Divertor

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Interior of Alcator C-Mod showing the lower divertor channel at the bottom of the torus Alcator C-Mod Tokamak Interior.jpg
Interior of Alcator C-Mod showing the lower divertor channel at the bottom of the torus
Divertor design for K-DEMO, a planned future tokamak experiment K-DEMO divertor module.jpg
Divertor design for K-DEMO, a planned future tokamak experiment
Divertor of COMPASS BPP LP divertor field 2017 new.jpg
Divertor of COMPASS

In magnetic confinement fusion, a divertor or diverted configuration is a magnetic field configuration of a tokamak or a stellarator which separates the confined plasma from the material surface of the device. The plasma particles which diffuse across the boundary of the confined region are diverted by the open, wall-intersecting magnetic field lines to wall structures which are called the divertor targets, usually remote from the confined plasma. The magnetic divertor extracts heat and ash produced by the fusion reaction, minimizes plasma contamination, and protects the surrounding walls from thermal and neutronic loads.

Contents

The term divertor usually describes the magnetic configuration itself or the region between the confined plasma and the target. Sometimes divertor target and divertor are used interchangeably. For example, the ITER divertor refers to the heavily engineered plasma-facing components designed to handle the intense plasma-wall interactions foreseen.

History

The divertor was initially introduced during the earliest studies of fusion power systems in the 1950s. It was realized early on that successful fusion would result in heavier ions being created and left in the fuel (the so-called "fusion ash"). These impurities were responsible for the loss of heat, and caused other effects that made it more difficult to keep the reaction going. The divertor was proposed as a solution to this problem. Operating on the same principle as a mass spectrometer, the plasma passes through the divertor region where heavier ions are flung out of the fuel mass by centrifugal force, colliding with some sort of absorber material, and depositing its energy as heat. [1] Initially considered to be a device required for operational reactors, few early designs included a divertor.

When early long-shot reactors started to appear in the 1970s, a serious practical problem emerged. No matter how tightly constrained, plasma continued to leak out of the main confinement area, striking the walls of the reactor core and causing problems. A major concern was sputtering in reactors with higher power and particle flux density, [2] which caused ions of the vacuum chamber's wall metal to flow into the fuel and to cool it.

During the 1980s it became common for reactors to include a feature known as the limiter, which is a small piece of material that projects a short distance into the outer edge of the main plasma confinement area. Ions from the fuel that are travelling outwards strike the limiter, thereby protecting the walls of the chamber from this damage. However, the problems with material being deposited into the fuel remained; the limiter simply changed where that material was coming from.

This led to the re-emergence of the divertor, as a device for protecting the reactor itself. In these designs, magnets pull the lower edge of the plasma to create a small region where the outer edge of the plasma, the "Scrape-Off Layer" (SOL), hits a limiter-like plate. The divertor improves on the limiter in several ways, mainly because modern reactors try to create plasmas with D-shaped cross-sections ("elongation" and "triangularity") so the lower edge of the D is a natural location for the divertor. In modern examples the plates are replaced by lithium metal, which better captures the ions and causes less cooling when it enters the plasma. [3]

In ITER and the latest configuration of Joint European Torus, the lowest region of the torus is configured as a divertor, [4] while Alcator C-Mod was built with divertor channels at both top and bottom. [5] A divertor design called Super-X has been designed to reduce the heat density in the divertor by adopting a design resembling a funnel. [6]

Tokamak divertors

A tokamak featuring a divertor is known as a divertor tokamak or divertor configuration tokamak. In this configuration, the particles escape through a magnetic "gap" (separatrix), which allows the energy absorbing part of the divertor to be placed outside the plasma. The divertor configuration also makes it easier to obtain a more stable H-mode of operation. The plasma facing material in the divertor faces significantly different stresses compared to the majority of the first wall.

