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A digital–analog integrator has been developed for use with inductive magnetic sensors in long-pulse tokamaks. Continuous compensation of input offsets is accomplished by alternating analog-to-digital convertor samples from the sensor and a dummy load, while a RC network provides passive integration between samples. Typically a sampling rate of 10 kHz is used. In operational tests on the DIII-D tokamak, digital and analog integration of tokamak data show good agreement. The output drift error during a 1200 s integration interval corresponds to a few percent of the anticipated signal for poloidal field probes in International Thermonuclear Experimental Reactor, and bench tests suggest that the error can be reduced further.

In highly elongated DIII-D divertor discharges, it was found that the line integrated X-ray sawtooth inversion radius scales as rinv/a = 1/q95 for a wide range of plasma parameters including low q and high beta discharges. No density or elongation dependence was observed. The sawtooth inversion radius showed little dependence on injected beam power in H-mode discharges at full toroidal field (2.1 T). At lower fields, the sawtooth inversion radius was observed to increase somewhat with increasing input power. Also, it was found that the sawtooth mixing radius scales as rmix/a = 1/q95 + 0.3 and is independent of elongation.

Using the outside launch with vertical polarization, electron cyclotron heating (ECH) experiments in DIII-D were carried out at 60 GHz at fundamental and second harmonic frequencies. The results are consistent with wave absorption theory for accessibility. For fundamental heating with a cut-off layer between the launch location and the resonant surface, good absorption efficiency was obtained. A process for mode conversion at the wall from the extraordinary mode to the ordinary mode is postulated to explain this result. For most efficient heating it was important to avoid the presence of a locked mode. For low density plasmas at the fundamental, the global energy confinement exceeds Kaye-Goldston scaling by about a factor of 1.4. However, heating at the second harmonic agrees numerically with the Kaye-Goldston scaling formula, for ECH alone or for ECH combined with neutral beam injection (NBI), and shows a mass dependence consistent with TE ~ √Ai. The global energy increase was insensitive to the resonance location for qres 1, global energy confinement scales as TE ~ Bt−0.3. Since plasma current and plasma geometry were not varied for these discharges, a scaling with safety factor q is also consistent. Using ECH alone, H-mode confinement was achieved for second harmonic heating in the central core of the plasma at a power threshold of ~0.75 MW for plasma parameters Ip = 0.5 MA, and e = 1.1 × 1019 m−3. The energy confinement times during the H-mode are similar to Ohmic values with a normalized confinement time τE/Ip ~ 200 ms/MA. For ECH and NBI H-mode plasmas, the transition from the L-mode to the H-mode was correlated with reduced magnetic fluctuations in the divertor region. For ECH H-mode plasmas, transport analysis shows that (1) the H-mode obtained with ECH is accompanied by an improvement of the electron thermal diffusivity χe relative to the L-mode over at least part of the plasma, (2) the ion thermal diffusivity χi is larger than predicted by neoclassical theory for the L- and H-modes, and (3) χe is approximately equal to χi for the ohmically heated discharge and for the L-mode and H-mode discharges heated by ECH discussed in the paper.

Magnetic diagnostics are critical for future tokamaks such as the International Thermonuclear Experimental Reactor (ITER). Existing tokamaks use inductive coils and loops as magnetic sensors. The sensors are located outside the plasma boundary but within the plasma's external magnetic fields. The electrical signals from these sensors are integrated over time to determine the plasma shape and are used for plasma control. Present magnetic diagnostics are designed for plasmas lasting seconds to tens of seconds. In their present form, the inductive magnetic sensors and integrators; are not suitable for future tokamaks with their long duration plasmas lasting for 1000 s or more. Namely, magnetic diagnostics for future tokamaks require: (1) sensors that will not overheat due to plasma radiation heating, (2) sensor materials that can survive the neutron fluence, (3) sensors designed for the remote handling requirements, and (4) integrators that have a 1000 s measurement capability. This paper presents a high frequency inductive magnetic sensor being developed for ITER.

