Divertor plasma studies on DIII-D: experiment and modelling (original) (raw)
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Radiative divertor and scrape-off layer experiments in open and baffled divertors on DIII-D
Nuclear Fusion, 1999
Recent progress towards an increased understanding of the physical processes in the divertor and scrape-off layer (SOL) plasmas in DIII-D has been made possible by a combination of new diagnostics, improved computational models and changes in divertor geometry. The work focused primarily on ELMing H mode discharges. The physics of partially detached divertor plasmas, with divertor heat flux reduction by divertor radiation enhancement using D2 puffing, was studied in two dimensions, and a model of the heat and particle transport was developed that includes conduction, convection, ionization, recombination and flows. Plasma and impurity particle flows were measured with Mach probes and spectroscopy and compared with the UEDGE model. The model now includes self-consistent calculations of carbon impurities. Impurity radiation was increased in the divertor and SOL with 'puff and pump' techniques using SOL D2 puffing, divertor cryopumping and argon puffing. The important physical processes in plasma-wall interactions were examined with a DiMES (divertor material evaluation system) probe, plasma characterization near the divertor plate and the REDEP code. Experiments comparing single null plasma operation in baffled and open divertors demonstrated a change in the edge plasma profiles. These results are consistent with a reduction in the core ionization source calculated with UEDGE. Divertor particle control in ELMing H mode with pumping and baffling resulted in a reduction in H mode core densities to ne/nGr ≈ 0.25 (with nGr the Greenwald density). Divertor particle exhaust and heat flux were studied as the plasma shape was varied from a lower single null to a balanced double null, and finally to an upper single null.
Compatibility of the radiating divertor with high performance plasmas in DIII-D
Journal of Nuclear Materials, 2007
A radiating divertor approach was successfully applied to high performance 'hybrid' plasmas [M.R. Wade et al., in: Proceedings of the 20th IAEA Fusion Energy Conference, Vilamoura, Portugal, 2004]. Our techniques included: (1) injecting argon near the outer divertor target, (2) enhancing the plasma flow into the inner and outer divertors by a combination of particle pumping and deuterium gas puffing upstream of the divertor targets, and (3) isolating the inner divertor from the outer by a structure in the private flux region. Good hybrid conditions were maintained, as the peak heat flux at the outer divertor target was reduced by a factor of 2.5; the peak heat flux at the inner target decreased by 20%. This difference was caused by a higher concentration of argon at the outer target than at the inner target. Argon accumulation in the main plasma was modest (n Ar /n e 6 0.004 on axis), although the argon profile was more peaked than the electron profile.
Comprehensive 2D measurements of radiative divertor plasmas in DIII-D
Journal of Nuclear Materials, 1997
This paper presents a comparison of the total radiated power profile and impurity line emission distributions in the SOL and divertor of DIII-D. This is done for ELMing H-mode plasmas with heavy deuterium injection (Partially Detached Divertor operation, PDD) and those without deuterium puffing. Results are described from a series of dedicated experiments performed on DIII-D to systematically measure the 2-D (R,Z) structure of the divertor plasma. The discharges were designed to optimize measurements with new divertor diagnostics including a divertor Thomson scattering system. Discharge sequences were designed to produce optimized data sets against which SOL and divertor theories and simulation codes could be benchmarked. During PDD operation the regions of significant radiated power shift from the inner divertor leg and SOL to the outer leg and X-point regions. Da emission shifts from the inner strikepoint to the outer strikepoint. Carbon emissions (visible CII and CIII) shift from the inner SOL near the X-point to a distributed region from the X-point to partially down the outer leg during moderate D2 puffing. In heavy puffing discharges the carbon emission coalesces on the outer separatrix near the X-point and for very heavy puffing it appears inside the last closed flux surface above the X-point. Calibrated spectroscopic measurements indicate that hydrogenic and carbon radiation can account for all of the radiated power. La and CIV radiation are comparable and when combined account for as much as 90% of the total radiated power along chords viewing the significant radiating regions of the outer leg.
Physics of Plasmas, 1997
The impact of two dimensional (2D) effects of energy transport on impurity radiation fronts in a tokamak Scrape off Layer (SOL) plasma is considered. It is shown that 2D effects significantly alter both the physics of the fronts and the radiation loss from the SOL plasma, explain the experimental observations of impurity radiation region jumps from the target to the X-point after transition to a detached regime, and suggest an explanation for the easier access to a detached divertor regime in "vertical" target geometry in comparison with the "horizontal" case.
Plasma recombination and molecular effects in tokamak divertors and divertor simulators
Physics of Plasmas, 1997
Analysis of the experimental data from tokamaks and linear divertor simulators leads to the conclusion that plasma recombination is a crucial element of plasma detachment. Different mechanisms of plasma recombination relevant to the experimental conditions of the tokamak scrape-off layer ͑SOL͒ and divertor simulators are considered. The physics of Molecular Activated Recombination ͑MAR͒ involving vibrationally excited molecular hydrogen are discussed. Although conventional Electron-Ion Recombination ͑EIR͒ alone can strongly alter the plasma parameters, MAR impact can be substantial for both tokamak SOL plasma and divertor simulators. Investigation of the effects of EIR on the plasma flow in divertor simulators shows that due to the balances of ͑a͒ energy transport and electron cooling, and ͑b͒ the plasma flow and recombination, that EIR extinguishes the simulator plasma at an electron temperature about 0.15 eV.
