Stellarator
A stellarator is a toroidal system for the magnetic confinement of a hot plasma with the aim of generating energy through nuclear fusion (see fusion by means of magnetic confinement and nuclear fusion reactor ). The name of this fusion concept should be reminiscent of the nuclear fusion as an energy source of the stars ( Latin stella "star").
A purely toroidal magnetic field cannot completely trap charged particles. The stellarator solves this problem with a complex, non-rotationally symmetrical magnetic field geometry. In the alternative concept of the tokamak , complete confinement is achieved by an electric current flowing in the fusion plasma .
Theoretical basics and differentiation from the tokamak
Magnetic confinement
A magnetic field can initially confine charged particles in two directions by forcing them on helical trajectories around the magnetic field lines ( gyration ) and they can only move freely along the direction of the field. For three-dimensional inclusion, the field is bent into a torus in which all field lines are closed in a circle. In doing so, however, the magnetic field inevitably becomes stronger towards the center than at the edge. This field gradient drives the particles out of the torus perpendicular to the direction of the magnetic field (torus drift). This is why the magnetic field has to be twisted as well so that the particles do not drift permanently in one direction on their course along a field line, but instead compensate for the drifts at different locations.
Stellarator
In the stellarator, unlike the tokamak, the twisted magnetic field is generated entirely by external current-carrying coils . Theoretically, it can be shown that a stellarator cannot be continuously rotationally symmetrical, i.e. it turns into itself when it rotates in a toroidal direction . The first stellarator concept had the shape of a horizontal figure 8, where two sections of opposite curvatures exist, the drifts of which should compensate each other. In modern stellarators, the stellarator field consists of a number of equal sections, the field periods , e.g. B. five in the Wendelstein 7-X , ten in the Large Helical Device (LHD), and thus has a discrete symmetry : The configuration only transitions into itself when rotated through an angle of 360 ° / in a toroidal direction. The so-called stellarator symmetry can also be present as a second symmetry: With this, a field period merges when it is rotated by 180 ° around a special axis.
Because of the discontinuous symmetry it can happen - unlike with the tokamak - that the magnetic field lines no longer run everywhere on nested surfaces, but behave chaotically in places . Since this has a negative effect on the confinement of the plasma, these areas ( ergodic areas and magnetic islands ) must be as small as possible.
The stellarator has two major advantages over the tokamak concept: Since no toroidal current has to be maintained in the plasma,
- instabilities associated with the plasma flow, which can lead to a breakdown of the plasma confinement, are avoided;
- a stellarator could later function as a power plant in continuous operation. With the tokamak concept, on the other hand, the question of how a current in the plasma can be maintained permanently is still the subject of current research.
These advantages are offset by the fact that the three-dimensional structure of the plasma fundamentally makes it more difficult to contain it in the hot state, so that an optimization of the magnetic field geometry is necessary. The coil system of a stellarator is also more complex than that of a tokamak. The tokamak and stellarator otherwise have many similar components; the technical requirements are also largely similar.
history
The concept of the stellarator was developed in 1951 by Lyman Spitzer in Princeton, USA, who initially proposed a configuration in which a torus was bent into the shape of a number 8. The experimental results of the successor, the “racetrack-shaped” Model-C, only showed insufficient plasma confinement. The experiments classified as secret against the backdrop of the Cold War were called Project Matterhorn. This is why the work that was continued after its publication in Garching near Munich in 1958 was given the name of the Bavarian mountain Wendelstein .
In principle experiments it could be shown that inaccuracies in the construction of the coils and the poor symmetry of the first arrangements were the reason for their poor containment. More symmetrical circular configurations have therefore been developed (the large radius R of the plasma in the torus is given below as the size measure): the classic Wendelstein 7-A stellarator (Garching, 1976-85, R = 2 m), the Heliotron-E , Kyoto (R = 2.2 m), the Torsatrons, Advanced Toroidal Facility ATF (1988, Oak Ridge, USA, R = 2 m) and Uragan (Charkow, Ukraine).
