A computational study of operating regimes for poloidal divertors

A drop in divertor plasma density was observed during the transition to the ELM free H-mode phase as well as the drop in Dα emission intensity in JT-60U. The time scale for the drop in divertor plasma density ranged from a few ms to several tens of ms. 2-D fluid divertor code simulations indicated that the time scale for the divertor plasma density drop was ranged from several tens of ms to a hundred ms, when thermal and particle diffusivities were instantly reduced in the main plasma edge with a constant recycling coefficient. When it was assumed that the recycling coefficient decreased with a decrease in heat flux and increased with a decrease in particle flux, the time scale was reduced to a few ms in the simulations accompanied by a reduction of edge thermal diffusivity. These results indicate that the sharp drop of the divertor plasma density concomitant with the sharp drop of the Dα emission intensity on a time scale of a few ms can be ascribed to the change of plasma-wall interaction depending on the heat flux to the divertor plates.

Access conditions for H modes with detached divertor plasmas are derived by considering power threshold, divertor plasma temperature and density limit constraints. It is found that there is a minimum scrape-off layer power and main plasma density for accessing the detached divertor H mode, and that the required divertor radiated power fraction must increase as the machine size and field increase. The model describes, at least qualitatively, experimental findings on ASDEX Upgrade, DIII-D, JET and JT-60

Three dimensional (3D) divertor particle flux footprints induced by the lower hybrid wave (LHW) have been systematically investigated in the EAST superconducting tokamak during the recent experimental campaign. We find that the striated particle flux (SPF) peaks away from the strike point (SP) closely fit the pitch of the edge magnetic field line for different safety factors q95, as predicted by a field line tracing code taking into account the helical current filaments (HCFs) in the scrape-off-layer (SOL). As LHW power increases, it requires the fuelling to be increased e.g. by super molecular beam injection (SMBI), to maintain a similar plasma density, which may be attributed to the pump-out effect due to LHW, and may thus be beneficial for EAST steady state operations. The 3D SPF structure is observed with a LHW power threshold (PLHW ~ 0.9 MW). The ratio of the particle fluxes between SPF and outer strike point (OSP), i.e. , increases with the LHW power. Upon transition to divertor detachment, the particle flux at the main OSP decreases, as expected, however, the particle flux at SPF continues increasing, in contrast to the RMP-induced striations that vanish with increasing divertor density. In addition, we also find that the in–out asymmetry of the 3D particle flux footprint pattern exhibits a clear dependence on the toroidal field direction (B × ∇ B ↓ and B × ∇ B↑). Experiments using neon impurity seeding show a promising capability in 3D particle and heat flux control on EAST. LHW-induced particle and heat flux striations are also present in the H-mode plasmas, reducing the peak heat flux and erosion at the main strike point, thus facilitating long-pulse operation with a new steady-state H-mode over 60 s being recently achieved in EAST.

The design for the BPX power and particle handling system consists of a double-null (DN) poloidal divertor with an inner-wall bumper limiter. In addition, a substantial portion of the remainder of the inside of the torus is covered with graphite tiles to withstand high heat flux. In divertor operating modes, the diverted plasma is swept across the divertor plates to reduce the time-average heat loads on the plates. The poloidal divertor is designed both to provide access to H-mode operation and to ensure high recycling in the divertor region, which minimizes impurity production and consequent contamination of the main plasma. The device is capable of both single-null (SN) and DN divertor operation. In addition, an inner bumper limiter is designed to provide an alternative to divertor operation. Additional toroidally and/or poloidally localized limiters are provided on the outboard wall for plasma startup and for protection of the radio-frequency (RF) antennas. In this paper the distribution of plasma power on the first-wall components is discussed. Power handling and the choice of material for the divertor plates and limiters is discussed. A discussion is included of the surface temperature limits and erosion mechanisms. The effect of disruptions on the plasma-facing components follows.more » Finally, tritium retention and particle pumping issues related to the choice of graphite as the first-wall material are discussed.« less

Development of Langmuir probe array for the new lower tungsten divertor in EAST

Calculations of the profiles along the field lines within the divertor and scrape-off layer (SOL) of differences in the plasma ion density, temperature, parallel current, parallel flow velocity, and electrostatic potential, which result from using different poloidal distributions of the particle and heat influxes crossing the separatrix from the core plasma into the SOL, are presented and discussed vis-à-vis experimental observations. The calculations show that the stronger outboard than inboard particle and heat fluxes into the SOL caused by the geometric compression/expansion of flux surfaces predicted by magnetohydrodynamic equilibrium calculations lead to a prediction of higher plasma temperature at the outboard divertor than at the inboard divertor, a result that is consistent with experimental observation and that confirms a previous prediction (made without accounting for drifts) of a possible cause of the observed in-out divertor power asymmetry. The calculations also illustrate the effect of the poloidal distribution of particle and power influx into the SOL on the flow velocity, parallel current, and electrostatic potential distributions in the SOL and divertor.

