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Plasma-facing material

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In nuclear fusion power research, the plasma-facing material (or materials) (PFM) 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.

Interior of Alcator C-Mod showing the molybdenum tiles used as first wall material
Interior of Tokamak à configuration variable showing the graphite tiles used as first wall material

Plasma-facing materials for fusion reactor designs must support the overall steps for energy generation, these include:

  1. Generating heat through fusion,
  2. Capturing heat in the first wall,
  3. Transferring heat at a faster rate than capturing heat.
  4. Generating electricity.

In addition PFMs have to operate over the lifetime of a fusion reactor vessel by handling the harsh environmental conditions, such as:

  1. Ion bombardment causing physical and chemical sputtering and therefore erosion.
  2. Ion implantation causing displacement damage and chemical composition changes
  3. High-heat fluxes (e.g. 10 MW/m) due to ELMS and other transients.
  4. Limited tritium codeposition and sequestration.
  5. Stable thermomechanical properties under operation.
  6. 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,[1] 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.[2]

Requirements

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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.
  • 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.[3]

Proposed materials

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Materials currently in use or under consideration include:

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.[6]
  • A liquid lithium layer on graphite tiles.[8]
  • A liquid lithium layer on top of a boron layer on graphite tiles.[9]
  • A liquid lithium layer on tungsten-based solid PFC surfaces or divertors.[10]

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).[11]

Beryllium was used to reline JET in 2009 in anticipation of its proposed use in ITER.[12]

Tungsten is used for the divertor in JET, and will be used for the divertor in ITER.[12][13] It is also used for the first wall in ASDEX Upgrade.[14] Graphite tiles plasma sprayed with tungsten were used for the ASDEX Upgrade divertor.[15] Studies of tungsten in the divertor have been conducted at the DIII-D facility.[16] 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 (LL) was used to coat the PFC of the Tokamak Fusion Test Reactor in the Lithium Tokamak Experiment (TFTR, 1996).[8]

Considerations

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Development of satisfactory plasma-facing materials is one of the key problems still to be solved by current programs.[17][18]

Plasma-facing materials can be measured for performance in terms of:[9]

  • Power production for a given reactor size.
  • Cost to generate electricity.
  • Self-sufficiency of tritium production.
  • Availability of materials.
  • Design and fabrication of the PFC.
  • Safety in waste disposal and in maintenance.

The International Fusion Materials Irradiation Facility (IFMIF) will particularly address this. Materials developed using IFMIF will be used in DEMO, 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."[19]

Recent developments

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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.[1]

Liquid metal plasma-facing components that enclose the plasma have been proposed to address challenges in the PFC. In particular, liquid lithium (LL) has been confirmed to have various properties that are attractive for fusion reactor performance.[1]

Tungsten

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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.[20][21][22]

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. 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.

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.

In the context of future fusion power plants, tungsten stands out for its resilience against erosion, the highest melting point among metals, and relatively benign behavior under neutron irradiation. However, its ductile to brittle transition temperature (DBTT) is a concern, especially as it increases under neutron exposure. To overcome this brittleness, several strategies are being explored, including the use of nanocrystalline materials, tungsten alloying, and W-composite materials.

Particularly notable are the tungsten laminates and fiber-reinforced composites, which leverage tungsten's exceptional mechanical properties. When combined with copper's high thermal conductivity, these composites offer improved thermomechanical properties, extending beyond the operational range of traditional materials like CuCrZr. For applications requiring even higher temperature resilience, tungsten-fibre reinforced tungsten-composites (Wf/W) have been developed, incorporating mechanisms to enhance toughness, thereby broadening the potential applications of tungsten in fusion technology.