With the availability of heating methods that were independent of a current driven in the plasma, it was also possible for the first time, in contrast to a tokamak, to examine currentless “pure” stellarator plasmas. As expected, a whole class of (current-driven) instabilities could thereby be avoided, as well as the sudden loss of confinement due to current interruption. The plasma confinement of this first stellarator generation corresponded to that of tokamaks of comparable size at the temperatures attainable at that time. However, it turned out that the particle losses, which increase sharply with increasing temperature, and the horizontal displacement of the plasma that occurs with increasing pressure would not allow a fusion reactor of an economically acceptable size. Another conceptual disadvantage was the large forces, especially in places where magnetic coils come close or cross one another.
The breakthrough came with the concept of modular coils (Wobig and Rehker, 1972). The forces that occur can be better absorbed in these; Crossing coil systems are avoided. At the same time, there were more degrees of freedom to optimize the generated magnetic field with regard to the meanwhile further developed understanding of plasma transport (important with increasing temperature), equilibrium (important with increasing pressure) and instabilities (important with increasing temperature and density differences). In order to check the basic feasibility of the modular concept and the correctness of the theoretically obtained optimization criteria, the Wendelstein 7-AS project (for Advanced Stellarator) was proposed in Garching, which partly continued to use components of the predecessor Wendelstein 7-A and therefore only carried out a partial optimization depicted. The results of the experiment carried out between 1988 and 2002 met or even exceeded expectations in some respects. In the 1990s this led to a revitalization of global stellarator activities and to the construction of a series of small and medium-sized experiments that were intended to investigate partial aspects and other magnetic field configurations: u. a. H-1 (Canberra, Australia), TJ-II, (Madrid, Spain, R = 1.5 m), Heliotron-J (Kyoto, Japan) and the Helically Symmetric Experiment (HSX) (Madison, Wisconsin, R = 1 , 2 m) . The last two experiments mentioned already use the possibilities that arise with modular coils.
In Princeton (USA), the construction of the comparatively compact (R = 1.4 m) National Compact Stellarator Experiment began, which pursued an alternative strategy for optimizing the magnetic field. The current in the plasma should not be minimized here, so that a hybrid between tokamak (twisting of the magnetic field by current flow in the plasma) and stellarator (twisting of the magnetic field by external coils) is created. The construction of this quasi-toroidal-symmetrical stellarator was canceled by the US government in 2008 for cost reasons.
The conventional Heliotron Large Helical Device , which has been in operation in Nagoya (Japan) since 1998, has demonstrated the feasibility of a reactor-relevant large superconducting coil system and investigates the properties of stellarator plasmas in long-term operation (large radius = 3.6 m, small radius = 0.6 m, plasma volume V = 26 m 3 ).
The so-called HELIAS (HELIcally Advanced Stellarator) was developed on the basis of the Wendelstein stellarators in Garching and with the possibilities of the modular coils: a concept in which several optimization criteria for the magnetic field are met at the same time. This led to the design of the Wendelstein 7-X in 1990 , with which this concept is to be examined for its suitability for a fusion reactor. Construction began in Greifswald in 2001; the first plasma was generated in late 2015.
Stellarator types
Stellarators are mainly developed for fusion plasmas. In addition, stellarators are now also used for basic plasma physical investigations. Examples are the Columbia Non-Neutral Torus in New York and the Torsatron TJ-K (University of Stuttgart). However, fusion research focuses on the following types:
Classic stellarator
The coil system consists of 2 closed helical conductors in which the current flows in opposite directions in the adjacent conductors. This coil system is surrounded by further coils that generate the toroidal magnetic field component. A classic stellarator thus has two interlaced coil systems. This can place high demands on the mechanical stability, since the forces occurring at the crossover points of the coils must be absorbed by the construction (example: Wendelstein 7-A).
Heliotron, Torsatron
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Here the current flows in closed helical conductors (with a natural number ) in the same toroidal direction. The coils thus also jointly generate the toroidal magnetic field component. There is therefore no need for a toroidal coil system, but vertical field coils to compensate for the vertical field generated by the helical coils. In contrast to the classic stellarator, the two coil systems are not interlocked, the forces between the coils are therefore lower and can therefore be more easily absorbed by support structures. If you go in the toroidal direction, the cross-section of the plasma at = 2 corresponds to a rotating ellipse. Examples are the Large Helical Device (Japan), the Advanced Toroidal Facility (Oak Ridge, USA) and Uragan 3M (with = 3, Kharkov, Ukraine). The Heliotron-J experiment (Kyoto, Japan) is a hybrid of heliotron and heliac: the plasma axis winds around the helical central conductor like in a heliac, but the toroidal field coils are arranged like in a classic stellarator.
