In nuclear fusion power research, the plasma-facing material (or materials) is any material used to construct the plasma-facing components (PFC) those components exposed to the plasma within which nuclear fusion occurs, and particularly the material used for the lining the first wall or divertor region of the reactor vessel.
Plasma-facing materials for fusion reactor designs must support the overall steps for energy generation, these include: Generating heat through fusion, Capturing heat in the first wall, Transferring heat at a faster rate than capturing heat. Generating electricity.
In addition plasma-facing materials have to operate over the lifetime of a fusion reactor vessel by handling the harsh environmental conditions, such as: Ion bombardment causing physical and chemical sputtering and therefore erosion. Ion implantation causing displacement damage and chemical composition changes High-heat fluxes (e.g. 10 MW/m) due to ELMs and other transients. Limited tritium codeposition and sequestration. Stable thermomechanical properties under operation. Limited number of negative nuclear transmutation effects
Currently, fusion reactor research focuses on improving efficiency and reliability in heat generation and capture and on raising the rate of transfer. Generating electricity from heat is beyond the scope of current research, due to existing efficient heat-transfer cycles, such as heating water to operate steam turbines that drive electrical generators.
Current reactor designs are fueled by deuterium-tritium (D-T) fusion reactions, which produce high-energy neutrons that can damage the first wall, however, high-energy neutrons (14.1 MeV) are needed for blanket and Tritium breeder operation. Tritium is not a naturally abundant isotope due to its short half-life, therefore for a fusion D-T reactor it will need to be bred by the nuclear reaction of lithium (Li), boron (B), or beryllium (Be) isotopes with high-energy neutrons that collide within the first wall.
Requirements Most magnetic confinement fusion devices (MCFD) consist of several key components in their technical designs, including: Magnet system: confines the deuterium-tritium fuel in the form of plasma and in the shape of a torus. Vacuum vessel: contains the core fusion plasma and maintains fusion conditions. First wall: positioned between the plasma and magnets in order to protect outer vessel components from radiation damage. Cooling system: removes heat from the confinement and transfers heat from the first wall. The core fusion plasma must not actually touch the first wall. ITER and many other current and projected fusion experiments, particularly those of the tokamak and stellarator designs, use intense magnetic fields in an attempt to achieve this, although plasma instability problems remain. Even with stable plasma confinement, however, the first wall material would be exposed to a neutron flux higher than in any current nuclear power reactor, which leads to two key problems in selecting the material: It must withstand this neutron flux for a sufficient period of time to be economically viable. It must not become sufficiently radioactive so as to produce unacceptable amounts of nuclear waste when lining replacement or plant decommissioning eventually occurs.
The lining material must also: Allow the passage of a large heat flux. Be compatible with intense and fluctuating magnetic fields. Minimize contamination of the plasma. Retain minimal amounts of plasma fuel species (mainly tritium) Be produced and replaced at a reasonable cost.
Some critical plasma-facing components, such as and in particular the divertor, are typically protected by a different material than that used for the major area of the first wall.
Proposed materials
Materials currently in use or under consideration include: Tungsten Molybdenum Beryllium Lithium Tin Boron carbide Silicon carbide Carbon fibre composite (CFC) Graphite
Multi-layer tiles of several of these materials are also being considered and used, for example: A thin molybdenum layer on graphite tiles. A thin tungsten layer on graphite tiles. A tungsten layer on top of a molybdenum layer on graphite tiles. A boron carbide layer on top of CFC tiles. A liquid lithium layer on graphite tiles. A liquid lithium layer on top of a boron layer on graphite tiles. A liquid lithium layer on tungsten-based solid plasma-facing component surfaces or divertors.
Graphite was used for the first wall material of the Joint European Torus (JET) at its startup (1983), in Tokamak à configuration variable (1992) and in National Spherical Torus Experiment (NSTX, first plasma 1999).
Beryllium was used to reline JET in 2009 in anticipation of its proposed use in ITER.
Tungsten is used for the divertor in JET, and will be used for the divertor in ITER. It is also used for the first wall in ASDEX Upgrade. Graphite tiles plasma sprayed with tungsten were used for the ASDEX Upgrade divertor. Studies of tungsten in the divertor have been conducted at the DIII-D facility. These experiments utilized two rings of tungsten isotopes embedded in the lower divertor to characterize erosion tungsten during operation. Molybdenum is used for the first wall material in Alcator C-Mod (1991).
Liquid lithium was used to coat the plasma-facing component of the Tokamak Fusion Test Reactor in the Lithium Tokamak Experiment (TFTR, 1996) and is currently used in Princeton Plasma Physics Laboratory's (PPPL) Lithium Tokamak Experiment-𝛽 (LTX-𝛽).
Considerations Development of satisfactory plasma-facing materials is one of the key problems still to be solved by current programs.