Lithium

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Lithium (Li) is an alkali metal with a low Z (atomic number). Li has a low first ionization energy of ~5.4 eV and is highly chemically reactive with ion species found in the plasma of fusion reactor cores. In particular, Li readily forms stable lithium compounds with hydrogen isotopes, oxygen, carbon, and other impurities found in D-T plasma.[1]

The fusion reaction of D-T produces charged and neutral particles in the plasma. The charged particles remain magnetically confined to the plasma. The neutral particles are not magnetically confined and will move toward the boundary between the hotter plasma and the colder PFC. Upon reaching the first wall, both neutral particles and charged particles that escaped the plasma become cold neutral particles in gaseous form. An outer edge of cold neutral gas is then “recycled”, or mixed, with the hotter plasma. A temperature gradient between the cold neutral gas and the hot plasma is believed to be the principal cause of anomalous electron and ion transport from the magnetically confined plasma. As recycling decreases, the temperature gradient decreases and plasma confinement stability increases. With better conditions for fusion in the plasma, the reactor performance increases.[23]

Initial use of lithium in 1990s was motivated by a need for a low-recycling PFC. In 1996, ~ 0.02 grams of lithium coating was added to the PFC of TFTR, resulting in the fusion power output and the fusion plasma confinement to improve by a factor of two. On the first wall, lithium reacted with neutral particles to produce stable lithium compounds, resulting in low-recycling of cold neutral gas. In addition, lithium contamination in the plasma tended to be well below 1%.[1]

Since 1996, these results have been confirmed by a large number of magnetic confinement fusion devices (MCFD) that have also used lithium in their PFC, for example:[1]

  • TFTR (US), CDX-U (2005)/LTX(2010) (US), CPD (Japan), HT-7 (China), EAST (China), FTU (Italy).
  • NSTX (US), T-10 (Russia), T-11M (Russia), TJ-II (Spain), RFX (Italy).

The primary energy generation in fusion reactor designs is from the absorption of high-energy neutrons. Results from these MCFD highlight additional benefits of liquid lithium coatings for reliable energy generation, including:[1][23][8]

  1. Absorb high-energy, or fast-moving, neutrons. About 80% of the energy produced in a fusion reaction of D-T is in the kinetic energy of the newly produced neutron.
  2. Convert kinetic energies of absorbed neutrons into heat on the first wall. The heat that is produced on the first wall can then be removed by coolants in ancillary systems that generate electricity.
  3. Self-sufficient breeding of tritium by nuclear reaction with absorbed neutrons. Neutrons of varying kinetic energies will drive tritium-breeding reactions.

Liquid lithium

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Newer developments in liquid lithium are currently being tested, for example:[9][10]

  • Coatings made of increasingly complex liquid lithium compounds.
  • Multi-layered coatings of LL, B, F, and other low-Z metals.
  • Higher density coatings of LL for use on PFC designed for greater heat loads and neutron flux.

Silicon carbide

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Silicon carbide (SiC), a low-Z refractory ceramic material, has emerged as a promising candidate for structural materials in magnetic fusion energy devices. While the remarkable properties of SiC once attracted attention for fusion experiments, past technological limitations hindered its wider use. However, the evolving capabilities of SiC fiber composites (SiCf/SiC) in Gen-IV fission reactors have renewed interest in SiC as a fusion material.[24]

Modern versions of SiCf/SiC combine many desirable attributes found in carbon fiber composites, such as thermo-mechanical strength and high melting point. These versions also present unique benefits: they exhibit minimal degradation of properties when exposed to high levels of neutron damage. However, tritium retention in silicon carbide plasma-facing components is about 1.5-2 times higher than in graphite, leading to reduced fuel efficiency and increased safety risks in fusion reactors. SiC traps more tritium, limiting its availability for fusion and increasing the potential for hazardous buildup, which complicates tritium management.[25][26] Additionally, the chemical and physical sputtering of SiC is still significant and contributes to the key issue of increasing tritium inventory through co-deposition over time and with particle fluency. For those reasons, carbon-based materials have been ruled out in ITER, DEMO, and other devices.[27] SiC has demonstrated a tritium diffusivity lower than that observed in other structural materials, a property that can be further optimized by applying a thin layer of monolithic SiC on a SiC/SiCf substrate. [28][29]

Siliconization, as a wall conditioning method, has been demonstrated to reduce oxygen impurities and enhance plasma performance.[30][31] Current research efforts focus on understanding SiC behavior under conditions relevant to reactors, providing valuable insights into its potential role in future fusion technology. Silicon-rich films on divertor PFCs were recently developed using Si pellet injections in high confinement mode scenarios in DIII-D, prompting further research into refining the technique for broader fusion applications.[32]