A radiative divertor experiment with neon gas puff was carried out with the aim of investigating impurity behaviour, after the open divertor was modified to the pumped W-shaped divertor. To evaluate neon radiation, we analysed lines from the neon ion measured with a VUV spectrometer in the divertor plasma. As a result, lines from Ne IV-Ne VIII were identified. By combining the measurement of an absolutely calibrated multichannel interference filter spectrometer with that of the VUV spectrometer, the radiation loss through neon line emission was estimated in the divertor plasma. In the case of pumping-off, the detached plasma evolved into multifaceted asymmetric radiation from the edge (MARFE). It was observed that the line intensities of highly ionized neon (Ne VII, Ne VIII) increased simultaneously with the formation of the MARFE and furthermore increased after the MARFE formation. It is considered that after the divertor plasma was detached, the plasma flow velocity and the friction force were weaker in the case of pumping-off than in the case of pumping-on and more impurities moved toward the X-point region. These observations imply impurity back flow from the divertor upstream by the reduction of friction force.

Measurement and modeling of the DIII-D divertor plasma

The realization of fusion energy requires materials that can withstand high heat and particle fluxes at the plasma material interface. In this work, an electrothermal (ET) plasma source has been designed as a transient heat flux source for a linear plasma material interaction device. An ET plasma source operates in the ablative arc regime driven by a DC capacitive discharge. The current channel width is defined by the 4 mm bore of a boron nitride liner. At large plasma currents, the arc impacts the liner wall, leading to high particle and heat fluxes to the liner material, which subsequently ablates and ionizes. This results in a high density plasma with a large unidirectional bulk flow out of the source exit. The pulse length for the ET source has been optimized using a pulse forming network to have durations of 1 and 2 ms. The peak currents and maximum source energies seen in this system are 1.9 kA and 1.2 kJ for the 2 ms pulse and 3.2 kA and 2.1 kJ for the 1 ms pulse, respectively. This work is a proof of the principal project to show that an ET source produces electron densities and heat fluxes comparable to those anticipated in transient events in large future magnetic confinement fusion devices. Heat flux, plasma temperature, and plasma density were determined for each shot using infrared imaging and optical spectroscopy techniques. This paper will discuss the assumptions, methods, and results of the experiments.

INTOR divertor in a realistic 2-D geometry

A zero-dimensional time dependent computer model is developed to describe post implosion behavior of a radially diffuse plasma column in linear theta pinch machines. The model consists of five first-order differential equations for electron temperature Te, ion temperature Ti, plasma column area Ap, magnetic field embedded in the plasma column Bi, and number of electrons in the plasma column N, and incorporates the effects of particle end loss, electron and ion thermal conduction, magnetic compression of the plasma, electron-ion energy transfer, Ohmic heating of the electrons, and magnetic field diffusion. End-on interferometric experimental data, which indicate that particle confinement time scales as L[mi/(Te+Ti)]1/2, where mi is the ion mass and L is the compression coil length, are used as a guide to the choice of an end loss parameter. Numerical predictions of on-axis density and plasma temperature are found to be in reasonably good agreement with experimental results from both the collisional Scylla I-C experiment and the hot collisionless Scylla IV-P experiment. These comparisons are used to identify important physical mechanisms characterizing the behavior of a particular machine including: the significance of thermal conduction on both plasma temperature and particle end loss; the importance of magnetic field diffusion; and the impact of using a time independent end loss scaling parameter.

Calculations have been performed which demonstrate the possibility of operating poloidal divertors at high densities and low temperatures. Ionization of recycling neutral gas near the divertor neutralizer plate amplifies the input particle flux thereby raising the plasma density and lowering the plasma temperature. Low-temperature, high-density operation should reduce the erosion rate of the divertor walls and plate, and ease the pumping requirements in future large tokamaks.

Bohm criterion and plasma particle/power exhaust to and recycling at the wall

The influence of electric fields and neutral particles on the plasma sheath at ITER divertor conditions

During the first test divertor campaign of the stellarator experiment Wendelstein 7-X (Pedersen et al 2022 Nucl. Fusion 62 042022), OP1.2b, 13 neutral gas pressure gauges collected data in different locations in the plasma vessel, enabling a detailed analysis of the neutral gas pressures, the compression ratios and the particle exhaust rates via the turbomolecular pumps in the different magnetic field configurations. In Wendelstein 7-X, the edge magnetic islands are intersected by the divertor target plates and used to create a plasma-wall interface. As the number and position of the magnetic islands varies in different magnetic field configurations, the position of the strike line on the target plates and thus the neutral gas pressure in the subdivertor differs between the configurations. Neutral gas pressures on the order of few 10−4 mbar were measured in the subdivertor region. The highest neutral gas pressure of mbar was obtained in the so-called high iota configuration featuring four edge magnetic islands per cross section. The neutral particle flux through the pumping gaps into the subdivertor volume was provided by EMC3-EIRENE simulations and allowed to analyze the relation between the particle flux entering the subdivertor and the pressure distribution in the subdivertor. Finite element simulations in ANSYS provide a detailed picture of the pressure distribution in the subdivertor volume and agree with the neutral gas pressure measurements in the subdivertor in the standard configuration featuring an island chain of 5 edge magnetic islands. Surprisingly high neutral gas pressures that were not predicted by the simulation were measured in the subdivertor region away from the main strike line for discharges in the most used magnetic configuration, the standard configuration. While the pressure ratio between the two sections of the subdivertor volume, the low and high iota section is 0.06 in high iota configuration, a ratio of 2–5 was obtained in the other configurations, indicating significant particle loads and exhaust rates on the high iota section of the subdivertor in magnetic configurations with the main strike line on the low iota divertor targets.