Assessment of alternative divertor configurations as an exhaust solution for DEMO

Abstract

Plasma exhaust has been identified as a major challenge towards the realisation of magnetic confinement fusion. To mitigate the risk that the single null divertor (SND) with a high radiation fraction in the scrape-of-layer (SOL) adopted for ITER will not extrapolate to a DEMO reactor, the EUROfusion consortium is assessing potential benefits and engineering challenges of alternative divertor configurations. Alternative configurations that could be readily adopted in a DEMO design include the X divertor (XD), the Super-X divertor (SXD), the Snowflake divertor (SFD) and the double null divertor (DND). The flux flaring towards the divertor target of the XD is limited by the minimum grazing angle at the target set by gaps and misalignments. The characteristic increase of the target radius in the SXD is a trade-off with the increased TF coil volume, but, ultimately, also limited by forces onto coils. Engineering constraints also limit XD and SXD characteristics to the outer divertor leg with a solution for the inner leg requiring up-down symmetric configurations. Capital cost increases with respect to a SND configuration are largest for SXD and SFD, which require both significantly more poloidal field coil conductors and in the case of the SXD also more toroidal field coil conductors. Boundary models with increasing degrees of complexity have been used to predict the beneficial effect of the alternative configurations on exhaust performance. While all alternative configurations should decrease the power that must be radiated in the outer divertor, only the DND and possibly the SFD also ease the radiation requirements in the inner divertor.

Analysis of the performances of a fusion reactor in a reduced H-mode confinement

“…It results lq≈3 mm, as for the present version of the EU DEMO [6]. The reasons for a value higher than that of the present empirical scaling laws [24], lq,scaling≈1 mm, are detailed in [6], and also in [21,22] for the I-DTT device. They can be summarised briefly as follows: i) the scaling is based on a data base with lq>ri,pol, that makes its extrapolation uncertain to DEMO where the opposite is true.…”

“…The cross-field transport coefficients are adjusted in order to give a cross-field e-folding decay length inside the power transport channel, lq, whose magnitude at the outer equator is quite close to that of the ion poloidal gyroradius, ri,pol, as done in [11,14]. It results lq≈3 mm, as for the present version of the EU DEMO [6]. The reasons for a value higher than that of the present empirical scaling laws [24], lq,scaling≈1 mm, are detailed in [6], and also in [21,22] for the I-DTT device.…”

“…The concentration for both needs to be maintained approximately < 0.01% for maintaining the H mode. However, the total power to plate should simultaneously be kept below 100 MW at least, for a technically sustainable heat removal rate, according to an optimistic rough scaling for the larger major radius from recent studies on the EU-DEMO [6]. The figure shows the incompatibility of the two requirements with Xe or Kr only.…”

“…The key points of this study are that the following constraints should hold: 1) fusion gain Q, ratio of the fusion power to that of the auxiliary heating, QºPfus/Paux≥30; 2) power crossing the separatrix PSOL>PL-H, threshold value for the L-H mode back transition [1]; 3) peak load onto either divertor target below the technically manageable level, here assumed qpk,max=10 MW/m 2 , following ITER [1]. Other reference values anyway can range from 5 MW/m 2 , in the prudent approach of EU-DEMO [6], to 15 MW/m 2 for more aggressive views as I-DTT [7] or even beyond in case of liquid metals targets [8]. The plan of the paper is as follows.…”

Global fluid simulation of plasma turbulence in a stellarator with an island divertor

“…Most tokamaks are currently operated in an axisymmetric single-null diverted configuration, consisting of a central region of closed field lines, surrounded by the scrape-off layer (SOL), a region of open field lines that are diverted by using external coils and end up on the solid surfaces of the divertor plates at a certain distance from the core. The single-null configuration will be adopted by ITER [1], while alternative configurations such as the snowflake or the double-null are being explored for DEMO [2]. As for stellarators, diverted solutions also exist.…”