Tokamak Experimental Reactor Research Articles

On heat removal from the first-wall elements and warm toroidal field coils in the JUST-T tokamak

Heat removal from the structural elements of a compact tokamak with warm coils intended for the transmutation of minor actinides is an especially critical problem. This paper is concerned with the thermal processes on the first-wall elements and the toroidal field windings, ignoring the most stressed component of JUST-T, i.e. the blanket. The analysis has shown that the peak load on the first-wall elements in JUST-T does not exceed a similar load in the International Thermonuclear Experimental Reactor tokamak. The two-dimensional numerical code for calculation of the current distribution density, Ohmic heating and heat transfer due to heat conductivity and extraction of heat by the water flux has been developed.

Potential for D-3He Experiments in Next-Generation Tokamaks

In this paper the potential for D-{sup 3}He experiments in the proposed Compact Ignition Tokamak (CIT) and International Thermonuclear Experimental Reactor (ITER) tokamak test devices is examined. In CIT, an energy multiplication Q of {approximately}0.3 can be obtained with an injection power of {approximately}100 MW. Without modifications to ITER, except for the change of fuel, it is found that Q of the order of 0.3 to 0.5 can be obtained. Breakeven with D-{sup 3}He requires modification to the device to increase the elongation to 2.4, reduce the major radius to 5.6 m, and increase the magnetic field at the plasma from 4.9 to 5.6 T. Operation with a small amount of tritium seeding can reduce the auxiliary power required to achieve breakeven and leads to Q = 2 in an unmodified device.

Superconducting induction coil for a doublet tokamak experimental fusion power reactor

A superconducting induction coil system was designed for a noncircular tokamak experimental fusion power reactor based on the Doublet concept. This coil system is used for inducing and sustaining the plasma current during the thermonuclear burn. The coil must be cycled in about 2 seconds; thus, it is important to design it for low AC loss during pulsing operation. A braided superconducting cable capable of carrying 450 kA at a central field of 5.65 Tesla was chosen as the conductor. The coil system consists of a central solenoid and 12 outer coils, with a total of 157 series connected turns. The central solenoid has a single layer winding of 134 turns on a fiberglass epoxy spool. The outer coils are located so that the error field generated at the plasma location is limited to a few gauss. The winding is reinforced by prestressed stainless steel bands on the outer surface, so that there is no relative motion between the cable and the coil spool or insulation during pulsing.

Nuclear Design of the Blanket/Shield System for a Tokamak Experimental Power Reactor

The various options and trade-offs in the nuclear design of the blanket/ shield for a Tokamak Experimental Power Reactor (TEPR) are investigated. The TEPR size and cost are particularly sensitive to the blanket/shield thickness, $delta$/sub BS/, on the inner side of the torus. Radition damage to the components of the superconducting magnet and refrigeration power requirements set lower limits on $delta$/sub BS/. These limits are developed in terms of TEPR design parameters such as the wall loading, duty cycle, and frequency of magnet anneals. The study of the nuclear performance of various material compositions shows that mixtures of tungsten, or tantalum, or stainless-steel alloys and boron carbide require the smallest $delta$/sub BS/ for a given attenuation. This $delta$/sub BS/ has to be doubled if the low induced activation materials graphite and aluminum are used. The space problems are greatly eased in the Argonne National Laboratory ANL-TEPR reference design by using two separate segments of the blanket/shield. The inner segment occupies the region of the high magnetic field, uses very efficient attenuators (tungsten- or tantalum- or stainless-steel-boron carbide mixtures), and is only 1 m thick. The outer more » blanket/shield is 131 cm and consists of an optimized composition of stainless steel and boron carbide. For the design parameters of 0.2 MW/m$sup 2$ neutron wall loading and 50 percent duty cycle, the reactor components can operate satisfactorily up to (a) 10 yr for the stainless-steel first wall, (b) 10 yr for the superconductor composite after which magnet warmup becomes necessary, and (c) 30 yr for the Mylar insulation. Nuclear heat generation rates in the blanket/ shield and magnet are well within the practical limits for heat removal. « less