See also
- ADITYA (tokamak)
- Megaproject
- Fusion for Energy, EU organisation managing ITER
- Nuclear power in India
| Steady State Superconducting Tokamak | |
|---|---|
| Device type | Tokamak |
| Location | Gandhinagar, India |
| Affiliation | Department of Atomic Energy |
| Technical specifications | |
| Major radius | 1.1 m (3 ft 7 in) |
| Minor radius | 0.2 m (7.9 in) |
| Magnetic field | 3 T (30,000 G) |
| History | |
| Year(s) of operation | 2005–present |
| Links | |
| Website | www |
SST-1 (or Steady State Superconducting Tokamak) is a plasma confinement experimental device in the Institute for Plasma Research (IPR), an autonomous research institute under Department of Atomic Energy, India. It belongs to a new generation of tokamaks with the major objective being steady state operation of an advanced configuration ('D' Shaped) plasma. It has been designed as a medium-sized tokamak with superconducting magnets.
The SST-1 project helped India become capable of conceptualizing and making a fully functional fusion based reactor device. The SST-1 System is housed in Institute for Plasma Research, Gandhinagar. The SST-1 mission has been chaired by Indian plasma physicists Prof. Y.C. Saxena, Dr. Chenna Reddy, and is headed by Dr. Subrata Pradhan.
Next stage of the SST-1 mission, the SST-2, dubbed as 'DEMO', has already been initiated. [1]
The first talks about SST Mission started in 1994. The technical details and mechanical drawings of the system were finalized in 2001. The machine was fabricated by 2005. Godrej-Boyce Pvt. Ltd. played a crucial role in fabrication of the SST-1 coils. The assembly of SST-1 convinced the top brass of Indian bureaucracy to give a green flag to the claim of Indian physicists to join the ITER program [See Info Box]. On 17 August 2005, PM Sayeed, then India's power minister informed the Rajya Sabha about India's claim to join ITER. [2] A team from ITER, France visited the SST-1 mission control housed in Institute for Plasma Research to see the advances Indian scientists had made. Finally on 6 December 2005, India was officially accepted as a full partner of the ITER project. [3] To improve and modify some of the components, the SST-1 machine was subsequently disassembled. The improved version of the machine was completely assembled by January 2012.
It was fully commissioned in 2013. And by 2015, produces repeatable plasma discharges up to ~ 500 ms with plasma currents in excess of 75000 A at a central field of 1.5 T. [4] "SST-1 is also the only tokamak in the world with superconducting toroidal field magnets operating in two-phase helium instead of supercritical helium in a cryo-stable manner, thereby demonstrating reduced cold helium consumption. " [4] [5]
As of Dec 2015 it is having upgrades including to the plasma facing components to allow longer pulses. [5] [ needs update ]
Traditionally the tokamaks have operated with a `transformer' action- with plasma acting as a secondary, thus having the vital `self-generated' magnetic field on top of the `externally generated' (toroidal and equilibrium) fields. This is a pretty good scheme in which creation, current-drive and heating are neatly integrated and remained a choice of the fusion community for many years until the stage came to heat the plasma to multi-keV temperatures. Heating was then accomplished separately by radio frequency (RF) waves and/or energetic neutral beam injection (NBI).
Subsequently, excellent control got established on tokamak plasma performance by controlling the plasma-wall interaction processes at the plasma boundary so the plasma duration was limited primarily by the `transformer pulse length'. However, for relevance to future power reactors it is essential to operate these devices in a steady state mode. The very idea of steady state operation presents a series of physics and technology challenges. For example, the excellent plasma performance which was accomplished earlier, was with the surrounding material wall acting as a good 'pump' of particles, a fact which may not be true in steady state.
So one has to try and accomplish an equally good performance in presence of a possibly `saturated' wall. Secondly, a host of engineering and technical considerations spring up. The magnets must be superconducting type, otherwise the power dissipation in conventional (resistive) types can reach uneconomical levels. They have to be specially designed to remain superconducting in spite of their proximity to the other `warm' objects (like vacuum vessel etc.). The heat and particle exhaust must be handled in steady state with specialized tiles and active cooling. The advanced, so-called double null divertor plasma configuration has to be maintained through efficient feedback control avoiding plasma disruptions over long discharge durations. [6]
| Toroidal field, Bθ | 3 T |
| Plasma current, IP | 0.22 MA |
| Major radius, R0 | 1.1 m |
| Minor radius, a | 0.2 m |
| Aspect ratio, R/a | 5.5 |
| Elongation, κ | <=1.9 |
| Triangularity, δ | <=0.8 |
| Ion cyclotron resonance heating (ICRH) | 1 MW |
| Lower hybrid current drive (LHCD) | 1 MW |
| Neutral beam injection (NBI) | 1 MW |
| Discharge Duration | 1000 s |
| Configuration | Double-null divertor |
SST-1 will feature many new plasma diagnostic devices, many of which are being used for the first time in fusion research in India. Some of the novel plasma diagnostics devices incorporated in SST-1 are:
Almost all of the diagnostic devices installed on SST-1 are indigenous and are designed and developed by Diagnostics Group of Institute for Plasma Research. This group is the only group working on plasma diagnostics and related technologies in Indian Subcontinent.