Heliac
In contrast to the heliotron or the classic stellarator, the plasma axis in the heliac does not form a circle, but rather winds around a central, circular magnetic field coil . The toroidal field coils surrounding the plasma follow this plasma axis. This creates a helically twisted component of the magnetic field in the reference system of the plasma. Vertical field coils are required to compensate for the vertical field. Heliacs offer good access to the plasma between the toroidal field coils. B. is advantageous for measurements. On the other hand, the plasma comes very close to the central conductor. Since neutron shielding and a breeding blanket are difficult to implement there, there is currently no concept for a fusion reactor based on the Heliac. Examples of heliacs: TJ-II (Madrid, Spain) and H-1 (Canberra, Australia).
Stellarator with modular coils
The ability to create a stellarator magnetic field with modular coils, i.e. H. Coils that are poloidally closed, but not flat , allow great freedom of design when choosing the magnetic field. At the same time, the magnetic forces in and between the coils can be better absorbed. Since no toroidal revolving coils would be required, much smaller superconducting coils could be used in a reactor, which would bring decisive technical and economic advantages. A stellarator configuration with modular coils allows almost any current distribution to be generated on a surface around the plasma. This results in more degrees of freedom to optimize the shape and strength of the magnetic field (example: Wendelstein 7-AS , Wendelstein 7-X ).
Optimized stellarator
Due to their three-dimensional geometry, stellarators offer a high degree of freedom of design. This freedom is used in modern stellarators in order to optimize the magnetic configuration with regard to certain criteria. The shape of the plasma is changed using numerical optimization algorithms until a set of previously established conditions is met that represent the requirements for the physical behavior of the stellarator (e.g. stability of the plasma against small disturbances, good inclusion of particles) . First the shape of the plasma is calculated and then in a second step the (modular) coil system that generates the required magnetic field. A more recent development are mixed forms between tokamak and stellarator, which have both a three-dimensional geometry and a total toroidal current.
The first examples of modular stellarators that follow such optimization criteria are Wendelstein 7-AS (optimized with regard to the Shafranov shift), Helically Symmetric Experiment HSX (see figure below) (partial aspects of optimization: quasi-helical symmetry, Madison, Wisconsin), NCSX (Partial aspects of optimization: quasi-toroidal symmetry, Princeton, USA, construction canceled) and Wendelstein 7-X (Greifswald).
Status of the stellarator development
The experimental results obtained with stellarators largely correspond to those of tokamaks and can therefore be traced back to the basic properties of a toroidal plasma confinement. This applies, for example, to heat and particle transport, such as B. is carried by instabilities, turbulence and currents in the plasma. The heating methods used, the diagnostics required and the important material issues for the first wall also largely coincide.
Wendelstein 7-AS and LHD have each shown with different concepts that - as with the tokamak - stable operation of a divertor is possible.
The experiments showed or confirmed the following essential differences to the tokamak:
- The convective transport, which increases sharply with temperature as a result of the three-dimensional magnetic field structure, is observed as expected; it should be reduced to an acceptable level by the stellarator optimization.
- Stellarators can work at significantly higher densities than tokamaks , since there is no risk of a power failure (with increasing density, the temperature decreases and the resistance increases with the same heating power - a plasma current then disappears). The higher density n has the advantage of increasing fusion power in the reactor of a given size ( P fusion ~ n 2 ). In addition, the experimentally observed plasma inclusion improves proportionally to the square root of n . The load on the wall is lower because of the simultaneous drop in temperature.
- In operation, stellarators behave comparatively moderately in the vicinity of operating limits (maximum density, maximum pressure). There are no abrupt instabilities that lead to heavy loads on the first wall, for example. Instead, the plasma may cool down on a moderate time scale (ultimately a consequence of the lack of plasma flow).
- A specific problem of stellarators could, however, be that, as a result of the drifts, an increasing concentration of impurities could build up inside in the long term, which would cool down the plasma through increased radiation. There are currently no meaningful long-pulse experiments.
