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The DIII-D tokamak was a major upgrade of the Doublet III device and was commissioned in 1986. It featured a large noncircular plasma cross section and was optimized for understanding the role of t...

This overview of the DIII-D fusion program provides an introduction to the research program carried out on the DIII-D tokamak since its inception in 1986. It serves as the introduction and summary ...

Small non-axisymmetric magnetic fields are known to cause serious loss of stability in tokamaks, leading to loss of confinement and abrupt termination of plasma current (disruptions). The best known examples are the locked mode and the resistive wall mode. Understanding of the underlying field anomalies (departures in the hardware-related fields from ideal toroidal and poloidal fields on a single axis) and the interaction of the plasma with them is crucial to tokamak development. Results of both locked mode experiments (Scoville J.T. and La Haye R.J. 2003 Nucl. Fusion 43 250) and resistive wall mode experiments (Garofalo A.M., La Haye R.J. and Scoville J.T. 2002 Nucl. Fusion 42 1335) done in DIII-D tokamak plasmas have been interpreted to indicate the presence of a significant anomalous field. New measurements of the magnetic field anomalies of the hardware systems have been made in DIII-D. The measured field anomalies due to the plasma shaping coils in DIII-D are smaller than previously reported (La Haye R.J. and Scoville J.T. 1991 Rev. Sci. Instrum. 61 2146). Additional evaluations of systematic errors have been made. New measurements of the anomalous fields of the Ohmic heating and toroidal coils have been added. Such detailed in situ measurements of the fields of a tokamak are unique. The anomalous fields from all the coils are one-third the values indicated from the stability experiments (Garofalo et al 2002, Scoville and La Haye 2003). These results indicate limitations in the understanding of the interaction of the plasma with the external field. They indicate that it may not be possible to deduce the anomalous fields in a tokamak from plasma experiments and that we may not have the basis needed to project the error field requirements of future tokamaks.

Non-axisymmetric (error) fields in tokamaks lead to a number of instabilities including so-called locked modes [J.T. Scoville, R.J. La Haye, Nucl. Fusion 43 (4) (2003) 250–257] and resistive wall modes (RWM) [A.M. Garofab, R.J. La Haye, J.T. Scoville, Nucl. Fusion 42 (11) (2002) 1335–1339] and subsequent loss of confinement. They can also cause errors in magnetic measurements made by point probes near the plasma edge, error in measurements made by magnetic field sensitive diagnostics, and they violate the assumption of axisymmetry in the analysis of data. Most notably, the sources of these error fields include shifts and tilts in the coil positions from ideal, coil leads, and nearby ferromagnetic materials excited by the coils. New measurements have been made of the n=1 coil-related field errors in the DIII–D plasma chamber. These measurements indicate that the errors due to the plasma shaping coil system are smaller than previously reported and no additional sources of anomalous fields were identified. Thus they fail to support the suggestion of an additional significant error field suggested by locked mode and RWM experiments.

The DIII-D tokamak evolved from the earlier Doublet III device in 1986. Since then, the facility has undergone a number of changes including the installation of divertor baffles and pumping chambers in the vacuum vessel, the addition of a radiation shield, the development of extensive neutral beam and RF heating systems, and the addition of a comprehensive plasma control system. The facility has become the focus of a broad fusion plasma science research programme. An integrated picture of the facility and its capabilities is presented.

A particle exhaust scheme using a cryocondensation pump in the advanced divertor configuration of the DIII-D tokamak is described. In this configuration, the pump is located inside a baffle chamber...

An attached poloidal current, which flows in a circuit lying partly in the vacuum vessel wall and partly in the scrape-off layer of the plasma, is observed during vertical instabilities in the DIII-D tokamak. A direct measurement of the current, using Rogowski loops on several protective tiles at locations where the plasma contacts the wall, is in good agreement with the value determined from MHD equilibrium reconstructions using measured values of magnetic field and flux. This attached current, which can reach transient peaks of several hundred kiloamperes, interacts with the toroidal magnetic field to create a large vertical force on the vacuum vessel. The predicted motion of the vessel resulting from the measured currents agrees well with the observed displacement of the vacuum vessel.

A survey of the boundary physics research on the DIII-D tokamak and an outline of the DIII-D Advanced Divertor Program (ADP) is presented. We will present results of experiments on impurity control, impurity transport, neutral particle transport, and particle effects on core confinement over a wide range of plasma parameters, Ip ≲ 3 MA, βT ≲ 10.7%, P(auxiliary)≲ 20 MW. Based on the understanding gained in these studies, we in collaboration with a number of other laboratories have devised a series of experiments (ADP) to modify the core plasma conditions through changes in the edge electric field, neutral recycling, and plasma-surface interactions.

