, Tokyo, Japan
Hydrogen generation by high temperature reactions between neutron multiplier (i.e. metallic Be-compounds) and steam during an in-box loss-of-coolant accident (LOCA) is one of the severe accident scenarios in a water-cooled ceramic breeding blanket. To allow inherent safety against the in-box LOCA, this study focuses on synthesis and characterization of Li-Be hybrid ceramics. The hybrid ceramic system was designed using machine-learning-based crystal structure prediction (CSPML). Experimental synthesis via solid-state reaction confirmed the formation of Li2BeSiO4 with BeO, validated by X-ray diffraction and Rietveld refinement. Steam exposure tests at up to 1200°C demonstrated negligible hydrogen generation from the hybrid ceramics sample, contrasting sharply with metallic Be which produced H₂ under steam above 400 °C. In our presentation, thermal conductivities of Li2BeSiO4 with Li depletion at elevated temperatures will be also reported. The hybrid ceramics is the first example of multi-functional oxide to breed sufficient fuel tritium with no metallic neutron multiplier, which allows a novel design of ceramic breeding blanket with enhanced safety margins during in-box LOCA.
IC-5.C:IL23 Permeation Barrier Layers for fusion Components
TAKUMI CHIKADA, Shizuoka University, Shizuoka, Japan
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 a 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.
Session IC-5.E Modeling fundamental radiation effects
IC-5.E:IL27 Atomistic Simulations of Radiation Damage in Materials: A Self-consistent Approach Bridging Macro- and Microscales
S.L. DUDAREV, UK Atomic Energy Authority, Culham Campus, Abingdon, Oxfordshire, UK
The operating conditions for materials in a fusion power plant can be derived from full fusion device simulations, requiring an approach combining experimental observations and quantitative interpretation of experiments using microstructural evolutionary models. The development of such models, and the formulation of appropriate experimental tests, realisable without extraordinary effort and/or involving excessive costs, is reviewed in this presentation. Given that the macroscopic engineering models use a limited set of thermodynamic macroscopic variables, focusing solely on the mechanics of materials, whereas the radiation damage effects are microscopic phenomena, bridging the scales in a consistent mathematical manner represents the central aspect of the required analysis. We shall review the microscopic ab initio simulations of radiation defects in metals across the Periodic Table, focusing in particular on the effects of symmetry breaking, the X-ray and TEM-based experimental observations of these effects, and on the role that symmetry breaking plays in the microstructural evolution of materials. We also discuss and address the specific computational requirements derived from the fact that computational models for materials must be compatible with full device simulations.
IC-5.E:IL28 In Search of Prototypical Neutron Sources for Fusion Materials Testing using Computational Materials Modeling
J. MARIAN, University of California Los Angeles, Los Angeles, CA, USA
This presentation will highlight activities in the United States over the past 18 months to assess the suitability of several high-energy neutron sources as a prototypical fusion materials testing facility. The analysis performed involves the participation of experts in the areas of neutronics, chemical inventory changes, computational thermodynamics, primary damage production, and long-term microstructural evolution. The studies have been conducted for materials catalogued as primary candidates for structural applications in fusion reactors: reduced-activation ferritic-martensitic steels, vanadium alloys, tungsten alloys, and silicon carbide. The neutron spectra evaluated for prototypicality include several MFE fusion designs, the IFMIF-DONES concept, spallation neutron sources, and a pulsed irradiation spectrum representative of inertial fusion energy concepts. We have defined quantitative materials performance metrics to evaluate the proximity of these sources to a true fusion neutron spectrum in terms of He and H to displacement damage ratios, cascade damage production, transmutation inventory, and the formation of transmutation-induced thermodynamic phases. As such, this study is intended to inform the decision to select an appropriate fusion materials testing facility.
Session IC-5.F Materials modelling and database
IC-5.F:IL29 Improved Experimental Accuracy in Light Ion Interactions with Plasma-Facing Materials and its Effect on Sputter Yield Predictions
E. PITTHAN1, M. FELLINGER2, B. BURAZOR DOMAZET2, P.M. WOLF1, J. SHAMS-LATIFI1, M. RUBEL1,3, F. AUMAYR2, D. PRIMETZHOFER1, 1Department of Physics and Astronomy, Ångström Laboratory, Uppsala University, Uppsala, Sweden; 2Institute of Applied Physics, TU Wien, Fusion@ÖAW, Vienna, Austria; 3KTH Royal Institute of Technology, Stockholm, Sweden
We investigate key parameters governing ion–solid interactions relevant for plasma-wall interaction in nuclear fusion devices. We further assess, how, uncertainties propagate and influence sputtering yield simulations in the binary collision approximation framework. Our study examines recent experimental evaluations of electronic stopping power and short-range repulsive interatomic potentials for light plasma species (H, D, He) in candidate plasma-facing materials (W, Fe, EUROFER97), and contrasts these findings with commonly used semi-empirical and theoretical models. At low ion energies, we observe deviations in specific energy loss up to 60% compared to SRIM-2013 and as much as 210% relative to SRIM-1997, underscoring the need for more accurate input data. Using SDTrimSP simulations, we analyze the impact of these corrections on sputter yields and compare with new experimental data obtained from re-deposited thin films of W, Fe, and EUROFER97 measured via a high-sensitivity quartz crystal microbalance. We also address experimental challenges in assessing quantities such as electronic energy deposition at low ion energies as well as remaining uncertainties and limitations in the models employing this input, including effects related to crystal structure and ion implantation.