Stellarator reactor concepts
Reactor concepts based on the stellarator confinement principle are similar in many technical aspects to those of tokamaks and benefit from their development. Continuous operation, however, avoids the mechanical alternating loads on the structural parts that occur in pulsed operation. On the other hand, the three-dimensionality of the magnetic field results in a high physical and technical complexity. Three concepts are currently (2016) being investigated.
Heliotron reactor
A Heliotron reactor would have the advantage of low forces between the superconducting coils and good accessibility through between the coils, e.g. for maintenance of the blanket . This is countered by the technical challenge of very large toroidal superconducting coils, as they have already been implemented at the LHD in a somewhat smaller size. The magnetic field structure leading to the divertor is created by the configuration at the corners of the approximately elliptical plasma cross-section and does not have to be generated by extra coils as in the tokamak. Correspondingly, the "helical divertor" with its baffle plates winds helically around the torus - in contrast to the tokamak, where the divertor rotates toroidally above or below. However, no overall concept is foreseeable for the classic heliotron, in which both sufficiently low heat transport and sufficient plasma pressure can be achieved with one and the same magnetic field configuration. Corresponding studies compensate for these disadvantages with the assumption of operation at relatively high density and very high magnetic fields (up to B = 12 T on the magnetic axis), the generation of which has yet to be demonstrated technically.
Reactors with modular coils
Both the US ARIES study and the HELIAS reactor examined in Europe provide for modular coils. With their moderate size, the coils could largely be realized with today's technology, (barely) transportable and could therefore be tested individually before assembly. However, in places where a strong curvature of the magnetic field is to be achieved on the inside of the torus, the coil and the plasma must come relatively close. In order to achieve a breeding blank and neutron shielding there, a minimum distance between the plasma and the coils of about 1.3 m is required, which could only be achieved in relatively large reactors. The resulting large wall surface would, however, also facilitate the dissipation of heat from the plasma and reduce the power density on the first wall and its load with neutrons . The high magnetic forces at points where the modular coils come close seem to be structurally controllable.
ARIES study
Based on the modular stellarator NCSX , a quasi-toroidal-symmetrical configuration with finite current, which was not implemented in the USA , a study was carried out on a comparatively compact ARIES stellarator reactor. Because of the desired small size, it is accepted that the plasma comes so close to the coils at narrow points that only a neutron shield, but no breeding blanket, could be accommodated there.
HELIAS reactor
The further development of the HELIAS concept applied in Wendelstein 7-X would lead to reactors with comparatively large radii (> 18 m). These are necessary in order to achieve a breeding blanket everywhere and to achieve ignition; both require a small radius of at least 1.8 m, if one conservatively assumes the technology available today and moderate magnetic fields (W = 5 T) for the superconducting coils. Such a reactor would be almost four times the size of the Wendelstein 7-X experiment.
literature
- "Helical Confinement Devices", Beidler, et al. in Fusion Physics, ed. by Kikuchi, Lackner, Tran, International Atomic Energy Agency Vienna 2012.
Individual evidence
- ↑ Peter Lobner: Return of the Stellarator ( in English ) August 30, 2017.
- ↑ historical summary of the Garching work in: Grieger G et al. 1985 Nucl. Fusion 25 1231-42
- ↑ H. Wobig, S. Rehker: A Stellarator coil system without helical windings . In: Proceedings of the 7th Symposium on Fusion Technology . Grenoble, France 1972, p. 345-353 .
- ↑ TJ-K
- ↑ Wendelstein 7-A at the IPP
- ↑ Heliotron-J ( Memento of the original from October 5, 2013 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice.
- ^ TJ-II, Madrid
- ↑ H-1, Canberra ( Memento of the original from April 9, 2013 in the Internet Archive ) Info: The archive link was inserted automatically and has not yet been checked. Please check the original and archive link according to the instructions and then remove this notice.
- ↑ Sagara, A., et al .; Fusion Eng. Of. 81 (2006) 2703-2712.
- ↑ Najmabadi, F., et al., Fusion Sci. Technol. 54 (3) (2008) 655-672
- ↑ Beidler, C., et al., Nucl. Fusion 43 (2003) 889-898.
