National Spherical Torus Experiment

Last updated
NSTX
National Spherical Torus Experiment
NSTX.jpg
NSTX in 2009
Device type Spherical tokamak
Location Princeton, New Jersey, US
Affiliation Princeton Plasma Physics Laboratory
Technical specifications
Major radius0.85 m (2 ft 9 in)
Minor radius0.68 m (2 ft 3 in)
Magnetic field 0.3 T (3,000 G)
Heating power11  MW
Plasma current1.4  MA
History
Year(s) of operation1999–present
Preceded by Tokamak Fusion Test Reactor (TFTR)
Links
Website NSTX-U official website
CAD drawing of NSTX NSTX CAD FInal (18842020631).jpg
CAD drawing of NSTX

The National Spherical Torus Experiment (NSTX) is a magnetic fusion device based on the spherical tokamak concept. It was constructed by the Princeton Plasma Physics Laboratory (PPPL) in collaboration with the Oak Ridge National Laboratory, Columbia University, and the University of Washington at Seattle. It entered service in 1999. In 2012 it was shut down as part of an upgrade program and became NSTX-U, U for Upgrade.

Contents

Like other magnetic confinement fusion experiments, NSTX studies the physics principles of thermonuclear plasmas—ionized gases with sufficiently high temperatures and densities for nuclear fusion to occur—which are confined in a magnetic field.

The spherical tokamak design implemented by NSTX is an offshoot of the conventional tokamak. Proponents claim that spherical tokamaks have dramatic practical advantages over conventional tokamaks. For this reason the spherical tokamak has seen considerable interest since it was proposed in the late 1980s. However, development remains effectively one generation behind mainline tokamak efforts such as JET. Other major spherical tokamak experiments include the START and MAST at Culham in the UK.

History

1999–2012

First plasma was obtained on NSTX on Friday, February 12, 1999 at 7:06 p.m.

Magnetic fusion experiments use plasmas composed of one or more hydrogen isotopes. For example, in 1994, PPPL's Tokamak Fusion Test Reactor (TFTR) produced a world-record 10.7 megawatts of fusion power from a plasma composed of equal parts of deuterium and tritium, a fuel mix likely to be used in commercial fusion power reactors. NSTX was a "proof of principle" experiment and therefore employed deuterium plasmas only. If successful it was to be followed by similar devices, eventually including a demonstration power reactor (e.g. ITER), burning deuterium-tritium fuel.

NSTX produced a spherical plasma with a hole through its center (a "cored apple" profile; see MAST), different from the doughnut-shaped (toroidal) plasmas of conventional tokamaks. The low aspect ratio A (that is, an R/a of 1.31, with the major radius R of 0.85 m and the minor radius a of 0.65 m) experimental NSTX device had several advantages including plasma stability through improved confinement. Design challenges include the toroidal and poloidal field coils, vacuum vessels and plasma-facing components. This plasma configuration can confine a higher pressure plasma than a doughnut tokamak of high aspect ratio for a given, confinement magnetic field strength. Since the amount of fusion power produced is proportional to the square of the plasma pressure, the use of spherically shaped plasmas could allow the development of smaller, more economical and more stable fusion reactors. NSTX's attractiveness may be further enhanced by its ability to trap a high "bootstrap" electric current. This self-driven internal plasma current would reduce the power requirements of externally driven plasma currents required to heat and confine the plasma.

Upgrade 2012–2015

Vacuum vessel during the upgrade Technicians inside NSTXU vacuum vessel. (15645933543).jpg
Vacuum vessel during the upgrade

The $94 million [1] NSTX-U (Upgrade) [2] was completed in 2015. It doubles the toroidal field (to 1 Tesla), plasma current (to 2 MA) and heating power. It increases the pulse duration by a factor of five. [3] To achieve this the central stack (CS) solenoid was widened, [4] and an OH coil, inner poloidal coils, and a 2nd neutral-ion beam line were added. [5] This upgrade consisted of a copper coil installation, not a superconducting coil.

