D.N. Hill

Divertor heat flux reduction by D2 injection in DIII-D

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.

Simulation of experimentally achieved DIII‐D detached plasmas using the UEDGE code

The introduction of a divertor Thomson scattering system in DIII‐D [J. Luxon et al., International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1986), p. 159] has enabled accurate determination of the plasma properties in the divertor region. Two plasma regimes are identified: detached and attached. The electron temperature in the detached regime is about 2 eV, much lower than 5–10 eV determined earlier. Fluid models of the DIII‐D scrape‐off layer plasma successfully reproduce many of the features of these two regimes, including the boundaries for transition between them. Detailed comparison between the results obtained from the fluid models and experiment suggest the models underestimate the spatial extent of the low‐temperature region associated with the detached plasma mode. Low‐temperature atomic physics processes that are not included in the present models may account for this discrepancy.

Initial operation of the divertor Thomson scattering diagnostic on DIII-D

The first Thomson scattering measurements of n{sub e} and T{sub e} in the divertor region of a tokamak are reported. These data are used as input to boundary physics codes such as UEDGE and DEGAS and to benchmark the predictive capabilities of these codes. These measurements have also contributed to the characterization of tokamak disruptions. A Nd:YAG laser (20 Hz, 1 J, 15 ns, 1064 nm) is directed vertically through the lower divertor region of the DIII-D tokamak. A custom, aspherical collection lens (f /6.8) images the laser beam from 1-21 cm above the target plates into eight spatial channels with 1.5 cm vertical and 0.3 cm radial resolution. 2D mapping of the divertor region is achieved by sweeping the divertor X-point location radially through the fixed laser beam location. Fiber optics carry the light to polychromators whose interference filters have been optimized for low T{sub e} measurements. Silicon avalanche photo diodes measure both the scattered and plasma background light. Temperatures and densities are typically in the range of 5-200 eV and 1 - 10 x 10{sup 19} m{sup -3} respectively. Low temperatures, T{sub e} 8x10{sup 20} m{sup -3} have been observed in detached plasmas. Background light levels have not been a significant problem. Reduction of the laser stray light permits Rayleigh calibration. Because of access difficulties, no in-vessel vacuum alignment target could be used. Instead, an in situ laser alignment monitor provides alignment information for each laser pulse. Results are compared with Langmuir probe measurements where good agreement is found except for regions of high n{sub e} and low T{sub e} as measured by Thomson scattering.

A fast scanning probe for DIII–D

A fast reciprocating probe has been developed for DIII–D which can penetrate the separatrix during H mode with up to 5 MW of NBI heating. The probe has been designed to carry various sensor tips into the scrape‐off layer at a velocity of 3 m/s and dwell motionless for a programmed period of time. The driving force is provided by a pneumatic cylinder charged with helium to facilitate greater mass flow. The first series of experiments have been done using a Langmuir probe head with five graphite tips to measure radial profiles of ne, Te, φf, ne, and φf. The amplitude and phase of the fluctuating quantities are measured by using specially constructed vacuum compatible 5‐kV coaxial transmission lines which allow us to extend the measurements into the MHz range. TTZ ceramic bearings and fast stroke bellows were also specially designed for the DIII–D probe. Initial measurements will be presented.

Infrared thermography of the DIII‐D divertor targets

The power flow to the DIII‐D divertor targets is routinely measured using infrared (IR) thermography. An IR television camera sensitive to radiation in the 8–12‐μm range views the divertor region using a set of germanium optics. Digital signal processing is used to extract the desired surface temperature profiles from the analog data (≂12 Mbytes) stored on videotape. Inversion of these data using a simple matrix formulation of the inverse heat conduction problem then yields the incident heat flux as a function of space and time. Results from a DIII‐D discharge are included.

Survey of target plate heat flux in diverted DIII-D tokamak discharges

A series of observations is presented concerning divertor heat flux, qdiv, in the DIII-D tokamak, and it is shown that many features can be accounted for by assuming that the heat flux flows preferentially along field lines because τ|| < τ⊥ in the scrape-off layer (SOL). Exceptions to this agreement are pointed out and the discrepancies explained by means of two dimensional (2-D) effects. About 80% of the discharge input power can be accounted for. The power deposited on the target plate due to enhanced losses during edge localized modes (ELMs) is less than 10% of the total target power in most cases. X point height scans for lower single null (LSN) diverted discharges show that the peak heat flux variation is primarily due to flux expansion and secondarily due to transport of energy across the magnetic field in the divertor. At the outer strike point qdiv,peak Pin(Ip - Ip,0)G(gin)(1/Bt)4/9(Bdiv/Bmp)f(Ldivχ⊥), where G is a linear function of the inner gap, gin, over a specified range and f describes cross-field energy transport in the divertor. Evidence of radial in-out asymmetries (comparing the outer strike point with the inner strike point or centre-post) and toroidal asymmetries in qdiv is presented and the heat flux peaking due to tile gaps and misalignment of tiles is examined. For magnetically balanced double null (DN) discharges with downward ∇B ion drift, it is found that qdiv is inherently higher in the lower divertor than in the upper divertor, having a 3:1 downward bias. Examples of heat flux reduction by gas puffing deuterium or neon in LSN and DN discharges are given. At least a threefold reduction of the peak heat flux in both the upper and lower divertors of a DN discharge, using D2 puffing, is reported.