Stellarator divertors

In stellarators, low-order magnetic islands can be used to form a divertor volume, the island divertor, for managing power and particle exhaust. [7] The island divertor has shown success in accessing and stabilizing detached scenarios and has demonstrated reliable heat flux and detachment control with hydrogen gas injection, and impurity seeding in the W7-X stellarator. [8] [9] The magnetic island chain in the plasma edge can control plasma fueling. [10] Despite some challenges, the island divertor concept has demonstrated great potential for managing power and particle exhaust in fusion reactors, and further research could lead to more efficient and reliable operation in the future. [11]

The helical divertor, as employed in the Large Helical Device (LHD), utilizes large helical coils to create a diverting field. This design permits adjustment of the stochastic layer size, situated between the confined plasma volume and the field lines ending on the divertor plate. However, the compatibility of the Helical Divertor with stellarators optimized for neoclassical transport remains uncertain. [12]

The non-resonant divertor provides an alternative design for optimized stellarators with significant bootstrap currents. This approach leverages sharp "ridges" on the plasma boundary to channel flux. The bootstrap currents modify the shape, not the location, of these ridges, providing an effective channeling mechanism. This design, although promising, has not been experimentally tested yet. [13]

Given the complexity of the design of stellarator divertors, compared to their two-dimensional tokamak counterparts, a thorough understanding of their performance is crucial in stellarator optimization. The experiments with divertors in the W7-X and LHD have shown promising results and provide valuable insights for future improvements in shape and performance. Furthermore, the advent of non-resonant divertors offers an exciting path forward for quasi-symmetric stellarators and other configurations not optimized for minimizing plasma currents. [14]

See also

Related Research Articles

<span class="mw-page-title-main">Stellarator</span> Plasma device using external magnets to confine plasma

A stellarator is a device that confines plasma using external magnets. Scientists researching magnetic confinement fusion aim to use stellarator devices as a vessel for nuclear fusion reactions. The name refers to stars as fusion also occurs in stars such as the Sun. It is one of the earliest fusion power devices, along with the z-pinch and magnetic mirror.

<span class="mw-page-title-main">Tokamak</span> Magnetic confinement device used to produce thermonuclear fusion power

A tokamak is a device which uses a powerful magnetic field generated by external magnets to confine plasma in the shape of an axially-symmetrical torus. The tokamak is one of several types of magnetic confinement devices being developed to produce controlled thermonuclear fusion power. The tokamak concept is currently one of the leading candidates for a practical fusion reactor.

<span class="mw-page-title-main">Plasma stability</span> Degree to which disturbing a plasma system at equilibrium will destabilize it

In plasma physics, plasma stability concerns the stability properties of a plasma in equilibrium and its behavior under small perturbations. The stability of the system determines if the perturbations will grow, oscillate, or be damped out. It is an important consideration in topics such as nuclear fusion and astrophysical plasma.

<span class="mw-page-title-main">Fusion power</span> Electricity generation through nuclear fusion

Fusion power is a proposed form of power generation that would generate electricity by using heat from nuclear fusion reactions. In a fusion process, two lighter atomic nuclei combine to form a heavier nucleus, while releasing energy. Devices designed to harness this energy are known as fusion reactors. Research into fusion reactors began in the 1940s, but as of 2024, no device has reached net power, although net positive reactions have been achieved.

This timeline of nuclear fusion is an incomplete chronological summary of significant events in the study and use of nuclear fusion.

<span class="mw-page-title-main">Large Helical Device</span>

The Large Helical Device (LHD) is a fusion research device located in Toki, Gifu, Japan. It is operated by the National Institute for Fusion Science, and is the world's second-largest superconducting stellarator, after Wendelstein 7-X. The LHD employs a heliotron magnetic field originally developed in Japan.

<span class="mw-page-title-main">Magnetic confinement fusion</span> Approach to controlled thermonuclear fusion using magnetic fields

Magnetic confinement fusion (MCF) is an approach to generate thermonuclear fusion power that uses magnetic fields to confine fusion fuel in the form of a plasma. Magnetic confinement is one of two major branches of controlled fusion research, along with inertial confinement fusion.

<span class="mw-page-title-main">ASDEX Upgrade</span>

ASDEX Upgrade is a divertor tokamak at the Max-Planck-Institut für Plasmaphysik, Garching that went into operation in 1991. At present, it is Germany's second largest fusion experiment after stellarator Wendelstein 7-X.