Poloidal coil problem 2016 and Recovery 2016–present

The NSTX-U (Upgrade) was stopped in late 2016 just after its update, due to a failure of one its poloidal coils. [5] The NSTX had been shut down since 2012 and only returned for 10 weeks at the end of 2016 just after it was updated. The origin of this failure is partly attributed to a non-compliance of the chilled copper winding, the manufacture of which had been sub-contracted. After a diagnostic phase requiring the complete dismantling of the device and coils, evaluation of the design, and a redesign of major components including the six inner poloidal coils, [5] [6] a restarting plan was adopted in March 2018, with reactivation scheduled for the end of 2020, [7] though this was later pushed back to 2022. [8] As of 2022, the restart was still delayed due to an insulation problem between the central solenoid and the coils around it. [9]

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 aim to use stellarators to achieve controlled nuclear fusion. It is one of many types of magnetic confinement fusion devices, the most common being the tokamak. The name "stellarator" 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">Princeton Plasma Physics Laboratory</span> National laboratory for plasma physics and nuclear fusion science at Princeton, New Jersey

Princeton Plasma Physics Laboratory (PPPL) is a United States Department of Energy national laboratory for plasma physics and nuclear fusion science. Its primary mission is research into and development of fusion as an energy source. It is known for the development of the stellarator and tokamak designs, along with numerous fundamental advances in plasma physics and the exploration of many other plasma confinement concepts.

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">Reversed field pinch</span> Magnetic field plasma confinement device

A reversed-field pinch (RFP) is a device used to produce and contain near-thermonuclear plasmas. It is a toroidal pinch that uses a unique magnetic field configuration as a scheme to magnetically confine a plasma, primarily to study magnetic confinement fusion. Its magnetic geometry is somewhat different from that of a tokamak. As one moves out radially, the portion of the magnetic field pointing toroidally reverses its direction, giving rise to the term reversed field. This configuration can be sustained with comparatively lower fields than that of a tokamak of similar power density. One of the disadvantages of this configuration is that it tends to be more susceptible to non-linear effects and turbulence. This makes it a useful system for studying non-ideal (resistive) magnetohydrodynamics. RFPs are also used in studying astrophysical plasmas, which share many common features.

<span class="mw-page-title-main">Tokamak Fusion Test Reactor</span> Former experimental tokamak at Princeton Plasma Physics Laboratory

The Tokamak Fusion Test Reactor (TFTR) was an experimental tokamak built at Princeton Plasma Physics Laboratory (PPPL) circa 1980 and entering service in 1982. TFTR was designed with the explicit goal of reaching scientific breakeven, the point where the heat being released from the fusion reactions in the plasma is equal or greater than the heating being supplied to the plasma by external devices to warm it up.

<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">Spheromak</span>

A spheromak is an arrangement of plasma formed into a toroidal shape similar to a smoke ring. The spheromak contains large internal electric currents and their associated magnetic fields arranged so the magnetohydrodynamic forces within the spheromak are nearly balanced, resulting in long-lived (microsecond) confinement times without external fields. Spheromaks belong to a type of plasma configuration referred to as the compact toroids. A spheromak can be made and sustained using magnetic flux injection, leading to a dynomak.

The Small Tight Aspect Ratio Tokamak, or START was a nuclear fusion experiment that used magnetic confinement to hold plasma. START was the first full-sized machine to use the spherical tokamak design, which aimed to greatly reduce the aspect ratio of the traditional tokamak design.

<span class="mw-page-title-main">Madison Symmetric Torus</span>

The Madison Symmetric Torus (MST) is a reversed field pinch (RFP) physics experiment with applications to both fusion energy research and astrophysical plasmas.

<span class="mw-page-title-main">Alcator C-Mod</span> Tokamak at MIT

Alcator C-Mod was a tokamak that operated between 1991 and 2016 at the Massachusetts Institute of Technology (MIT) Plasma Science and Fusion Center (PSFC). Notable for its high toroidal magnetic field, Alcator C-Mod holds the world record for volume averaged plasma pressure in a magnetically confined fusion device. Until its shutdown in 2016, it was one of the major fusion research facilities in the United States.