Fast reciprocating Langmuir probe for the DIII-D divertor

A new reciprocating Langmuir probe has been used to measure density and temperature profiles, ion flow, and potential fluctuation levels from the lower divertor floor up to the X-point on the DIII-D tokamak. This probe is designed to make fast (2 kHz swept, 20 kHz Mach, 500 kHz Vfloat) measurements with 2 mm spatial resolution in the region where the largest gradients on the plasma open flux tubes are found and therefore provide the best benchmarks for SOL and divertor numerical models. Profiles are constructed using the 300 ms time history of the probe measurements during the 25 cm reciprocating stroke. Both single and double null plasmas can be measured and compared with a 20 Hz divertor Thomson scattering system. The probe head is constructed of four different kinds of graphite to optimize the electrical and thermal characteristics. Electrically insulated pyrolytic graphite rings act as a heat shield to absorb the plasma heat flux on the probe shaft and are mounted on a carbon/carbon composite core for mechanical strength. The Langmuir probe sampling tips are made of a linear carbon fiber composite. The mechanical, electrical, data acquisition and power supply systems design will be described. Initial measurements will also be presented.

A review of ELMs in divertor tokamaks

Abstract Edge localized modes (ELMs) are the focus of increasing attention by the edge physics community because of the potential impact that the large divertor heat pulses due to ELMs would have on the divertor design of future high power tokamaks such as ITER. This paper reviews what is known about ELMs, with an emphasis on their effect on the scrape-off layer and divertor plasmas. ELM effects have been measured in the ASDEX-U, C-Mod, COMPASS-D, DIII-D, JET, JFT-2M, JT-60U and TCV tokamaks, and are reported here. At least three types of ELMs have been identified and their salient features determined. Type 1 giant ELMs can cause the sudden loss of up to 10–15% of the plasma stored energy, but their amplitude ( ΔW W ) does not increase with heating power. Type 3 ELMs are observed near the H-mode power threshold and produce small energy dumps (1–3% of the stored energy). All ELMs increase the scrape-off layer plasma and produce particle fluxes on the divertor targets which are as much as ten times larger than the quiescent phase between ELMs. The divertor heat pulse is largest on the inner target, unlike that of L-mode or quiescent H-mode; some tokamaks report radial structure in the heat flux profile which is suggestive of islands or helical structures. The power scaling of type 1 ELM amplitude and frequency has been measured in several tokamaks and has recently been applied to predictions of the ELM size in ITER. Concern over the expected ELM amplitude has led to a number of experiments aimed at demonstrating active control of ELMs. Impurity gas injection with feedback control on the radiation loss in ASDEX-U suggests that a promising mode of operation (the CDH-mode) with very small type 3 ELMs can be maintained with heating power well above the H-mode threshold, where giant type 1 ELMs are normally observed. While ELMs have many potential negative effects, the beneficial effect of ELMs in providing density control and limiting the core plasma impurity content in high confinement H-mode discharges should not be overlooked.

UEDGE and DEGAS modeling of the DIII-D scrape-off layer plasma☆

This paper presents work to develop benchmarked theoretical models of scrape-off-layer (SOL) characteristics in diverted tokamaks by comparing shot simulations using the UEDGE plasma fluid and DEGAS neutral transport codes to measurements of the DIII-D SOL plasma. The experimental data include the radial profiles of n{sub e} T{sub e}, and T{sub i}, the divertor exhaust power, the intensity of H{sub {alpha}} emission, and profiles of the radiated power. A very simple model of the anomalous perpendicular transport rates produces consistency between the calculated and measured radial profiles of the divertor power, and of the midplane densities and temperatures. Experimentally, the measured exhaust power is now 80--90% of the input power. The simulated peak power on the outer leg of the divertor floor is now within 20% of the measured power. Various sensitivities of these comparisons to model assumptions are described. Finally, these benchmarked models have been used to examine the effects of various baffle configurations for the Radiative Divertor Upgrade in DIII-D.

Recent DIII-D divertor research

DIII-D currently operates with a single- or double-null open divertor and graphite walls. Active particle control with a divertor cryopump has demonstrated density control, efficient helium exhaust, and reduction of the inventory of particles in the wall. Gas puffing of D{sub 2} and impurities has demonstrated reduction of the peak divertor beat flux by factors of 3--5 by radiation. A combination of active cryopumping and feedback-controlled D{sub 2} gas puffing has produced similar divertor heat flux reduction with density control. Experiments with neon puffing have shown that the radiation is equally-divided between a localized zone near the X-point and a mantle around the plasma core. The density in these experiments has also been controlled with cryopumping. These experimental results combined with modeling were used to develop the new Radiative Divertor for DIII-D. This is a double-null slot divertor with four cryopumps to provide particle control and neutral shielding for high-triangularity advanced tokamak discharges. UEDGE and DEGAS simulations, benchmarked to experimental data, have been used to optimize the design.