Ignitor: Physics and Progress Towards Ignition

Thermonuclear ignition condition for deuterium-tritium plasmas can be achieved in compact, high magnetic field devices such as Ignitor. The main scientific goals, the underlying physics basis, and the most relevant engineering solutions of this experiment are described. Burning plasma conditions can be reached either with ohmic heating only or with small amount of auxiliary power in the form of ICRH waves, and this condition can be sustained for a time considerably longer than all the relevant plasma time scales. In the reference operating scenario, no transport barriers are present, and the resulting thermal loads on the plasma facing component are estimated to be rather modest, thanks to the high edge density and low edge temperature that ensure an effective intrinsic radiating mantle in elongated limiter configurations. Enhanced confinement regimes can also be obtained in configurations with double X-points near the first wall

Time-extended inductive tokamak discharges with differentially-tilted toroidal field coils

The strong toroidal magnetic field required for plasma confinement in tokamaks is generated by a set of D-shaped coils lying equidistant on meridian planes toroidally located around the central axis of the device. A major technological challenge tied to this configuration is represented by the large Lorentz force acting on the coils and arising from the interaction of the coils’ currents with the magnetic field generated by the coil system itself. As this force is given by the cross product of the coil current and the magnetic field, various kinds of coil geometry modification have been proposed to alleviate this problem, from an inclination of the entire coil in order to maintain its planarity, to azimuthal tilting of all, or parts of, the coil profile. When the inner legs of the coils are tilted, apart from a reduction of the electromagnetic forces, a solenoid-like structure is formed which introduces additional magnetic flux linked to the plasma. Considering compact, high field devices, it is shown that when this additional flux is exploited, totally or in part, to ramp up the plasma current, the discharge time can be extended by a significant amount without resorting to noninductive current drive systems. Operational scenarios with inner-leg-tilted toroidal field coils are presented

An innovative approach for DEMO core fuelling by inboard injection of high-speed pellets

Core fuelling of DEMO tokamak fusion reactor is under investigation within the EUROfusion Work Package “Tritium, Fuelling and Vacuum”. An extensive analysis of fuelling requirements and technologies, suggests that pellet injection still represents, to date, the most realistic option. Modelling of both pellet penetration and fuel deposition profiles for different injection locations, assuming a specific plasma reference scenario and the ITER reference pellet mass (6 × 1021 atoms), indicates that: 1) Low Field Side (LFS) injection is inadequate; 2) Vertical injection may be effective only provided that pellets are injected at ∼ 10 km/s from a radial position ≤∼8 m; 3) effective core fuelling can be achieved launching pellets from the High Field Side (HFS) at ∼1 km/s. HFS injection was therefore selected as the reference scheme, though scenarios featuring less steep density and temperature gradients at the plasma edge could induce to reconsider vertical injection at speeds in the range of 4–5 km/s. To deliver intact pellets at 1 km/s from the HFS, the use of guide tubes with a bend radius ≥6 m is envisaged. The results of above simulations rely on the hypothesis that pellets are delivered at the plasma edge with the desired mass and speed. However, mass erosion and fracturing of pellets inside the guide tube (severely limiting the transfer speed), as well as pressure build up and speed losses at relevant injection rates, might hamper the use of curved guide tubes. An additional innovative approach, aimed at identifying inboard straight “free flight” injection paths, to inject pellets from the HFS at significantly higher speeds, is proposed and discussed as a backup solution. Outboard high-speed injection is still being considered, instead, for JT-60SA

Observations of central toroidal rotation in ICRF heated Alcator C-Mod plasmas

AC02-78ET51013. Reproduction, translation, publication, use and disposal, in whole or in part by or for the United States government is permitted

Modelling of the Ignitor scrape-off layer including neutrals

Ignitor is a tokamak project aimed at achieving ignition. In the reference scenario, plasma-surface interactions are controlled by a Mo first-wall/limiter, which constitutes a simple engineering solution but, at the same time, a special challenge for edge plasma modelling. Here the ASPOEL plasma fluid code, already applied to Ignitor in the recent past, is coupled with the neutral Monte Carlo code EIRENE. We study the effects of the neutrals on the plasma density and temperature profiles in the Ignitor scrape-off layer, and compute the particle and heat loads onto the Ignitor first-wall limiter