The DIII-D tokamak has made a remarkable series of contributions to the fusion science program. These contributions are due to the continual evolution of the experimental systems led by research results and technological developments. The recent completion of core and edge diagnostics allows comprehensive time-dependent profile data, including plasma current and rotation profiles to be available on each discharge. This data combined with rapid a MHD calculation of the flux surfaces is used to carry out extensive analysis between discharges (10-15 min cycle) and permits a shot-by-shot strategy to optimize research productivity. The control of the plasma internal profiles has emerged as key in developing advanced tokamak discharges. RF power has become a key tool in achieving this control. The DIII-D ECH system is presently being expanded to a capability of six MW level gyrotrons. The radiative divertor configuration has evolved as an effective tool for edge plasma management. The recent commissioning of the upper outer divertor cryopump and baffle along with the existing installation in the lower divertor has allowed understanding of the role of upper and lower divertors, double-null divertors, and the location of the X-point. This fall, installation of an upper inner cryopump and baffle will allow further investigation of the role of pumping of the private flux region of the high triangularity divertor.

A radiation shield has been added to the DIII-D tokamak facility to allow an increased neutron production rate without an increased radiation exposure to the public and to personnel. The goal is to reduce the biological dose level at the site boundary by a factor of 300. Shielding analyses were performed extensively using a 3D Monte Carlo neutron-photon coupled code (MCNP) for varied geometry and material combinations. The calculated total dose at the site boundary (83 m from the tokamak) is approximately 10/sup -17/ mrem per D-D neutron with the new shield in place, and 3*10/sup -15/ mrem without the new shield. >

The motions of the DIII-D vacuum vessel during vertical instabilities of elongated plasmas have been measured and studied over the past five years. The currents flowing in the vessel wall and the plasma scrapeoff layer were also measured and correlated to a physics model. These results provide a time history load distribution on the vessel which was input to a dynamic analysis for correlation to the measured motions. The structural model of the vessel using the loads developed from the measured vessel currents showed that the calculated displacement history correlated well with the measured values. The dynamic analysis provides a good estimate of the stresses and the maximum allowable deflection of the vessel. In addition, the vessel motions produce acoustic emissions at 21 Hz that are sufficiently loud to be felt as well as heard by the DIII-D operators. Time history measurements of the sounds were correlated to the vessel displacements. An analytical model of an oscillating sphere provided a reasonable correlation to the amplitude of the measured sounds. >

Error fields from misalignment of the toroidal field (TF) and poloidal field (PF) coils in TPX are presented in terms of the outward normal B-field (B/sub /spl perp//), expanded in poloidal and toroidal harmonics (m, n), on a simulated, D-shaped, plasma flux surface. Results are reported for n=1 toroidal mode number and low poloidal mode numbers, m, and for various displacements of the TF and PF coils. In particular, results are given for the m,n=2,1 error field which interacts with the q=2 surface to cause locked modes and loss of plasma performance. Based on existing experiments, maximum permissible field errors are 4 G for the 2,1 mode and 8 G for the n=1; m=1,3,4 modes. Results are presented for a rigid shift and rotation of a single TF coil and for a rigid, radial shift of each PF coil.

The DIII-D program focuses on developing fusion physics in an integrated program of tokamak concept improvement. The intent is both to support the present ITER physics R and D and to develop more efficient concepts for the later phases of ITER and eventual power plants. Progress in this effort can be best summarized by recent results for a diverted deuterium discharge with negative central shear which reached a performance level of Q{sub DT} = 0.32. The ongoing development of the tools needed to carry out this program of understanding and optimization continues to be crucial to its success. Control of the plasma cross-sectional shape and the internal distributions of plasma current, density, and rotation has been essential to optimizing plasma performance. Advanced divertor concepts provide edge power and particle control for future devices such as ITER and provide techniques to help manage the edge power and particle flows for advanced tokamak concepts. New divertor diagnostics and improved modeling are developing excellent divertor understanding. Many of the plasma physics issues being posed by ITER are being addressed. Scrapeoff layer power flow is being characterized to provide an accurate basis for the design of reactor devices. Ongoing studies of the density limit focus on identifying ways in which ITER can achieve the required densities in excess of the Greenwald limit. Better understanding of disruptions is crucial to the design of future reactors.

Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs.