Radiating divertor experiments in the HL-2A tokamak
Journal of Nuclear Materials, 2009
Radiating divertor experiments have been performed in HL-2A using different fueling methods, such as direct gas puffing (GP), supersonic molecular beam injection (SMBI), and noble gas injection in divertor. The plasma temperatures at inner and outer target plates can be decreased below 2 eV in completely detached plasma (CDP). Electron pressures at target plates, radiation power in divertor and the compression ratio (R p0 ) of neutral gas pressures between the divertor and main chamber gradually drop during the detachment. Partial detachment first appears at inner target plate even if plasma density is very low. No clear high-recycling regime is observed before the detachment. It is more difficult to observe the partial detachment if the drift direction of magnetic field gradient is away from X-point because the electron temperature at inner target plate is higher than that at outer one.
Divertor design for the tokamak physics experiment
Journal of Nuclear Materials, 1995
In this paper we discuss the divertor design for the planned TPX tokamak, which will explore the physics and technology of steady state (1000 s pulses) heat and particle removal in high confinement (up to 4 × L-mode), high beta (up to /3 N = 5) divertor plasmas sustained by non-inductive current drive. TPX will operate in the double-null divertor configuration, with actively cooled graphite targets forming a deep (0.57 m) slot at the outer strike point. The peak heat flux on the highly tilted (74 ° from normal) re-entrant divertor plate (tilted to recycle ions back toward the separatrix) will be in the range of 4-6 MW/m z with 17.5 MW of auxiliary heating power. The combination of pumping and gas puffing (D 2 plus impurities), along with higher heating power (45 MW maximum) will allow testing of radiative divertor concepts at ITER-like power densities.
On mechanisms of impurity leakage and retention in the tokamak divertor
Plasma Physics and Controlled Fusion, 2019
Impurity seeding into a tokamak divertor for radiative cooling is considered as a tool for achieving detached/semi-detached regimes required to meet the condition of acceptable heat loads on divertor plates. Experiments aimed at searching of the operational window with a significant reduction of poloidal heat fluxes due to the impurity radiation and without the decreasing of confinement are performed on many tokamaks. Critical issue in these experiments is which fraction of impurities is retained in the divertor region and which is extracted upstream to the scrape-off layer (SOL). In the present paper a physical mechanism of impurity transport from a divertor towards upstream and back to the divertor is analyzed. It is demonstrated that the widespread concept that the impurity leaks if the parallel thermal force exceeds the friction due to main ions and retains otherwise-is not correct. In contrast, the impurity leaks if it crosses the stagnation point of impurity ion poloidal velocity profile before the ionization, and retains if it ionizes closer to the target than the location of that stagnation point. Thus the leakage efficiency depends on the relative spatial positions of the impurity atom ionization source and the stagnation point of the impurity ion poloidal velocity profile. The impurity ion poloidal velocity is a sum of poloidal projection of its parallel velocity and the E × B drift velocity, where the former should be defined from the parallel impurity force balance equation. It is demonstrated that the solution of this equation may be approximated by the balance of friction and thermal forces in all regimes, while other terms are smaller. This allows for expressing the impurity parallel velocity through the main ion one and makes the distribution of the parallel (poloidal) fluxes of the main ions, including Pfirsch-Schlüter (PS) fluxes and E × B drift fluxes, to be an important element of the impurity transport. It is shown that impurity distribution in the edge plasma is rather sensitive to the value of the impurity ion ionization potential. This analysis is supported by the simulation results obtained for the ASDEX Upgrade tokamak with various seeding rates of N and Ne with the SOLPS-ITER code. The importance of inclusion of self-consistent drift flows is demonstrated by the comparison to result of corresponding simulations with drifts turned off.
Plasma detachment in JET Mark I divertor experiments
Nuclear Fusion, 1998
The experimental characteristics of divertor detachment in the JET tokamak with the Mark I pumped divertor are presented for Ohmic, L-mode and ELMy H-mode experiments with the main emphasis on discharges with deuterium fuelling only. The range over which divertor detachment is observed for the various regimes as well as the influence of divertor configuration, direction of the toroidal field, divertor target material and active pumping on detachment will be described. The observed detachment characteristics such as the existence of a considerable electron pressure drop along the field lines in the scrape-off layer, and the compatibility of the decrease in plasma flux to the divertor plate with the observed increase of neutral pressure and the D α emission from the divertor region will be examined in the light of existing results from analytical and numerical models for plasma detachment. Finally, a method to evaluate the degree and window of detachment is proposed and all the observations of the JET Mark I divertor experiments summarised in the light of this new quantitative definition of divertor detachment.
Journal of Nuclear Materials, 2001
We present results from DIII-D experiments and modeling focused on the divertor issues of an "Advanced Tokamak" (AT). Operation at high plasma pressure β with good energy confinement H requires core and divertor plasma shaping and current profile J(r) control with ECH current drive. Transport modeling indicates that the available DIII-D ECH power determines a density and temperature regime for sustained DIII-D AT experiments. We demonstrate that a high-δ, unbalanced double null divertor with cryopumping (D-2000) is a flexible AT divertor. Impurity levels in AT experiments have been reduced by careful alignment of the divertor tiles; this, in turn has changed the time evolution of the core J(r) profiles. New physics has been observed near the X-point and private flux regions, including flow reversal and recombination, that is important in understanding and controlling the flows and thereby the radiation in the divertor region, which reduces the divertor heat flux.