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

Presentation of the New SOLPS-ITER Code Package for Tokamak Plasma Edge Modelling

The basic physics of the processes playing the most important role in divertor plasma detachment is reviewed. The models used in the two-dimensional edge plasma transport codes that are widely used to address different issues of the edge plasma physics and to simulate the experimental data, as well as the numerical schemes and convergence issues, are described. The processes leading to ultimate divertor plasma detachment, the transition to and the stability of the detached regime, as well as the impact of the magnetic configuration and divertor geometry on detachment, are considered. A consistent, integral physical picture of ultimate detachment of a tokamak divertor plasma is developed.

/pdf/physics-of-ultimate-detachment-of-a-tokamak-divertor-plasma-54jitr1pmj.pdf

Physics of ultimate detachment of a tokamak divertor plasma

The erosion of tungsten (W), induced by the bombardment of plasma and impurity particles, determines the lifetime of plasma-facing components as well as impacting on plasma performance by the influ ...

Erosion, screening, and migration of tungsten in the JET divertor

The SPARC tokamak project, currently in engineering design, aims to achieve breakeven and burning plasma conditions in a compact device, thanks to new developments in high-temperature superconductor technology. With a magnetic field of 12.2 T on axis and 8.7 MA of plasma current, SPARC is predicted to produce 140 MW of fusion power with a plasma gain of Q ≈ 11, providing ample margin with respect to its mission of Q > 2. All tokamak systems are being designed to produce this landmark plasma discharge, thus enabling the study of burning plasma physics and tokamak operations in reactor relevant conditions to pave the way for the design and construction of a compact, high-field fusion power plant. Construction of SPARC is planned to begin by mid-2021.

https://iopscience.iop.org/article/10.1088/1741-4326/ac1654/pdf

Overview of the SPARC physics basis towards the exploration of burning-plasma regimes in high-field, compact tokamaks

The new SOLPS-ITER code package

Since ITER will be a full metallic wall machine, scattered photons from the strong emission in the divertor may distort the emission from the scrape-off layer (SOL). The influence of stray light for visible spectroscopy of H? and Be?I emissions in ITER was investigated with a ray-tracing simulation. It was found that the stray light would be always more than one order of magnitude greater than the real signal intensity for H? emission from the SOL. For Be?I emission, although the situation was better than the cases of H? measurement, the stray light could be much larger than the real signal from the SOL. In ITER, it will inevitably reduce the stray light somehow for visible spectroscopy. The effect of optical dumps embedded on first walls was investigated with the ray-tracing analysis.

http://iopscience.iop.org/article/10.1088/0741-3335/55/8/085020/pdf

Influence of stray light on visible spectroscopy for the scrape-off layer in ITER

Influence of stray light for divertor spectroscopy in ITER

The retention of deuterium fuel during MAST discharges has been studied using the method of global gas balance. The results show that with inter-shot 4He-GDC, the total number of particles retained for a discharge increases with the total number of injected particles, and the total retention fraction stays very high, and can be almost 100%. However, it is observed that disruptions can reduce the wall inventory. With inter-shot 4He-GDC to recondition surfaces on MAST, the retention for all shots analysed has not been observed to saturate, while if there is no 4He-GDC before a shot, the following plasma density can become uncontrollable owing to lower wall pumping capacity. This indicates that a majority of injected particles is retained in the walls/divertor targets during each discharge and that most of them can be recovered with 4He-GDC, suggesting that on the short time-scale of MAST pulses, the particle balance is dominated by direct implantation of ions and neutral particles in shallow surface layers.

http://iopscience.iop.org/article/10.1088/0741-3335/52/7/075012/pdf

Analysis of fuel retention on MAST by global gas balance

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The new SOLPS-ITER code package

Influence of stray light on visible spectroscopy for the scrape-off layer in ITER

Influence of stray light for divertor spectroscopy in ITER

Analysis of fuel retention on MAST by global gas balance