What is Magnetic Confinement Fusion?

Magnetic confinement fusion refers to the use of a special form of magnetic field to confine the ultra-high temperature plasma composed of light atomic nuclei such as deuterium and tritium and free electrons in a thermonuclear reaction state in a limited volume, so that it can control a large number of nuclear fusion reactions. To release energy. Research on magnetically constrained fusion in China began in the 1950s and 1960s. The Chinese Academy of Sciences Institute of Physics first built a linear discharge device and two angular pinch devices. The Institute of Atomic Energy has built a magnetic mirror "Xiaolong" device. Peking University, Fudan University and Tsinghua University have carried out relevant basic research.

Magnetic confinement fusion refers to the use of a special form of magnetic field to confine the ultra-high temperature plasma composed of light atomic nuclei such as deuterium and tritium and free electrons in a thermonuclear reaction state in a limited volume, so that it can control a large number of nuclear fusion reactions. To release energy ^[1]

After the establishment of the Northeast Institute of Technical Physics in 1962, a Z pinch device, an angular pinch device, and an ion source were built, and a steady-state magnetic mirror was designed.

Although China's nuclear fusion energy research has experienced difficulties for a long period of time, it has always been able to adhere to stable and gradual development, and has built two professional institutes combining science and technology, namely the nuclear industry group company.

There are two main methods of controllable nuclear fusion restraint, one is inertial restraint and the other is magnetic restraint. Inertial confinement refers to a method of using particles' inertia to constrain the particles themselves to achieve nuclear fusion reactions. The basic idea is: use the energy provided by the driver to make the nuclear fusion fuel (deuterium, tritium) in the target pellets form a plasma. In a very short period of time, due to the inertia of these plasma particles, they can not fly to the surroundings. The explosive fusion is compressed to a high temperature and high density state, so that a nuclear fusion reaction occurs. The research is mainly in the US National Ignition Facility (NIF) and China's Shenguang-III mainframe device ^[2]

HT-7 Magnetic confinement fusion HT-7

HT-7 device diagram

In the early 1990s, the Plasma Institute of the Chinese Academy of Sciences used the former Soviet Union's original T-15 device valued at about 15 million dollars to make a major transformation, making it a more advanced and complete superconducting tokamak-HT -7. Its main research objectives are to obtain and study long-pulse or quasi-steady-state high-temperature plasmas, test and develop related engineering technologies, and provide engineering technology and physical foundation for future steady-state advanced tokamak fusion reactors.

From December 1994 to March 1995, the HT-7 successfully performed the joint engineering adjustment for the first time, and received the first plasma on December 28, 1994. In 1998, the State Council's Science and Education Leading Group approved the establishment of the national Ninth Five-Year Plan major scientific project HT-7U. The HT-7 also partially undertakes the preliminary experimental tasks of the next-generation device HT-7U.

HT-7 is a large-scale experimental system, which includes the body of the HT-7 superconducting tokamak, a large ultra-high vacuum system, a large-scale computer control and data acquisition and processing system, a large-scale high-power pulse power supply and its loop system, and a national scale The largest cryogenic liquid helium system, MW-level low clutter current drive and RF wave heating systems, and dozens of complex diagnostic measurement systems. Over the past few years, the HT-7 superconducting tokamak device has undergone continuous transformation and has carried out more than a dozen rounds of experimental operations. It has achieved a number of scientific research results and has a certain international influence. In order to achieve the high-power, steady-state operation of the HT-7 superconducting tokamak device, in 2001, researchers made several major modifications to the HT-7's experimental system:

(1) Steady-state power supply and control of polar field;

(2) The use of vanadium steel to achieve significant improvement in longitudinal field waviness under steady state conditions;

(3) 1MW steady state low clutter current drive system;

(4) High-performance water-cooled graphite limiter and particle exclusion system;

(5) New RF antenna feed system;

(6) Real-time continuous collection system for massive data;

(7) Several advanced plasma diagnostic systems.

Physically, HT-7 is closely researched on the world's leading topic of steady-state high-constrained plasma operation. The experiments carried out for this purpose are as follows:

(1) Low clutter current drive and improvement constraints;

(2) Ion Bernstein wave heating and improving restraint;

(3) Research on boundary turbulence and transport;

(4) Fine distribution control of plasma parameters;

(5) Advanced wall treatment;

(6) Steady-state operation and control.