The next stage of SST mission, the SST-2 fusion reactor, dubbed as 'DEMO' among Indian scientific circles has already been conceived. A group of eminent scientists from Institute for Plasma Research is working towards making of a full-fledged fusion reactor capable of producing electricity. Many new features like D-T plasma, Test Blanket Module, Biological shielding and an improved divertor will be incorporated in SST-2. SST-2 will also be built in the Indian state of Gujarat. The land acquisition and other basic formalities have been completed for the same.
Other designs of fusion reactor are DEMO, [7] Wendelstein 7-X, [8] NIF, [9] HiPER, [10] JET (precursor to ITER), [11] and MAST. [12]
A tokamak is a device which uses a powerful magnetic field to confine plasma in the shape of a torus. The tokamak is one of several types of magnetic confinement devices being developed to produce controlled thermonuclear fusion power. As of 2016, it was the leading candidate for a practical fusion reactor.
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 to date, no design has produced more fusion power output than the electrical power input.
This timeline of nuclear fusion is an incomplete chronological summary of significant events in the study and use of nuclear fusion.
The Joint European Torus, or JET, is an operational magnetically confined plasma physics experiment, located at Culham Centre for Fusion Energy in Oxfordshire, UK. Based on a tokamak design, the fusion research facility is a joint European project with a main purpose of opening the way to future nuclear fusion grid energy. At the time of its design JET was larger than any comparable machine.
ITER is an international nuclear fusion research and engineering megaproject aimed at creating energy by replicating, on Earth, the fusion processes of the Sun. Upon completion of construction of the main reactor and first plasma, planned for late 2025, it will be the world's largest magnetic confinement plasma physics experiment and the largest experimental tokamak nuclear fusion reactor. It is being built next to the Cadarache facility in southern France. ITER will be the largest of more than 100 fusion reactors built since the 1950s, with ten times the plasma volume of any other tokamak operating today.
A reversed-field pinch (RFP) is a device used to produce and contain near-thermonuclear plasmas. It is a toroidal pinch which 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 the more common 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.
The T-15 is a Russian nuclear fusion research reactor located at the Kurchatov Institute, which is based on the (Soviet-invented) tokamak design. It was the first industrial prototype fusion reactor to use superconducting magnets to control the plasma. These enormous superconducting magnets confined the plasma the reactor produced, but failed to sustain it for more than just a few seconds. Despite not being immediately applicable, this new technological advancement proved to the USSR that they were on the right path. In the original shape, a toroidal chamber design, it had a major radius of 2.43 m and minor radius 0.7 m.
Magnetic confinement fusion 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 fusion energy research, along with inertial confinement fusion. The magnetic approach began in the 1940s and absorbed the majority of subsequent development.
The Experimental Advanced Superconducting Tokamak (EAST), internal designation HT-7U, is an experimental superconducting tokamak magnetic fusion energy reactor in Hefei, China. The Hefei Institutes of Physical Science is conducting the experiment for the Chinese Academy of Sciences. It has operated since 2006.
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, for Upgrade.
The KSTAR is a magnetic fusion device at the Korea Institute of Fusion Energy in Daejeon, South Korea. It is intended to study aspects of magnetic fusion energy which will be pertinent to the ITER fusion project as part of that country's contribution to the ITER effort. The project was approved in 1995 but construction was delayed by the East Asian financial crisis which weakened the South Korean economy considerably; however the construction phase of the project was completed on September 14, 2007. The first plasma was achieved in June 2008.
ASDEX Upgrade is a divertor tokamak, that went into operation at the Max-Planck-Institut für Plasmaphysik, Garching in 1991. At present, it is Germany's second largest fusion experiment after stellarator Wendelstein 7-X.
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.
WEST, Tungsten Environment in Steady-state Tokamak, is a French tokamak that originally began operating as Tore Supra after the discontinuation of TFR and of Petula. The original name came from the words torus and superconductor, as Tore Supra was for a long time the only tokamak of this size with superconducting toroidal magnets, allowing the creation of a strong permanent toroidal magnetic field. After a major upgrade to install tungsten walls and a divertor, the tokamak was renamed WEST.
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.
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 with a cored apple. The spherical tokamak is sometimes referred to as a spherical torus and often shortened to ST.
In nuclear fusion power research, a divertor is a device within a tokamak or a stellarator that allows the online removal of waste material from the plasma while the reactor is still operating. This allows control over the buildup of fusion products in the fuel, and removes impurities in the plasma that have entered into it from the vessel lining.
The ARC fusion reactor is a design for a compact fusion reactor developed by the Massachusetts Institute of Technology (MIT) Plasma Science and Fusion Center (PSFC). ARC aims to achieve an engineering breakeven of three. The key technical innovation is to use high-temperature superconducting magnets in place of ITER's low-temperature superconducting magnets. The proposed device would be about half the diameter of the ITER reactor and cheaper to build.
The history of nuclear fusion began early in the 20th century as an inquiry into how stars powered themselves and expanded to incorporate a broad inquiry into the nature of matter and energy, as potential applications expanded to include warfare, energy production and rocket propulsion.
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