On the eve of component procurement, this paper discusses the present physics basis for the first ITER tungsten (W) divertor, beginning with a reminder of the key elements defining the overall design, and outlining relevant aspects of the Research Plan accompanying the new “staged approach” to ITER nuclear operations which fixes the overall divertor lifetime constraint. The principal focus is on the main design driver, steady state power fluxes in the DT phases, obtained from simulations using the 2-D SOLPS-4.3 and SOLPS-ITER plasma boundary codes, assuming the use of the low Z seeding impurities nitrogen (N) and neon (Ne). A new perspective on the simulation database is adopted, concentrating purely on the divertor physics aspects rather than on the core-edge integration, which has been studied extensively in the course of the divertor design evolution and is published elsewhere. Emphasis is placed on factors which may increase the peak steady state loads: divertor target shaping for component misalignment protection, the influence of fluid drifts, and the consequences of narrow scrape-off layer heat flux channels. All tend to push the divertor into an operating space at higher sub-divertor neutral pressure in order to remain at power flux densities acceptable for the target material. However, a revised criterion for the maximum tolerable loads based on avoidance of W recrystallization, sets an upper limit potentially ∼50% higher than the previously accepted value of ∼10 MW m−2, a consequence both of the choice of material and the finalized component design. Although the simulation database is currently restricted to the 2-D toroidally symmetric situation, considerable progress is now also being made using the EMC3-Eirene 3-D code suite for the assessment of power loading in the presence of magnetic perturbations for ELM control. Some new results for low input power corresponding to the early H-mode operation phases are reported, showing that even if realistic plasma screening is taken into account, significant asymmetric divertor heat fluxes may arise far from the unperturbed strike point. The issue of tolerable limits for transient heat pulses is an open and key question. A new scaling for ELM power deposition has shown that whilst there may be more latitude for operation at higher current without ELM control, the ultimate limit is likely to be set more by material fatigue under large numbers of sub-threshold melting events.

Physics basis for the first ITER tungsten divertor

During the last decade, significant progress has been made in the field of pellet injection with (1) the identification of the drift of the deposited material in the inhomogeneous magnetic field that opened the possibility of fuelling efficiently the plasmas from the high-field side of the torus, (2) the technique to mitigate ELMS in H-mode discharges with shallow pellet injection at high frequency and (3) with the development of high density, high performance scenarios close to the ITER requirements. Both the experimental and theoretical aspects of this domain are reviewed in this paper.

https://iopscience.iop.org/article/10.1088/0741-3335/49/8/R01/pdf

Review: Pellet injection experiments and modelling

Pellet injection from the inner wall is planned for use in ITER as the primary core fuelling system since gas fuelling is expected to be highly inefficient in burning plasmas. Tests of the inner wall guide tube have shown that 5 mm pellets with up to 300 m s−1 speeds can survive intact and provide the necessary core fuelling rate. Modelling and extrapolation of the inner wall pellet injection experiments from present day's smaller tokamaks leads to the prediction that this method will provide efficient core fuelling beyond the pedestal region. Using pellets for triggering of frequent small edge localized modes is an attractive additional benefit that the pellet injection system can provide. A description of the ITER pellet injection system's capabilities for fuelling and ELM triggering is presented and performance expectations and fusion power control aspects are discussed.

Pellet fuelling and control of burning plasmas in ITER

This paper reports the impact on confinement and power load of the high-shape 2.5 MA ELMy H-mode scenario at JET of a change from all carbon plasma-facing components to an all metal wall. In preparation to this change, systematic studies of power load reduction and impact on confinement as a result of fuelling in combination with nitrogen seeding were carried out in JET-C and are compared with their counterpart in JET with a metallic wall. An unexpected and significant change is reported on the decrease in the pedestal confinement but is partially recovered with the injection of nitrogen.

Impact of nitrogen seeding on confinement and power load control of a high-triangularity JET

Future fusion reactors require a safe, steady state divertor operation. A possible solution for the power exhaust challenge is the detached divertor operation in scenarios with high radiated power ...

Power exhaust by SOL and pedestal radiation at ASDEX Upgrade and JET

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 Alfven 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.

https://pure.mpg.de/pubman/item/item_2137665_3/component/file_3311236/P3_155.PDF

Mass Deposition after Pellet Injection into a Tokamak

Using the new version of the SOLPS plasma boundary code package, SOLPS-ITER, this 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 development of reliable schemes for 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.

https://iopscience.iop.org/article/10.1088/1741-4326/ab73c1/pdf

SOLPS-ITER modelling of ITER edge plasma with drifts and currents

The injection of high-pressure supersonic jets into the tokamak plasma is considered a promising method of future thermonuclear reactor fuelling and as a tool for disruption mitigation. Successful experiments were correspondingly performed on Tore Supra and DIII-D. In the present paper the evolution of such a jet is analysed. 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 4 × 1024 m−3 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 the poloidal (perpendicular to the magnetic field) direction of the jet is sufficient to stop poloidal expansion of the jet by 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 corresponding drift. In the paper two mechanisms of polarization reduction are considered: Alfven conductivity of the ambient plasma and the ∇B-induced drift. It is shown that an 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 injection.

/pdf/penetration-of-supersonic-gas-jets-into-a-tokamak-1ijcjbo80h.pdf

I. Senichenkov

Papers

Physics basis for the first ITER tungsten divertor

Mass Deposition after Pellet Injection into a Tokamak

SOLPS-ITER modelling of ITER edge plasma with drifts and currents

Penetration of supersonic gas jets into a tokamak

Electric fields and currents in the detached regime of a tokamak