David N. Hill

2D tomography with bolometry in DIII‐Da)

A 48‐channel platinum‐foil bolometer system on DIII‐D was installed to achieve better spatial and temporal resolution of the radiated power in diverted discharges. Two 24‐channel arrays provide complete plasma coverage with optimized views of the divertor. The divertor radiation profile was measured for a series of radiative divertor and power balance experiments. A significant change in the magnitude and distribution of divertor radiation with heavy gas puffing was observed. Unfolding the radiation profile with only two views requires one to treat the core and divertor radiation separately. The core radiation is fitted to a function of magnetic flux and is then subtracted from the divertor viewing chords. The divertor profile is then fit to a 2D spline as a function of magnetic flux and distance from divertor floor.

Measurements of non-axisymmetric effects in the DIII-D divertor

Abstract Non-stationary toroidal asymmetries are observed in the DIII-D divertor heat flux and scrape-off layer (SOL) currents. Using the present DIII-D diagnostics, asymmetries are seen much less frequently in single-null H-modes ( 50%). Divertor heat flux asymmetries are characterized by toroidal variations in the radial profile (i.e., multiple or bifurcated peaks) at some toroidal locations and single peaks at other toroidal locations while SOL currents sometimes have a strongly bipolar toroidal structure. SOL current asymmetries are particularly large during Edge Localized Modes (ELMs). The measurements reported here indicate that the asymmetries are best described by a model in which non-axisymmetric radial magnetic perturbations create magnetic islands in the plasma boundary and SOL which then cause toroidal variation in the divertor heat flux and the SOL currents.

IMPURITY ENRICHMENT AND RADIATIVE ENHANCEMENT USING INDUCED SOL FLOW IN DIII-D

Abstract Experiments on DIII-D have demonstrated the efficacy of using induced scrape-off-layer (SOL) flow to preferentially enrich impurities in the divertor plasma. This SOL flow is produced through simultaneous deuterium gas injection at the midplane and divertor exhaust. Using this SOL flow, an improvement in enrichment (defined as the ratio of impurity fraction in the divertor to that in the plasma core) has been observed for all impurities in trace-level experiments (i.e., impurity level is non-perturbative), with the degree of improvement increasing with impurity atomic number. In the case of argon, exhaust gas enrichment using modest SOL flow is as high as 17. Using this induced SOL flow technique and argon injection, radiative plasmas have been produced that combine high radiation losses ( P rad / P input xa0>xa070%), low core fuel dilution ( Z eff τ E xa0>xa01.0 τ E,ITER93H ).

Effect of divertor bias on plasma flow in the DIII-D scrape-off layer

A toroidal ring bias electrode in contact with the single-null divertor scrape-off layer in the DIII-D tokamak achieved electric field control of plasma flow in the layer. Bias strongly altered the static pressure of neutralized gas in the new divertor chamber (designed for future pumping) during Ohmic, L-mode and ELMing H-mode plasma operation. The pressure increased or decreased, depending on the signs of the applied electric potential and toroidal magnetic field, in agreement with the E*B drift direction. The variation of the core plasma density was opposite to the variation of the pump chamber gas pressure and was proportionally much smaller. A large bias induced pressure rise appeared even when the scrape-off layer was not directed at the pump entrance aperture-a surprising effect that is interpreted qualitatively in terms of radial potential discontinuities. Bias also affected divertor heat fluxes, and preliminary observations are discussed qualitatively

The effect of ELMs on edge plasma scaling in DIII-D

In this paper we report results of scaling studies aimed at determining how the divertor conditions vary with plasma current, toroidal field, and neutral beam heating power in H-mode discharges with ELMs in the DIII-D tokamak. We find that ELMs produce relatively more direct particle losses (50% or more of the total) than energy losses (≤20%). The time-average peak divertor heat flux in these plasmas is found to scale as d α ( P NBI I p )( B p,mp / B p,div ). The linear power dependence suggests that the plasma sheath at the targets is primarily responsible for limiting the parallel energy flow, while the I p variation may mean that the radial energy transport in the SOL decreases with increasing plasma current, just as it does in the core.

DIVERTOR NEUTRAL PRESSURE ENHANCEMENT WITH A BAFFLE IN DIII-D

The open divertor in DIII-D has been modified by the installation of a baffled plenum with a narrow entrance aperture. Neutral particle pressures in the range of 2-20 mtorr have been measured in this new plenum during experiments in L and H mode plasmas in which the neutral beam power was varied and the divertor target location was scanned with respect to the entrance aperture. These pressures exceed the predicted minimum required to provide adequate particle exhaust for controlling the density in H mode discharges when pumping experiments are conducted in the future. The pressure measurements presented here, which were carried out in the absence of a pump, were modelled with the aid of two transport codes: an edge plasma code and a neutrals code. It was found in this modelling process that the divertor flux amplification factor required to explain the measured pressures is of the order of 30 to 40, much higher than the factors used in the modelling of data in the absence of the baffle. This higher flux amplification factor results from increased recycling at the divertor owing to the presence of the baffle

Partially detached radiative divertor with active divertor pumping

Deuterium fuelled radiative diverters, operating in the partially detached divertor (PDD) regime, are effective in reducing the heat load on the divertor target. These PDD discharges are shown to be compatible with active particle exhaust conditions-an important issue for ITER type operation. The measured global and local plasma properties of actively pumped PDD discharges are also shown to have important similarities with non-pumped PDD discharges. Active particle exhaust leads to quasi-equilibrium plasma conditions in both the divertor and the main body within several particle confinement times (i.e. ~800 ms). However, changes is the gas puffing and/or pumping programmes have an almost immediate impact on the divertor plasma (e.g., the PDD regime ends ~50 ms after the injected deuterium gas how is interrupted). In contrast to the pumping results with lower density non-radiative divertor discharges, the divertor cryopump particle exhaust rate is relatively insensitive to the location of the divertor separatrix strike point during radiative divertor operation. Edge localized modes (ELMs) are shown to trigger multifaceted asymmetric radiation from the edge (MARFE) at the divertor under slightly submarginal PDD conditions

Angular dependence of power in the DIII-D divertor

Irregularities in the DIII-D divertor tile angles are used to test the dependence of deposited power on the angle of the magnetic field in the range 0–3°. Results are presented which show that the sin θ law, which is assumed in thermal design calculations for CIT and ITER divertor plates, is accurate down to 0.3°.

Plasma diagnostics for the DIII-D divertor upgrade

The DIII‐D tokamak is being upgraded to allow for divertor biasing, baffling, and pumping experiments. This paper gives an overview of the new diagnostics added to DIII‐D as part of this advanced divertor program. They include tile current monitors, fast reciprocating Langmuir probes, a fixed probe array in the divertor, fast neutral pressure gauges, and Hα measurements with TV cameras and fiber optics coupled to a high‐resolution spectrometer.

Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (nmorexa0» e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). In conclusion, this paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.«xa0less