No abstract

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.

The shift of the ionized plasma cloud, which is created in the process of pellet ablation, in the low field side (LFS) direction is calculated. It is demonstrated that the shift is determined, on the one hand, by the acceleration of a cloud in the LFS direction, and, on the other hand, by the reduction of the cloud polarization. The latter is caused by the generation of Alfvén waves and compensation of the ∇B drifts in different parts of the cloud propagating along the magnetic field. Dissipative mechanisms of polarization reduction turned out to be less significant. An expression for the cloud displacement may be used to specify the initial density profile for simulation using one-dimensional transport codes. The calculated shift of the plasma cloud is consistent with the observations in ASDEX-Upgrade and DIII-D.

Using the new version of the SOLPS plasma boundary code package, SOLPS-ITER, the paper presents the first ever simulations of the ITER burning baseline H-mode edge plasma with drifts and currents activated. Neon (Ne) seeded discharges for divertor power dissipation are considered. The results for divertor and scrape-off layer (SOL) parameters with and without drifts are compared, both for the SOLPS-ITER simulations and against the earlier SOLPS-4.3 modelling (which did not include a drift description) constituting the bulk of the existing ITER divertor simulation database. Whereas the drift effect on the equatorial midplane (main chamber) density and temperature profiles is moderate, drifts increase the peak heat flow to the outer divertor target. This effect is more pronounced for regimes with low sub-divertor neutral pressure, when even drift-free SOLPS4.3 simulations find strong out-in target power asymmetries. An important conclusion is thus that if ITER operates as expected with partially detached divertor targets, drifts should not influence the power handling, but that in the case of divertor reattachment, they will act to worsen the target loading, increasing the need for vigilance in detachment control. Comparing SOLPS-4.3 and SOLPS-ITER results for the key peak target heat flux versus sub-divertor neutral pressure operating domain, SOLPS-ITER with drifts predicts a narrower operational window for the divertor pressure.

The existing Globus-M machine [1] is a low aspect ratio compact tokamak (R = 0.36 m, a = 0.24 m) with high specific ohmic and auxiliary heating power. First plasma was achieved in Globus-M in 1999. The machine has demonstrated practically all of the project objectives ever since. Target design parameters (aspect ratio-1.5, 2 − X-point configuration, vertical elongation-2.2, traiangularity-0.45, average density-1.0•10 20 m −3 , plasma current-0.3 MA, toroidal beta-12%, auxiliary heating power-1 MW) [2] were achieved and some of them overcame [3,4]. Also Globus-M

The injection of high-pressure supersonic jets into the tokamak plasma is considered as a promising way of future thermonuclear reactor fueling and as a tool for disruption mitigation. Successful experiments were performed on Tore Supra and DIII-D, correspondingly. In the present paper the evolution of such a jet is analyzed. The jet expansion, deceleration of the ambient electrons and ions by the jet, self-consistent electric field, elementary processes, radiation and adiabatic cooling of the ambient plasma are taken into account. The jet is simulated by a MHD code, which is similar to the code previously used for pellets. It is demonstrated that the ionization degree of the jet strongly depends on the jet parameters. Several simulations were performed for the range of parameters typical for DIII-D. The jet of initial density remains almost neutral, and only the outer regions are ionized. When the initial jet density is reduced by a factor of 2 or more the main part of the jet becomes ionized rather fast. It is demonstrated that ionization at the jet edge in poloidal (perpendicular to the magnetic field) direction of the jet is sufficient to stop poloidal expansion of the jet by -3 24 m 10 4 ⋅ B j r r × force. The final poloidal size of the jet remains of the order of its initial poloidal dimension (of the order of ten centimetres). The jet motion in the radial direction (direction of the injection) is provided by the polarization poloidal electric field and the correspoding B E r r × drift. In the paper two mechanisms of polarization reduction are considered: Alfvén conductivity of the ambient plasma and the B ∇ -induced drift. It is shown that almost neutral jet can penetrate deep into the tokamak while a modest ionization degree should prevent its penetration for the case of low field side (LFS) injection.

The structure of currents and electric fields in the edge tokamak plasma is analysed for detached regimes of a tokamak. Strong electric fields in the divertor regions and near the X‐point are typical for detached regimes. It is shown that the value of the poloidal electric field in the regions with the temperature 1–2 eV is determined by the Spitzer conductivity and parallel currents. The latter are a combination of thermal and Pfirsch–Schlüter (PS) currents. The E × B drifts have a significant impact on the distribution of the plasma in the divertor. Stability of the detached regimes with strong electric fields is also analysed. It has been demonstrated that the so‐called PS instability caused by PS current can develop in the cold low‐temperature regions of the detached divertor.

SOLPS-ITER, a comprehensive 2D scrape-off layer modeling package, is used to examine the physical mechanisms that set the scrape-off width ( q λ ) for inter-ELM power exhaust. Guided by Goldston's heuristic drift (HD) model, which shows remarkable quantitative agreement with experimental data, this research examines drift effects on q λ in a DIII-D H-mode magnetic equilibrium. As a numerical expedient, a low target recycling coefficient of 0.9 is used in the simulations, resulting in outer target plasma that is sheath limited instead of conduction limited as in the experiment. Scrape-off layer (SOL) particle diffusivity (D SOL ) is scanned from 1 to 0.1 m 2 s −1 . Across this diffusivity range, outer divertor heat flux is dominated by a narrow (∼3-4 mm when mapped to the outer midplane) electron convection channel associated with thermoelectric current through the SOL from outer to inner divertor. An order-unity up-down ion pressure asymmetry allows net ion drift flux across the separatrix, facilitated by an artificial mechanism that mimics the anomalous electron transport required for overall ambipolarity in the HD model. At D 0.1 SOL = m 2 s −1 , the density fall-off length is similar to the electron temperature fall-off length, as predicted by the HD model and as seen experimentally. This research represents a step toward a deeper understanding of the power scrape-off width, and serves as a basis for extending fluid modeling to more experimentally relevant, high-collisionality regimes.