The DIII-D tokamak employs eight neutral beam systems for plasma heating and current drive experiments. These positive ion source neutral beam systems have gone through several improvements in operational technique and in system hardware since the start of conditioning of the first long pulse ion source in December 1986. These improvements have led to the routine operation in deuterium at beam power levels of 20 MW. The improvements in operational technique include filament power supply operating mode, accelerator grid voltage holding capability, mid changes in grid potential gradients. The hardware improvements include installation of arc notching, arc discharge density regulation, and control of neutralizer gas puffing. Each of these improvements are discussed in this paper. Successful testing and operation of the ion source at 93 kV deuterium beam energy, well above the design value of 80 kV, also led to the possibility of enhancing system capability to 28 MW power level, nearly twice the original design value. Upgrading of the beam system to 60 second pulse duration at the currently achieved power level is under consideration. Studies have shown that this pulse length extension can be achieved with improvements in beamline heat handling components and auxiliary systems, especially the powermore » supply system. The drift duct (the section between neutral beam beamline and tokamak) upgrade and protection for the long pulse duration present the greatest challenges, and are crucial to achieving long pulse beam injection into the tokamak plasma.« less

The Big Dee device will be the first major tokamak to incorporate both divertor and pump limiter particle control techniques. The Big Dee is an upgrade of the Doublet III facility designed to reach reactor grade plasma parameters. The device, presently under construction, is designed to operate at plasma current of I p ≤ 5 MA (3.5 MA if diverted), input power of 20 MW and pulse duration of several seconds with either divertor or pump limiters. The diverted plasmas have a nominal major radius of 167 cm, minor radius of 55 cm, and elongation of 2. The pump limiters have a radial stroke of 50 cm and are compatible with a variety of plasma cross-sections from full size highly elongated dee-shaped plasmas ( R 0 = 167 cm, a = 67 cm, k = 2 ) to circular plasmas of reduced minor and major radii.

Ideal magnetohydrodynamic stability analysis has been carried out for configurations expected in the Big Dee tokamak, an upgrade of the Doublet III tokamak into a non-circular cross-section device which began operation early in 1986. The results of this analysis support theoretical predictions as follows: Since the maximum value of beta stable to ballooning and Mercier modes, which we denote βc, increases with inverse aspect ratio, elongation and triangularity, the Big Dee is particularly suited to obtain high values of βc and there exist high βc Big Dee equilibria for large variations in all relevant plasma parameters. The beta limits for the Big Dee are consistent with established theory as summarized in present scaling laws. High beta Big Dee equilibria are continuously accessible when approached through changes in all relevant input parameters and are structurally stable with respect to variations of input plasma parameters. Big Dee beta limits have a smooth dependence on plasma parameters such as βp and elongation. These calculations indicate that in the actual running of the device the Big Dee high beta equilibria should be smoothly accessible. Theory predicts that the limiting plasma parameters, such as beta, total plasma current and plasma pressure, which can be obtained within the operating limits of the Big Dee are reactor relevant. Thus the Big Dee should be able to use its favourable ideal MHD scaling and controlled plasma shaping to attain reactor relevant parameters in a moderate sized device.

Experiments are described in which a 2-mm-diameter cylindrical pellet of solid hydrogen is injected at low velocity to the centre of the upper half of the Doublet III tokamak, immediately before initiation of the plasma discharge. The pellet ablates in 6–9 ms and increases the initial plasma density by a factor of 2 to 3. The profiles of both electron density and current density of discharges initiated with a pellet are very flat compared to normally initiated discharges, and lower electron temperatures and lower impurity levels persist for hundreds of milliseconds. The resultant dee-shaped plasmas have height-to-width ratios of up to 1.8 on axis and 2.0 on the outermost flux surface, and are suitable targets for neutral-beam or electron cyclotron heating.

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 near-circular 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 Vs/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 Doublet III tokamak is to be reconfigured by replacing its indented (doublet) vacuum vessel with a larger one of a dee-shaped cross section. This change will permit significantly larger elongated plasmas than is presently possible and will allow higher plasma current (up to 5 MA) and anticipated longer confinement time. Reactor relevant values of stable beta and plasma pressure are predicted. This modification, while resulting in a significant change in capability, utilizes most of the existing coils, structure, systems and facility.

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.

Techniques are presented for determining the MHD equilibrium configuration for circular and non-circular plasmas from measurements of the magnetic field and flux made outside the discharge. Least-squares minimization techniques are introduced to provide an objective means for comparing the experimental data with the corresponding calculated equilibrium and for error estimation. Free-boundary equilibria using a simple model for the plasma-current profile with three free parameters – total current, poloidal beta, and a single parameter describing the peakedness of the current profile – are found to be statistically adequate and sufficient to describe the magnetic configuration of the discharge. In plasmas with significant non-circularity, the Shafranov dimensionless inductance, li, can be determined independent of βp. Examples of the application of the technique to discharges in the Doublet-Ill tokamak are presented.