<span class="mw-page-title-main">Tokamak à configuration variable</span> Swiss research fusion reactor at the École Polytechnique Fédérale de Lausanne

The tokamak à configuration variable is an experimental tokamak located at the École Polytechnique Fédérale de Lausanne (EPFL) Swiss Plasma Center (SPC) in Lausanne, Switzerland. As the largest experimental facility of the Swiss Plasma Center, the TCV tokamak explores the physics of magnetic confinement fusion. It distinguishes itself from other tokamaks with its specialized plasma shaping capability, which can produce diverse plasma shapes without requiring hardware modifications.

<span class="mw-page-title-main">Wendelstein 7-X</span> Modern stellarator for plasma fusion experiments

The Wendelstein 7-X reactor is an experimental stellarator built in Greifswald, Germany, by the Max Planck Institute for Plasma Physics (IPP), and completed in October 2015. Its purpose is to advance stellarator technology: though this experimental reactor will not produce electricity, it is used to evaluate the main components of a future fusion power plant; it was developed based on the predecessor Wendelstein 7-AS experimental reactor.

<span class="mw-page-title-main">Helically Symmetric Experiment</span>

The Helically Symmetric Experiment, is an experimental plasma confinement device at the University of Wisconsin–Madison, with design principles that are intended to be incorporated into a fusion reactor. The HSX is a modular coil stellarator which is a toroid-shaped pressure vessel with external electromagnets which generate a magnetic field for the purpose of containing a plasma. It began operation in 1999.

Magnetically confined fusion plasmas such as those generated in tokamaks and stellarators are characterized by a typical shape. Plasma shaping is the study of the plasma shape in such devices, and is particularly important for next step fusion devices such as ITER. This shape is conditioning partly the performance of the plasma. Tokamaks, in particular, are axisymmetric devices, and therefore one can completely define the shape of the plasma by its cross-section.

Plasma–Surface Interaction (PSI) studies study the interaction at the interface between plasma and materials. Focus of the research lies on providing both theoretical and experimental support to the design and validation of plasma facing materials for the fusion experiment ITER and future devices.

<span class="mw-page-title-main">Plasma-facing material</span>

In nuclear fusion power research, the plasma-facing material (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.

<span class="mw-page-title-main">Hybrid Illinois Device for Research and Applications</span> Toroidal magnetic fusion device

The Hybrid Illinois Device for Research and Applications (HIDRA) is a medium-sized toroidal magnetic fusion device housed in the Nuclear Radiation Laboratory and operated by the Center for Plasma-Material Interactions (CPMI) within the Department of Nuclear, Plasma and Radiological Engineering at the University of Illinois at Urbana–Champaign, United States. HIDRA had its first plasma at the end of April 2016 and started experimental campaigns by December of that year. HIDRA is the former WEGA classical stellarator that was operated at the Max Planck Institute for Plasma Physics in Greifswald Germany from 2001 to 2013.

<span class="mw-page-title-main">Princeton Large Torus</span> Experimental fusion reactor, first to hit 75 million degrees

The Princeton Large Torus, was an early tokamak built at the Princeton Plasma Physics Laboratory (PPPL). It was one of the first large scale tokamak machines and among the most powerful in terms of current and magnetic fields. Originally built to demonstrate that larger devices would have better confinement times, it was later modified to perform heating of the plasma fuel, a requirement of any practical fusion power device.

<span class="mw-page-title-main">Compact Toroidal Hybrid</span>

The Compact Toroidal Hybrid (CTH) is an experimental device at Auburn University that uses magnetic fields to confine high-temperature plasmas. CTH is a torsatron type of stellarator with an external, continuously wound helical coil that generates the bulk of the magnetic field for containing a plasma.

<span class="mw-page-title-main">Wendelstein 7-AS</span> Stellarator for plasma fusion experiments (1988-2002)

Wendelstein 7-AS was an experimental stellarator which was in operation from 1988 to 2002 by the Max Planck Institute for Plasma Physics (IPP) in Garching. It was the first of a new class of advanced stellarators with modular coils, designed with the goal of developing a nuclear fusion reactor to generate electricity.