This paper reviews the design of new diagnostics and the modifications to existing diagnostics needed to carry out radiative divertor experiments in DIII-D following installation in late 1996 of a set of baffle structures that will restrict the backflow to the core plasma of neutral deuterium atoms and impurity gases. The divertor slots formed by the new baffle structures will inhibit the easy view of the divertor legs and target plates that the open divertor geometry in DIII-D currently affords. We review a basic set of diagnostics that are needed to demonstrate the reduction of divertor heat loading and radiative dissipation of energy within the divertor. This will include IR cameras, bolometry, foil bolometers, and Langmiur probes. Within the limits of available funding, we will implement a supplemental set of instruments which provide a more detailed understanding of the underlying physical processes. Many existing diagnostics require only re-aiming to provide proper coverage of the initial 23 cm long divertor plasma configuration (X-point to floor distance). Other diagnostics need extensive reconfiguration using in-vessel fiber-optic bundles or high power laser mirrors. The new divertor baffle panels provide a protective shelf for diagnostic hardware mounted underneath them, but the water cooling channels in the panels limit the permissible size of through holes and, thereby, restrict the available views of under-the-baffle diagnostics. The successful resolution of the design and implementation of these diagnostic modifications is dependent on a strong coordination between GA and its many diagnostic collaborators.

Interferometer measurements on Doublet III indicate that a region of high-density cold plasma exists near the inside limiter. This cold plasma region appears above a certain threshold density in the main plasma and can be more than five times the average plasma density. The formation of the high-density cold region occurs somewhat before the maximum achievable plasma density.

Ultrasoft x‐ray diode (USXRD) arrays have been used on DIII‐D and ASDEX to study plasma edge radiation, in the photon energy range from 10 eV to 10 keV. The detectors are extremely useful and versatile due to their simplicity and compactness. Furthermore, absolute quantum efficiencies (QE) of many photocathodes such as vitreous C, Al, Cu, CuI, CsI, and Au have been measured in recent years. With filter technique, broadband resolution, E/ΔE∼1, is possible. Quantum‐efficiency comparison of USXRD with semiconductor XRD is also presented to better understand the regions of applicability for each detector.

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.

Summary The first Neutral Injection System Beamline became operational on the Doublet III tokamak in September 1981. Power delivery tests and system debugging proceeded through December and initial experiments on plasma heating were started in January 1982. This first beamline has been running well and as of February 1982 is routinely delivering about 2.0 MW of neutral hydrogen power to various circular and elongated dee-shaped deuterium plasmas. The second beamline is now being brought on-line and we expect to have 5–6 MW of heating power available for definitive β studies in mid 1982. We find that the change in the total plasma stored energy caused by the beams is essentially independent of toroidal field and plasma current, but increases weakly with density. The total stored energy increases with current and density, but is independent of toroidal field. A volume-average βT= 2.2% is obtained in low toroidal field discharges.

In SIII-D, good H-mode confinement has been obtained with up to 6 MW of neutral-beam injection (NBI). In this report we summarize the essential features of divertor operation for DIII-D H-mode plasma. Briefly, our measurements show that: Large edge density and temperarture gradients are obtained with scale lengths of a few cm at the midplane; In the steady state, approximately 10-20% of the input power is deposited on the divertor plates in a region that is peaked on the separatrix and extends radially over )approx equal) 2-4 cm at the outboard intercept. More power (2:1) reaches the outboard intercept than the inboard; Low electron temperature (less than or equal to20eV) and high density (n/sub e//sup div/ )approx equal) n/sub e//sup main/ greater than or equal to 3-10 )times)10/sup 19/ m/sup -3/ at the divertor plates indicate that DIII-D obtains a high-recycling divertor during H-mode. DEGAS neutral transport modeling is consistent with these results; The duration of good H-mode confinement is limited by Edge Localized Modes (ELMs) rather than increasing impurity radiation; The dominant impurities, C and Ni are concentrated in the edge plasma (ra greater than or equal to 0.5) for I/sub p/ greater than or equal to 1 MA;more » and Global impurity concentrations fall markedly with increasing plasma current. The remainder of this report contains a brief description of the DIII-D tokamak and the relavant diagnostics, a summary of the divertor plasma parameters obtained during the H-mode and the heat loads on the diverotr target plates. The results of measurements of the central radiation and impurity content using data from an EUV spectrometer and a 21 channel bolometer array are discussed, and finally a discussion of the implication of these results for the CIT and ITER tokamak designs is given. 8 refs., 8 figs« less