Plasma-facing materials can be measured for performance in terms of: Power production for a given reactor size. Cost to generate electricity. Self-sufficiency of tritium production. Availability of materials. Design and fabrication of the plasma-facing components. Safety in waste disposal and in maintenance. The International Fusion Materials Irradiation Facility will particularly address this. Materials developed using this facility will be used in DEMOnstration Power Station, the proposed successor to ITER.
French Nobel laureate in physics Pierre-Gilles de Gennes said of nuclear fusion, "We say that we will put the sun into a box. The idea is pretty. The problem is, we don't know how to make the box."Michio Kaku, Physics of the Impossible, pp.46-47.
Recent developments Solid plasma-facing materials are known to be susceptible to damage under large heat loads and high neutron flux. If damaged, these solids can contaminate the plasma and decrease plasma confinement stability. In addition, radiation can leak through defects in the solids and contaminate outer vessel components.
Graphite is being phased out as a plasma-facing wall material because it undergoes strong chemical sputtering in hydrogen plasmas, leading to high erosion, tritium co-deposition and retention, dust formation, and impurity influx, while also suffering from neutron-induced degradation and limited lifetime under reactor heat loads; for example, measurements in the W7-X graphite divertor showed net erosion of up to 20 μm at the strike line and a total of 48 ± 14 g of carbon removed over 3776 s of plasma exposure, corresponding to effective erosion rates of ~5–8 nm s⁻¹.
Liquid metal plasma-facing components that enclose the plasma have been proposed to address challenges in the plasma-facing components. In particular, liquid lithium has been confirmed to have various properties that are attractive for fusion reactor performance. In 2026, researchers from PPPL, Princeton University and Pennsylvania State University demonstrated one effective approach to clearing a roadblock for using liquid lithium to protect the inner surfaces of fusion systems: scrubbing porous tungsten samples inside the LTX-𝛽 by heating the tiles and bombarding them with particles from a neon plasma to knock contaminants out of tiles. The approach could be used after the tiles have been sealed inside a vacuum vessel.
Tungsten Tungsten is widely recognized as the preferred material for plasma-facing components in next-generation fusion devices, largely due to its unique combination of properties and potential for enhancement. Its low erosion rates make it particularly suitable for the high-stress environment of fusion reactors, where it can withstand the intense conditions without degrading rapidly. Additionally, tungsten's low tritium retention through co-deposition and implantation is crucial in fusion contexts, helping to minimize the accumulation of this radioactive isotope.
Another key advantage of tungsten is its high thermal conductivity, essential for managing the extreme heat generated in fusion processes. This property ensures efficient heat dissipation, reducing the risk of damage to the reactor's internal components. Furthermore, the potential for developing radiation-hardened alloys of tungsten presents an opportunity to enhance its durability and performance under the intense radiation conditions typical in fusion reactors.
Despite these benefits, tungsten is not without its drawbacks. One notable issue is its tendency to contribute to high core radiation, a significant challenge in maintaining the plasma performance in fusion reactors. Another notable drawback of tungsten is the formation of tendril-like structures, also known as 'tungsten fuzz' by the plasma-facing materials community, on the plasma-facing surface, particularly under helium-ion irradiation at reactor-relevant temperatures. This phenomenon results from helium bubble formation and migration, which is common in tokamak environments because energetic helium particles are a byproduct of the DT fusion reaction. The formation of tungsten fuzz reduces the PFC's thermal conductivity and could further contaminate the plasma if tendrils detach from the surface. Nevertheless, tungsten has been selected as the plasma-facing material for the ITER project's first-generation divertor, and it is likely to be used for the reactor's first wall as well.
Another drawback of the use of tungsten as plasma-facing material is that it is not very effective in removing, or gettering, impurities, such as oxygen and water, from the vacuum environment as opposed to for example beryllium. The presence of impurities in the vessel can lead to significant radiation losses in the core plasma and has also led to disruptions of the plasma. For this reason, a method known as boronization has been employed in several fusion devices with tungsten plasma-facing components and is planned to be used for ITER as well. A consequence of boronization of the tungsten plasma-facing components will be the formation of tungsten-boron (W-B) mixtures, the properties of which may be very different from those of tungsten by itself. Of special concern is the change in fuel retention properties of the W-B mixtures that form.
Understanding the behavior of tungsten in fusion environments, including its sourcing, migration, and transport in the scrape-off-layer (SOL), as well as its potential for core contamination, is a complex task. Significant research is ongoing to develop a mature and validated understanding of these dynamics, particularly for predicting the behavior of high-Z (high atomic number) materials like tungsten in next-step tokamak devices.
To address tungsten's intrinsic brittleness, which limits its operational window, a composite material known as W-fibre enhanced W-composite (Wf/W) has been developed. This material incorporates extrinsic toughening mechanisms to significantly increase toughness, as demonstrated in small Wf/W samples.