IC-5.F:IL30 Modelling of Thermo-mechanical Properties and Damage Evolution in SiC-SiC Composites
YANG CHEN, Department of Mechanical Engineering, University of Bath, Bath, UK
The application of SiC/SiC composites in nuclear systems requires a deep understanding of irradiation-induced dimensional changes and their impact on mechanical and thermal properties. Neutron irradiation can cause swelling mismatches between fibres and matrix, leading to residual stresses, matrix cracking, and interface debonding. These effects degrade both mechanical integrity and thermal conductivity, which is critical for heat management in reactor environments. Microscale modelling offers insight into these phenomena but demands high computational resources due to the need for large RVEs with thousands of fibres and sub-micron interphase resolution. Conventional finite element tools struggle with efficiency at this scale. To overcome this, we present a fast and scalable numerical solver based on the Fast Fourier Transform (FFT) method, optimised for parallel processing and HPC platforms. This approach enables efficient simulation of coupled thermo-mechanical behaviour of irradiated SiC/SiC composites.
Session IC-5.G Crosscutting materials issues, present status, challenges and directions for nuclear fission and fusion science and technology
IC-5.G:IL31 From Perfection to Pragmatism: A Materials-Anchored Path Toward Timely Fusion Power
J.-W. COENEN, Forschungszentrum Jülich GmbH, Jülich, Germany; University of Wisconsin-Madison, Madison, WI, USA; Gauss Fusion GmbH, Garching bei München, Germany
Fusion energy is approaching technological maturity, yet progress is slowed by excessive pre-deployment conservatism. Structural steels, plasma-facing components, breeder ceramics and tritium systems already reach Technology Readiness Level >4, but final qualification requires operation in real reactor environments. This contribution advocates a structured-pragmatism approach in which a first-of-a-kind (FOAK) fusion plant acts as the materials substantiation platform. A bounded materials baseline—Grade 91 or EUROFER steels, W/Cu divertor targets, Li₄SiO₄–Li₂TiO₃ breeder ceramics, and Pb multipliers—is coupled to in-service surveillance and staged approach substantion Recent IAEA assessments confirm that only integrated operation can advance to TRL 7–8, while critical-materials analyses show material resources are sufficient if recycling and early supply-chain activation begin now. Structured pragmatism enables earlier licensing, faster learning, and real industrial engagement, bridging the gap between laboratory maturity and commercial viability, and accelerating fusion from perpetual promise to sustainable, deployable technology.
IC-5.G:IL32 Synergistic Effects of Atomic Displacement Damage and Solid Transmutations In Neutron Irradiated Tungsten
YUTAI KATOH1, WEICHENG ZHONG1, CHAD PARISH1, TAKAAKI KOYANAGI1, H. GIETL1,6, J. TRELEWICZ1,2, Z. UNTERBERG1, W. GERINGER1, FRANK CHEN1, XIAO-YING YU1, L.M. GARRISON1,6, D. SPROUSTER2, TAKESHI MIYAZAWA3,6, MAKOTO FUKUDA3,6, TATSUYA HINOKI4, MASAFUMI AKIYOSHI5, SHUHEI NOGAMI3,6, AKIRA HASEGAWA3, 1Oak Ridge National Laboratory, Oak Ridge, TN, USA; 2Stony Brook University, Stony Brook, NY, USA; 3Tohoku University, Sendai, Miyagi-ken, Japan; 4Kyoto University, Kyoto-shi, Kyoto-fu, Japan; 5Osaka Metropolitan University, Sakai-shi, Osaka-fu, Japan; 6 Currently affiliated with other institutions
As the candidate plasma-facing material for the near-term fusion technology demonstration devices, tungsten materials are faced by challenges associated with nuclear environments anticipated in power reactors. Studies conducted using fission neutrons have revealed many irradiation effect factors contributing to the performance limits of tungsten materials, including the ductile-to-brittle transition, recrystallization, and intermediate temperature embrittlement. A unique challenge in understanding the neutron effects in tungsten is arising from complex interactions of atomic displacement damage and relatively rapid nuclear transmutations. In the present work, synergistic effects of atomic displacement and solid transmutations in tungsten are discussed based upon the results from irradiation studies conducted in the High Flux Isotope Reactor (HFIR) of ORNL with and without thermal neutron shielding, followed by comparative analyses for mechanical behaviors, thermophysical properties, and microstructural evolutions.
Research supported by the U.S. Department of Energy, Office of Science, Fusion Energy Sciences and U.S.-Japan PHENIX/FRONTIER collaboration project; Abstract authored by UT-Battelle, LLC, under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy.