Patent Number: 043022847
Section: summary

This invention relates generally to plasma devices and particularly to the stabilization of toroidal fusion devices. More particularly, the present invention relates to the combination of helical magnetic field and a poloidal magnetic field for the stabilization of such devices. Toroidal plasma devices are devices in which plasma is created in a toroidal space and is confined therein by appropriate confining fields. Such devices are useful in the study and analysis of plasmas and particularly in the generation, confinement, study and analysis of hydrogenous plasmas. Such devices are useful in respect to plasma devices for the reaction of deuterium and tritium, with the production of high energy neutrons as reaction products. The present invention finds particular utility in such devices and their applications, including experimental devices and the use thereof in experimentation and investigation in respect to toroidal plasma devices. The problems in fusion devices are largely in heating the plasma to a high enough temperature to enable the desired reactions to occur and to confine the heated plasma for a time long enough to release energy in excess of that required to produce the reactions. The present invention is directed to the confinement of such plasma. A number of toroidal plasma devices have been suggested and built. These include the tokamak, the stellarator, and the reversed field pinch. In such devices, gas is confined in a toroidal confinement vessel and is heated to form a plasma which is generally held away from the walls of the confinement vessel by appropriate fields. In tokamak devices, a toroidal plasma current is produced as by a transformer with the toroidal confined gas acting as the secondary and with the primary being a central solenoid. Upon creation or extinction of the magnetic field produced by current in the solenoid, a toroidal electric field is produced to ionize the gas and drive plasma current around the torus. The pinch effect of the flowing current causes the charged plasma particles to be urged toward the center of the plasma current. However, the plasma current by itself is unstable and some of the plasma would strike the confinement vessel, hence cooling the plasma and hampering any reaction. For this reason, the tokamak also includes a toroidal field coil disposed around the confinement vessel to produce a very large toroidal magnetic field. The interaction of the toroidal magnetic field with the poloidal magnetic field produced by the plasma current produces a relatively stable plasma confinement. In stellarators the confinement is by magnetic fields produced by external coils and does not rely upon plasma current. In stellarators a toroidal field coil, like that of the tokamak, provides a relatively large toroidal magnetic field in which the plasma is created. In addition to the toroidal magnetic field, a helical field is produced by coils helically disposed about the toroidal confinement vessel. The combination of the toroidal magnetic field with the helical field produces a net twisted magnetic field providing relative stability to the plasma device. Helical coils and toroidal field coils can be combined as in the torsatron device. The difficulty with the stellarators has been the problem of producing the plasma in devices of reasonable size and in providing the very large magnetic field required. In the reversed field pinch confinement is achieved by trapping a toroidal field in a pinching plasma and inducing a toroidal field of the opposite sign between the plasma and the wall. The device of the present invention has certain aspects in common with the prior devices, such as the tokamak and the stellarator, but is generically different, particularly in the absence of heavy toroidal field coils. In accordance with the present invention, stability is achieved by the combination of the poloidal magnetic field produced by plasma current and the helical magnetic field produced by helical windings. The helical field superposed on the poloidal field produces a translational transform whereby the flux lines become helical and form twisted flux surfaces. In order to be magnetohydrodynamically (MHD) stable, toroidal plasma devices must satisfy necessary conditions for the safety factor q, where q is defined as an average length traversed in the toroidal direction per unit poloidal angle of rotation of a magnetic field line on a flux surface, divided by the major radius of the torus, i.e.: ##EQU1## where z is the distance traversed in the toroidal direction, .theta. is the poloidal angle of displacement, and R is the major radius. A flux surface is defined as a surface on which the magnetic flux density has no component normal thereto. If r is the minor radius, then these conditions are: ##EQU2## must be large enough to satisfy the Mercier criterion. Tokamak devices and those stellarator devices which carry substantial plasma current generally satisfy condition (a) by operating with .vertline.q.vertline.&gt;1 throughout the plasma. In contrast, the reversed field pinch device operates with .vertline.q.vertline.