The DIII-D tokamak has demonstrated the feasibility of the divertor configuration for future reactor relevant devices. H-mode (good) confinement has been demonstrated for deuterium discharges in both single-null and double-null divertor configurations and under a variety of heating scenarios including neutral beam injection, electron cyclotron heating, and ohmic heating. Plasma beta values (..beta.. = 6.8%) near expected theoretical limits (3.5 I/aB) have been demonstrated. Plasma currents are limited by safety factors q > 2 near the surface of the plasma. Energy confinement improves with increasing plasma current until q decreases below 3. Divertor heat flow has been documented and initial boundary physics studies carried out. Non-inductive current drive has been demonstrated using neutral beam injection. 7 refs., 8 figs., 1 tab.

Analysis of the expanded boundary divertor in Doublet III reveals a medium density operating regime ( n e ≲ 4 × 10 13 cm -3 ), in which (1) the divertor density ne(Div) rises with the cube of the main plasma line average density, n e , and (2) the neutral pressure at the exhaust end of the divertor has a quadratic dependence on n e (with divertor plasma temperatures ≲ 15 eV). In this regime, injection of argon into the discharge results in a steady state concentration of argon, which is two orders of magnitude higher in the divertor plasma than in the main discharge. This trapping effect suggests that the argon concentration in the divertor region may be useful as an externally controllable parameter; that is, injected argon effectively cools the plasma boundary, yet is not deleterious to the plasma as a whole. A model which successfully describes the behavior of the edge and divertor plasmas in ohmically heated discharges suggests that these desirable trapping and radiative features may be maintained in the high powered neutral beam heating phase.

Tokamak discharges using the expanded boundary divertor in the DIII-D device exhibit H-mode confinement. With neutral-beam power up to 6 MW, energy confinement remains comparable to the Ohmic value at a plasma current of 1 MA. Confinement is also independent of plasma density and toroidal field. Confinement increases with plasma current, but the exact functional dependence is, as yet, uncertain. These results show that the H mode can be achieved in a reactor-compatible open divertor configuration.

Current diffusion was studied in near-circular discharges in Doublet III when the plasma current was programmed to rise linearly from an initial steady level to a second, higher, steady level. Classical current diffusion was observed for Ip up to 7 MAs−1; discharges with Ip>2 MAs−1, however, disrupted at the transition from the current ramp to the second steady level.

This document summarizes the Experimental Plan for the DIII-D tokamak facility for the fiscal year 1989. The long-range DIII-D 5 yr plan is directed ultimately at the goal of achieving good confinement at high beta in a plasma with non-inductively driven current. This is important to the design of a steady-state reactor. This program may be thought of as occurring in two phases. In the first phase of the program we axe separately investigating high beta plasma confinement in inductively-driven plasmas, and non-inductive current drive. In the second phase we will combine these two elements to investigate high beta plasma confinement with non-inductive current drive. The FY 89 plan continues the first phase of the DIII-D experimental effort that contains a strong focus on beta and confinement in non-circular plasma configurations and in the divertor configuration in particular. Important work also continues in the development of rf heating systems for heating, profile control, and current drive. This research is coupled to theoretical efforts at General Atomics. The FY 89 research program outlined herein is diverse and multifaceted. However, it is also characterized by a greater synthesis of techniques toward a common goal. An example is the application of ECH formore » sawtooth suppression that would improve the low q confinement and allow higher {beta} to be obtained. We believe this research program will provide a solid foundation for the continued development of the tokamak toward high beta steady-state reactor application. The DIII-D FY 89 research program will provide results that will help resolve many CIT and ITER Physics R&D issues. In addition, DIII-D confinement studies will be an important input to the newly formed National Transport Task Force.« less

The results of the in-situ pellet experiments in Doublet III are summarized, with particular attention to the recent measurements of the electron temperature and density during the early phase of the discharge when large vertical elongations are achieved.

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.

Doublet IIA electron heating experiments utilize lower hybrid waves launched by slow wave structures with various n/sub parallel/ = ck/sub parallel//..omega.. (11, 14, 16) to achieve spatially localized heating by electron Landau damping. Radiofrequency power of 350 kW at 800 MHz and 200 kW at 915 MHz is available to heat circular discharges with ohmic input power of 100 kW. Significant increases in the electrical conductivity have been observed, but plasma response to the rf power is sensitive to the impurity level of the discharge. Power absorption mechanisms and energy balance will be discussed.

The program proposed here is designed to support and build toward the long-term plan put forward during 1987 for the DIII-D facility. This plan has as its ultimate goal developing sufficient understanding and predictive capability to enable the demonstration of a high beta plasma with non-inductively driven toroidal current. The early stages of this plan call for the optimization of the plasma configuration for good confinement at high beta while simultaneously developing the need rf power systems for current drive, profile control, and heating.