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We have used experiments and modeling to develop a new radiative divertor configuration for DIII-D. Gas puffing experiments with the existing open divertor have shown the creation of a localized (∼ 10 cm diameter) radiation zone which results in substantial reduction (3–10) in the divertor heat flux while τ E remains ∼ 2 times ITER-89P scaling. However, n e increases with D 2 puffing, and Z eff increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity (δ ≈ 0.8) is important for high τ E VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented.

In this paper, we examine a denser and broader scrape-off layer (SOL) plasma, one class of which was first seen in VH mode, in the DIII-D tokamak. An enhanced SOL appears in many types of double-null (DN) discharges and is not a property of VH mode only. The L and H mode DN enhanced SOL density and temperature profiles exhibit a 5–6 cm broad profile outside the separatrix. The VH mode enhanced SOL is associated with a higher separatrix density and temperature and is only 1.5–2 cm wide. Both types of SOL enhancement are only observed in high triangularity discharges. The origin of the enhanced SOL is, however, not yet understood.

D{sub 2} gas injected into ELMing H-mode discharges in DIII-D reduced total integrated heat flux to the divertor by {approximately}2{times} and peak heat flux by {approximately}5{times}, with only modest degradation to plasma stored energy. Steady gas injection without particle pumping results in eventual degradation in stored energy. The initial reduction in peak heat flux at the divertor tiles may be primarily due to the increase in radiated power from the X-point/divertor region. The eventual formation of a high density region near the X-point appears to play a role in momentum (and energy) transfer from the flux surfaces near the outboard strike point to flux surfaces farther out into the scrapeoff. This may also contribute to further reduction in peak heat flux.

A new modeling effort is made in order to understand the observed relatively low levels of impurity contamination during the VH-mode phase on DIII-D, as compared to those observed during the H-mode phase of selected discharges. The key element is the inclusion of the real 2-D flux surface geometry in the prediction of impurity penetration of sputtered atoms through the scrape-off layer into the core plasma. Of the elements which determine the impurity content in the plasma: sputtering yield, penetration, and core transport, the penetration through the scrape-off layer is found to be the most determinative factor. The low impurity content in VH-mode is attributed to the development of a scrape-off layer with higher density and temperature properties than those normally obtained in H-mode.

A thin boron film has been applied to the DIII-D tokamak plasma facing surfaces to reduce impurity influx, particularly oxygen and carbon. A direct result of this surface modification was the observation of a regime of very high energy confinement, VH-mode, with confinement times from 1.5 to 2 times greater than predicted by H-mode scaling relation for the same set of parameters. VH-mode discharges are characterized by low ohmic target densities, low edge neutral pressure, and reduced cycling. These conditions have reduced the collisionality, {nu}*, in the edge region producing a higher edge pressure gradient and a significant bootstrap current, up to 30% of the total current. We will describe the edge plasma properties after boronization including reductions in recycling inferred from measurements of {tau}{sup p}*. In particular we will discuss the edge plasma conditions necessary for access to VH-mode including the boronization process and properties of the deposited film.

The DIII-D tokamak has been carbonized by the application of a carbon film covering all plasma facing surfaces. Carbonization was done in order to reduce the metal impurity influx and central metal accumulation especially during beam heated D{sup O} {yields} D{sup +} H-mode discharges. After carbonization, nickel impurity line radiation was reduced by a factor of 10 during the ohmic phase of the discharge and up to a factor of 30 during the H-mode phase. The reduction of metal impurities also produced a reduction of total radiated power and allowed high current operation. After carbonization, the highest plasma current in a double null divertor, and the highest stored energy ever achieved on DIII-D were observed, 3 MA and 3.6 MJ respectively. A toroidal beta, {beta}{sub T} = 5.1%, at full field, 2.1 T, was obtained. On the first day after carbonization, H-mode density profiles were more peaked than pre-carbonization discharges. 20 refs., 6 figs.

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.