With the continuous deepening of physical experiments, the winter experiment in 2001 has made significant progress and obtained many research results:

(1) Achieving a repeatable high-temperature plasma discharge with an electron temperature of more than five million degrees, a center density greater than 20 seconds, and low-noise drive;

(2) Achieving a high-parameter plasmon discharge that is greater than 10 seconds, has an electron temperature exceeding 10 million degrees, and has a center density greater than this. This is the world s second high-parameter quasi-steady-state plasma with a discharge length of 1,000 times the energy constraint time;

(3) Achieving high-constraint steady-state operation with a discharge pulse length greater than 100 times the energy confinement time and an electron temperature of 20 million degrees under the synergistic action of ion Burns wave and low clutter;

(4) The maximum electron temperature exceeds 30 million degrees.

The main physical and technical indicators achieved by the HT-7 superconducting tokamak are:

(1) Plasma parameters: discharge time 20 seconds, electron temperature> 30 million degrees, electron density, plasma current 240 kA;

(2) Device operating parameters: magnetic field strength 2.2 Tesla, background vacuum;

(3) Low clutter system index: the maximum injected power is 700 kW, the loop voltage is reduced to 0, and the transformer is reverse-charged;

(4) Ion cyclotron heating and IBW indicators: the maximum injection power is 330 kilowatts, and the plasma electron temperature and ion temperature increase significantly;

(5) Plasma and wall interaction: RF cleaning and RF boronization and silicidation have obvious effects, and the effective Zeff is close to 1;

(6) Diagnostic technology and achieved indicators: 35 types of diagnostics, more than 400 diagnostic signals;

(7) Feeding technology: Collaborative experiments of projectile injection and IBW, found that the core constraint is improved; Laval nozzle experiments have obtained preliminary results;

(8) Plasma control: Multi-variable control, plasma current, displacement feedback, flexible adjustment of plasma parameters, high discharge repetition rate.

The above indicators fully demonstrate that the HT-7 superconducting tokamak device has entered the ranks of advanced devices that can perform plasma physics research under high-parameter steady-state conditions.

EAST Magnetic confinement fusion EAST

In order to study the steady state and advanced operation of plasma under high-parameter conditions near the core, and to further explore the engineering and physical issues of achieving fusion energy, the Institute of Plasma Physics of the Chinese Academy of Sciences has built a superconducting tokamak HT-7 , Proposed the "HT-7U full superconducting non-circular cross-section Tokamak device construction" plan, later renamed EAST. EAST is spelled with the initials of the words "Experimental", "Advanced", "Superconducting", and Tokamak. Its Chinese meaning is "Advanced Experimental Superconducting Tokamak", and it also has The meaning of "Orient".

EAST device

The EAST device is a fully superconducting tokamak device designed and developed by our country. Its main technical characteristics and indicators are: 16 large "D" shaped superconducting longitudinal field magnets will produce longitudinal field strength; 12 large polar field superconductors Magnets can provide a change in magnetic flux 10 volt-seconds; through these polar-field superconducting magnets, a plasma current of 1 million amps can be generated; the duration will reach 1000 seconds, and the temperature will exceed 1 under high-power heating conditions 100 million degrees.

The main part of the EAST device is 11 meters high, 8 meters in diameter, and weighs 400 tons. It consists of six major components: an ultra-high vacuum chamber, a vertical field coil, a polar field coil, internal and external cold screens, an external vacuum dewar, and a support system. Its experimental operation requires large-scale low-temperature helium cooling, large high-power pulsed power supplies, large superconductors, large-scale computer control and data acquisition processing, megawatt-level low-clutter current drive and RF wave heating, large ultra-high vacuum, and many advanced System support such as diagnostic measurement. The subject covers a wide range of subjects, and the technology is difficult. Many key technologies have no international experience. In particular, EAST operation requires extreme environments such as ultra-high current, ultra-strong magnetic field, ultra-high temperature, ultra-low temperature, and ultra-high vacuum. From the core high temperature of 100 million degrees to the low temperature of 269 degrees in the coil, it proposes the device design, manufacturing process and materials. It is extremely demanding.

Tokamak schematic

EAST is not only a fully superconducting tokamak (pictured to the right is a schematic diagram of tokamak), but also has a large elongated non-circular section that will improve the confinement of the plasma.

The surface plasma configuration, its completion will effectively promote the development of magnetically constrained nuclear fusion in China. During the period of 10-15 years after the installation of the device, it is possible to carry out exploratory experimental research on the cutting-edge physical problems of constructing a steady state-of-the-art Tokamak nuclear fusion reactor.

Although the size and radius of EAST is only 1/3 and 1/4 of the International Thermonuclear Fusion Experimental Reactor (ie ITER), its configuration is similar to that of ITER, and it was put into operation 10-15 years earlier than ITER. EAST is an important experimental platform for high-parameter near-core and scientific problems of steady-state advanced plasma operation. It will become the most important international physics experimental base for Tokamak for steady-state deflectors before ITER ^[3] .