A Poynting's Theorem method is used for evaluating the volt-second consumption in a tokamak discharge. The method accurately identifies the inductive and resistive components of the volt-second consumption, and allows both quantities to be determined from magnetic measurements made outside the plasma. Only simple computational techniques are required. Application of the method to typical Doublet III nearcircular plasmas (R = 1.43 m, a = 0.44 m, b/a = 1.2) indicates that the flux at the plasma surface required to establish the current flat-top is 2.0 ± 0.2 V-s/MA. Approximately 40% of this flux is consumed in resistive dissipation. This division between resistive and inductive flux differs significantly from that obtained using an alternative data analysis method in which the resistive loss is evaluated at the plasma axis. The reasons for the difference are discussed.

The dependence of plasma energy confinement on minor radius, density and plasma current is described for Ohmically heated near-circular plasmas in Doublet III. A wide range of parameters is used for the study of scaling laws; the plasma minor radius defined by the flux surface in contact with limiter is varied by a factor of 2 (a = 44, 32 and 23 cm) , the line average plasma density, n̄e, is varied by a factor of 20 from 0.5 to 10 × 1013 cm−3 (n̄e R0/BT = 0.3 to 6 × 1014 cm−2·kG−1) and the plasma current, I, is varied by a factor of 6 from 120 to 718 kA. The range of the limiter safety factor, qL, is from 2 to 12. – For plasmas with a = 23 and 32 cm, the scaling law at low n̄e for the gross electron energy confinement time can be written as (s, cm) where qc = 2πa2BT/μ0IR0. For the 44-cm plasmas, is about 1.8 times less than predicted by this scaling, possibly owing to the change in limiter configuration and small plasma-wall separation and/or the aspect ratio change. At high n̄e, saturates and in many cases decreases with n̄e but increases with I in a classical-like manner. The dependence of on a is considerably weakened. The confinement behaviour can be explained by taking an ion thermal conductivity 2 to 7 times that given by Hinton-Hazeltine's neoclassical theory with a lumped-Zeff impurity model. Within this range the enhancement factor increases with a or a/R0. The electron thermal conductivity evaluated at half-temperature radius where most of the thermal insulation occurs sharply increases with average current density within that radius, but does not depend on a within the uncertainties of the measurements.

Neutral-beam current-drive experiments in the DIII-D tokamak with a single null poloidal divertor are described. A plasma current of 0.34 MA has been sustained by neutral beams alone, and the energy confinement is of //-mode quality. Poloidal p values reach 3.5 without disruption or coherent magnetic activity suggesting that these plasmas may be entering the second stability regime.PACS numbers: 52.55.Fa, 52.50.GjThe tokamak magnetic fusion configuration requires a toroidal current within the plasma. Generally this current is inductively coupled. Tokamaks can therefore only operate for finite-duration pulses. Also, the current concentrates in regions of high electrical conductivity (regions of high electron temperature) and thereby does not necessarily produce an optimum radial current profile. Numerous noninductive current-drive methods 1 have been proposed, including injection of electromagnetic waves and neutral beams. These methods could allow steady-state tokamak operation and optimization of the radial current profiles to possibly improve confinement and provide access to the second stability region, in which increasing plasma pressure increases plasma stability.The concept of neutral-beam current drive was proposed by Ohkawa 2 and the basic principle was demonstrated in the Culham Levitron. 3 First tokamak results were obtained in DITE 4 and subsequently in TFTR 5 and JET. 6 This paper presents new results from the DIII-D 7 tokamak in which the plasma current was sustained entirely by neutral beams from 1.5 s without assistance from the Ohmic-heating transformer. After Ohmic startup, the Ohmic-heating-coil current was held constant so that the plasma current could freely adjust. This technique provides a striking demonstration of neutral-beam current drive. The poloidal beta, fi p , reached 3.5, raising the possibility that the plasma is entering the second stability region, as described later.The DIII-D tokamak 7 was operated with a singlenull-divertor configuration having a 1.70-m major radius, 0.6-m minor radius, 1.75 vertical elongation, and 2.1-T toroidal magnetic field. The experiments were carried out with a helium plasma having a line-averaged density n e =2x 10 19 m ~3. Eight hydrogen neutral beams, 8 consisting of 52% neutral power at 75 keV, 30% at 37 keV, and 18% at 25 keV, were injected in the same direction as the plasma current. Four beams intersected the vacuum system axis at 47° and four beams intersected at 63°.Plasma parameters are shown in Fig. 1 as functions of time. Initially, a 0.22-MA Ohmic discharge was estab-lished without sawteeth, indicating an on-axis safety factor <7o>L At 1.1 s the Ohmic-heating-primary-coil current was held constant, so that without beam injection the plasma current decayed, as shown by the dashed line of Fig. 1(a). With 10 MW of absorbed neutralbeam injection [ Fig. 1(b)], the plasma current increased to 0.34 MA. During the period when the current was sustained, the loop voltage [ Fig. 1(c)] was zero, except for periodic voltage spikes associated with edgel...

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Using a neutral-beam injection power of 3.4 M W, volume-averaged toroidal betas of up to ⟨βT⟩ = 4.5% have been obtained in low-toroidal-field, low-qψ, vertically elongated discharges in the Doublet III tokamak. This level of ⟨βT⟩ is above the minimum level required for a tokamak reactor, thus demonstrating that reactor level values of ⟨βT⟩ are possible in a tokamak device. The observed enhancement of ⟨βT⟩ with vertical elongation lends confidence in the design of future devices which rely on vertical elongation.