<span class="mw-page-title-main">Omnigeneity</span> A concept in stellarator physics

Omnigeneity is a property of a magnetic field inside a magnetic confinement fusion reactor. Such a magnetic field is called omnigenous if the path a single particle takes does not drift radially inwards or outwards on average. A particle is then confined to stay on a flux surface. All tokamaks are exactly omnigenous by virtue of their axisymmetry, and conversely an unoptimized stellarator is generally not omnigenous.

References

  1. ↑ "RF Absorbers material types". www.masttechnologies.com. Retrieved 30 August 2015.
  2. ↑ "Fusrev". Archived from the original on 2014-01-10. Retrieved 2014-01-10.] T N Todd and C G Windsor, Progress in Magnetic Confinement Fusion Research, Contemporary Physics, 1998, volume 39, number 4, pages 255-282
  3. ↑ "Limiters and Divertors" Archived January 10, 2014, at the Wayback Machine , EFDA
  4. ↑ Stoafer, Chris (14 April 2011). "Tokamak Divertor System Concept and the Design for ITER" (PDF). www.apam.columbia.edu. Archived from the original (PDF) on 2013-12-11. Retrieved 11 September 2012.
  5. ↑ "MIT Plasma Science & Fusion Center: Research>alcator>information". Archived from the original on 2012-06-17. Retrieved 2012-09-11. retrieved September 11, 2012
  6. ↑ "First results from UK tokamak offers a STEP towards commercial fusion". 25 May 2021.
  7. ↑ Feng, Y; et al. (2006). "Physics of island divertors as highlighted by the example of W7-AS". Nucl. Fusion. 46 (8): 807–819. Bibcode:2006NucFu..46..807F. doi:10.1088/0029-5515/46/8/006. hdl: 11858/00-001M-0000-0027-0DC4-8 . S2CID   62893155.
  8. ↑ Schmitz, O; et al. (2021). "Stable heat and particle flux detachment with efficient particle exhaust in the island divertor of Wendelstein 7-X". Nucl. Fusion. 61 (1): 016026. Bibcode:2021NucFu..61a6026S. doi:10.1088/1741-4326/abb51e. hdl: 21.11116/0000-0007-A4DC-8 . OSTI   1814444. S2CID   225288529.
  9. ↑ Effenberg, F; et al. (2019). "First demonstration of radiative power exhaust with impurity seeding in the island divertor at Wendelstein 7-X" (PDF). Nucl. Fusion. 59 (10): 106020. Bibcode:2019NucFu..59j6020E. doi:10.1088/1741-4326/ab32c4. S2CID   199132000.
  10. ↑ Stephey, L; et al. (2018). "Impact of magnetic islands in the plasma edge on particle fueling and exhaust in the HSX and W7-X stellarators". Physics of Plasmas. 25 (6): 062501. Bibcode:2018PhPl...25f2501S. doi:10.1063/1.5026324. hdl: 21.11116/0000-0001-6AE2-9 . S2CID   125652747.
  11. ↑ Jakubowksi, M; et al. (2021). "Overview of the results from divertor experiments with attached and detached plasmas at Wendelstein 7-X and their implications for steady-state operation". Nucl. Fusion. 61 (10): 106003. Bibcode:2021NucFu..61j6003J. doi: 10.1088/1741-4326/ac1b68 . S2CID   237408135.
  12. ↑ Morisaki, T; et al. (2013). "Initial experiments towards edge plasma control with a closed helical divertor in LHD". Nucl. Fusion. 53 (6): 063014. Bibcode:2013NucFu..53f3014M. doi:10.1088/0029-5515/53/6/063014. S2CID   122537627.
  13. ↑ Boozer, A.H. (2015). "Stellarator design". Journal of Plasma Physics. 81 (6): 515810606. Bibcode:2015JPlPh..81f5106B. doi:10.1017/S0022377815001373.
  14. ↑ Bader, Aaron (December 6, 2018). "Progress in Divertor and Edge Transport Research for Stellarator Plasmas" (PDF). Archived from the original (PDF) on 2023-07-26.

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