<span class="mw-page-title-main">National Compact Stellarator Experiment</span>

The National Compact Stellarator Experiment, NCSX in short, was a magnetic fusion energy experiment based on the stellarator design being constructed at the Princeton Plasma Physics Laboratory (PPPL).

The beta of a plasma, symbolized by β, is the ratio of the plasma pressure (p = nkBT) to the magnetic pressure (pmag = B2/2μ0). The term is commonly used in studies of the Sun and Earth's magnetic field, and in the field of fusion power designs.

Compact toroids are a class of toroidal plasma configurations that are self-stable, and whose configuration does not require magnet coils running through the center of the toroid. They are studied primarily in the field of fusion energy, where the lack of complex magnets and a simple geometry may allow the construction of dramatically simpler and less expensive fusion reactors.

<span class="mw-page-title-main">Spherical tokamak</span> Fusion power device

A spherical tokamak is a type of fusion power device based on the tokamak principle. It is notable for its very narrow profile, or aspect ratio. A traditional tokamak has a toroidal confinement area that gives it an overall shape similar to a donut, complete with a large hole in the middle. The spherical tokamak reduces the size of the hole as much as possible, resulting in a plasma shape that is almost spherical, often compared to a cored apple. The spherical tokamak is sometimes referred to as a spherical torus and often shortened to ST.

<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.

The Tokamak Physics Experiment (TPX) was a plasma physics experiment that was designed but not built. It was designed by an inter-organizational team in the USA led by Princeton Plasma Physics Laboratory. The experiment was designed to test theories about how Tokamaks would behave in a high-performance, steady-state regime.

Rajesh Maingi is a physicist known for his expertise in the physics of plasma edges and program leadership in the field of fusion energy. He is currently the head of Tokamak Experimental Sciences at the U.S. Department of Energy's (DOE) Princeton Plasma Physics Laboratory (PPPL). He is a Fellow of both the American Physical Society and the American Nuclear Society and has chaired or co-chaired numerous national and international conferences.

References

  1. "NSTX-U Press Kit". Princeton Plasma Physics Lab.
  2. "Diagram of NSTX-U changes". Archived from the original on 2021-03-23. Retrieved 2020-11-14.
  3. The Role of the Spherical Tokamak in the U.S. Fusion Energy Sciences Program Menard, 2012
  4. "PPPL to launch major upgrade of key fusion energy test facility". Princeton Plasma Physics Lab. Jan 2012. Archived from the original on 2020-12-20. Retrieved 2015-12-12.
  5. 1 2 3 "Overview of the NSTX-U Recovery Project Physics and Engineering Design" (PDF). S. P. Gerhardt, et al. Archived from the original (PDF) on 2022-12-21. Retrieved 2019-09-07.
  6. "NSTX-U recovery plan: Environmental Evaluation Notification Form" (PDF). NSTX-U recovery project. August 2017. $65,000,000 ... * Redesign and Replace the Inner Poloidal Field (PF) Coils : The six PF- I magnet coils would be replaced with new coils or improved design: they would be mandrel-less, have no joggles, and no braze joints. * Redesign and Replace Polar Regions of NSTX-U : The top and bottom of the NSTX-U device would be redesigned with numerous design improvements. All single 0-ring seals would be replaced by double 0-rings or a metallic structure, the PF-1c vacuum interface would be made more robust, one of either the upper or lower ceramic insulators would be eliminated, and the PF-lb coil supports would be thermally isolated from the vessel. * Redesign and Replace Plasma Facing Components.
  7. "[1st] Review of NSTX-U Recovery plans notes progress and outlines challenges" (PDF). Princeton Plasma Physics Lab. 12 Feb 2018.
  8. Cho, Adrian (2020-02-06). "After decades of decline, the U.S. national fusion lab seeks a rebirth". Science | AAAS. Retrieved 2020-02-07.
  9. Gerhardt, Stefan (2022-03-16). "Team Meeting 3/16/2022" (PDF). NSTX-U Team Meeting. p. 4. Retrieved 2022-10-14.

Sources