See also

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References

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  1. ^ a b c d e f g Ono, Masayuki (2012). Lithium as Plasma Facing Component for Magnetic Fusion Research (Report). Princeton Plasma Physics Lab. doi:10.2172/1056493. OSTI 1056493.
  2. ^ Ihli, T; Basu, T.K; Giancarli, L.M; Konishi, S; Malang, S; Najmabadi, F; Nishio, S; Raffray, A.R.; Rao, C.V.S; Sagara, A; Wu, Y (December 2008). "Review of blanket designs for advanced fusion reactors". Fusion Engineering and Design. 83 (7–9): 912–919. Bibcode:2008FusED..83..912I. doi:10.1016/j.fusengdes.2008.07.039.
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  24. ^ Abrams, T.; et al. (2021). "Evaluation of silicon carbide as a divertor armor material in DIII-D H-mode discharges". Nuclear Fusion. 61 (6): 066005. arXiv:2104.04083. Bibcode:2021NucFu..61f6005A. doi:10.1088/1741-4326/abecee. S2CID 233204645.
  25. ^ Mayer, M.; Balden, M.; Behrisch, R. (1998). "Deuterium retention in carbides and doped graphites". Journal of Nuclear Materials. 252 (1): 55–62. Bibcode:1998JNuM..252...55M. doi:10.1016/S0022-3115(97)00299-7.
  26. ^ Koller, Markus T.; Davis, James W.; Goodland, Megan E.; Abrams, Tyler; Gonderman, Sean; Herdrich, Georg; Frieß, Martin; Zuber, Christian (2019). "Deuterium retention in silicon carbide, SiC ceramic matrix composites, and SiC coated graphite". Nuclear Materials and Energy. 20: 100704. Bibcode:2019NMEne..2000704K. doi:10.1016/j.nme.2019.100704.
  27. ^ Roth, Joachim; Tsitrone, E.; Loarte, A.; Loarer, Th.; Counsell, G.; Neu, R.; Philipps, V.; Brezinsek, S.; Lehnen, M.; Coad, P.; Grisolia, Ch.; Schmid, K.; Krieger, K.; Kallenbach, A.; Lipschultz, B.; Doerner, R.; Causey, R.; Alimov, V.; Shu, W.; Ogorodnikova, O.; Kirschner, A.; Federici, G.; Kukushkin, A. (2009). "Recent analysis of key plasma wall interactions issues for ITER". Journal of Nuclear Materials. 390–391: 1–9. Bibcode:2009JNuM..390....1R. doi:10.1016/j.jnucmat.2009.01.037. hdl:11858/00-001M-0000-0026-F442-2. ISSN 0022-3115.
  28. ^ Katoh, Y.; et al. (2012). "Radiation effects in SiC for nuclear structural applications". Current Opinion in Solid State and Materials Science. 16 (3): 143–152. Bibcode:2012COSSM..16..143K. doi:10.1016/j.cossms.2012.03.005.
  29. ^ Koyanagi, T.; et al. (2018). "Recent progress in the development of SiC composites for nuclear fusion applications". Journal of Nuclear Materials. 511: 544–555. Bibcode:2018JNuM..511..544K. doi:10.1016/j.jnucmat.2018.06.017. S2CID 104235507.
  30. ^ Winter, J.; et al. (1993). "Improved plasma performance in TEXTOR with silicon coated surfaces". Phys. Rev. Lett. 71 (10): 1549–1552. Bibcode:1993PhRvL..71.1549W. doi:10.1103/PhysRevLett.71.1549. PMID 10054436.
  31. ^ Samm, U.; et al. (1995). "Plasma edge physics with siliconization in TEXTOR". Journal of Nuclear Materials. 220–222: 25–35. Bibcode:1995JNuM..220...25S. doi:10.1016/0022-3115(94)00444-7.
  32. ^ Effenberg, F.; Abe, S.; Sinclair, G.; et al. (2023). "In-situ coating of silicon-rich films on tokamak plasma-facing components with real-time Si material injection". Nuclear Fusion. 63 (10): 106004. arXiv:2304.03923. Bibcode:2023NucFu..63j6004E. doi:10.1088/1741-4326/acee98. S2CID 258049235.
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