&lt;1 throughout the plasma. A sufficiently large ##EQU3## an implied shear, is obtained in the reversed field pinch by having a reverse sign near the edge of the plasma. In the case of the reversed field pinch, the flux surfaces are axisymmetric and circular in cross section and q has a simple definition in terms of the toroidal magnetic field B.sub.T, the poloidal magnetic field B.sub.P, the major radius of the torus R, and the minor radius r; in particular ##EQU4## for the case of the circular reversed field pinch, as well as for the circular tokamak. Since B.sub.P is unidirectional, the reversal of q can only be obtained by a corresponding reversal in B.sub.T in this case. The reversed field pinch achieves this transiently over the time scale for magnetic flux diffusion by trapping a toroidal field in a pinching plasma and inducing a toroidal field of the opposite sign between the plasma and the wall. The reversed field pinch had the disadvantages that (1) the plasma must be created prior to the creation of the desired field configuration for confinement, and involves either fast field programming or a turbulent initial phase in which the plasma can contact the wall introducing impurities, and (2) the plasma must be resistive for the externally applied magnetic field to penetrate the plasma and produce the desired configuration. However, the lifetime of the plasma is determined by magnetic field diffusion which occurs on the same time scale as the penetration. Therefore, it is difficult to produce the configuration and maintain it over a substantial length of time. The reversed field pinch has the advantages of (1) relatively higher beta (.beta.), the ratio of the plasma pressure to magnetic pressure, than tokamaks and (2 ) efficient ohmic heating since the relatively low q operation allows a relatively larger plasma current and aspect ratio R/r of the torus. The present invention is a generically different device from previous ones. It generates the desired magnetic confinement configuration by currents in helical windings and plasma current, and operates with .vertline.q.vertline.&lt;1 and with a reversed q near the plasma edge, as does the reversed field pinch. However, in the present invention with non-circular flux surfaces, the safety factor q defined as an average on a flux surface can be finite even in the absence of a net toroidal field (averaged over a circle). A reversed q configuration can be set up with no net toroidal field outside the plasma. The invention also allows for the introduction of a small external net toroidal field generated by slightly unbalancing the positive and negative helical coil currents for purposes of shifting the q=0 point to the radial position which is optimal for stability of the plasma. In the present invention, the value of q near the center of the plasma is substantially due to poloidal and toroidal plasma currents. The poloidal plasma currents generate a net toroidal field within the plasma. This net field decreases toward the edge of the plasma and vanishes outside the plasma. Near the edge of the plasma, q tends to reverse sign because of the toroidal field from the helical coils averaged over a flux surface, this field being in opposition to the net toroidal field generated by the poloidal plasma current. The present invention achieves a q profile suitable for plasma stability at high beta and suitable for efficient ohmic heating in a configuration which is not limited by magnetic flux diffusion but which persists as long as the currents in the helical coils and in the plasma are maintained. The plasma current is induced by a central solenoid as in the usual tokamak case. The present invention does not depend on toroidicity to achieve stability of the plasma and can be operated as a high-aspect ratio torus, like a bicycle tire, to relax design constraints on the central induction coil for driving the plasma current and on blanket design in a reactor application. The present invention does not require toroidal field coils other than the helical coils. These helical coils are preferably operated to produce a zero or slight net toroidal field. The absence of a large net toroidal field relaxes interwinding forces and stresses. The present invention has a separatrix which bounds the plasma. This separatrix defines a closed surface within which closed and nested magnetic flux surfaces exist. The radial position of the separatrix increases with the plasma current, and the separatrix acts as a magnetic limiter to confine the plasma current channel away from the wall during the start-up of the discharge. A suitable magnetic configuration is achieved at the beginning of the discharge without the necessity of fast field programming or a turbulent transition to the desired state. The separatrix also facilitates the introduction of a divertor, which might be desirable to reduce impurities in the plasma. More particularly, in accordance with the present invention, a toroidal confinement vessel defines a toroidal space and confines gas therein. A central solenoid generates magnetic flux linking the toroidal space to produce a toroidal electric field which drives plasma current therein. A plurality of first windings are wound substantially helically around the vessel substantially equally spaced around its minor circumference. A plurality of second windings are wound substantially helically around the vessel substantially midway between successive first windings. Direct current is passed through the respective first and second windings in opposite directions with the current in the respective first and second windings equal or slightly unbalanced. The magnitude of the plasma current relative to the magnitude of the currents in the first and second windings produces a separatrix in the toroidal space, this separatrix defining a closed surface which limits and encloses a region within which closed and nested magnetic flux surfaces exist. The sense of rotation of the first and second windings and the direction of the plasma current produces a variation in the safety factor q with minor radius at any poloidal angle, whereby the sign of q reverses near the outer edge of the plasma, q being an average over a flux surface of the number of transits made around the torus in the toroidal direction by a magnetic flux line in making a single transit in the poloidal direction. The sign of q is determined by the sense of the direction in which the toroidal transit is made.