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This paper discusses the criteria to be used in the preliminary design phases of the EU-DEMO reactor to ensure the performance of the divertor without compromising the stability of core plasma or the fusion power generation. This work refers to a lower single null conventional divertor using actively cooled solid metal plasma-facing components and with extrinsic seeding for heat flux dissipation, which is the solution currently being adopted for EU-DEMO. The analysis does not consider the role of edge localised modes, and also neglects major off-normal events like disruptions. It is shown that it is necessary to fulfil two high-level requirements, namely: (i) the concentration of seeded impurities has to be lower than some critical value in order to not compromise the fusion plasma performance or stability and (ii) damage to the divertor plate in the case of accidental plasma reattachment must be avoided for a sufficiently long time in order to ensure safe, controlled termination of the plasma discharge. These requirements are needed because in a device like EU-DEMO, other strategies relying on mass injection are considered more likely to cause a loss of plasma stability at full current, with dramatic consequences for the integrity of plasma facing components. Two figures of merit, corresponding to these criteria, have been identified in the existing literature and discussed. The dependence of such figures of merit on the relevant machine parameters (major radius and toroidal magnetic field) is analysed. Initially, the analysis is carried out using a simple 0D physics approach and subsequently by means of the systems code PROCESS, which allows for consideration of further parameters, such as aspect ratio and elongation. The main conclusion of the present work is that the simultaneous fulfilment of both requirements limits the viable EU-DEMO size both in terms of major radius and in terms of toroidal magnetic field . Finally, an attempt to extend the EU-DEMO related conclusions to a more general level is made. [ABSTRACT FROM AUTHOR]

An updated and improved version of the 0D divertor and scrape-off layer (SOL) model published in Siccinio et al (2016 Plasma Phys. Control. Fusion58 125011) was coupled with the 1.5D transport code ASTRA (Pereverzev 1991 IPP Report 5/42, Pereverzev and Yushmanov 2002 IPP Report 5/98 and Fable et al 2013 Plasma Phys. Control. Fusion55 124028). The resulting numerical tool was employed for various scans in the major radius R and in the toroidal magnetic field BT—for different safety factors q, allowable loop voltages Vloop and H factors—in order to identify the most convenient choices for an electricity producing tokamak. Such a scenario analysis was carried out evaluating self-consistently, and simultaneously, the core profile and transport effects, which significantly impact on the fusion power outcome, and the divertor heat loads, which represent one of the most critical issues in view of the realization of fusion power plants (Zohm et al 2013 Nucl. Fusion53 073019 and Wenninger et al 2017 Nucl. Fusion57 046002). The main result is that, when divertor limits are enforced, the curves at constant electrical power output are closed on themselves in the plane, and a maximum achievable power exists—i.e. no benefits would be obtained from a further increase in R and BT once the optimum is reached. This result appears as an intrinsic physical limit for all those devices where a radiative SOL is needed to deal with the power exhaust, and where a lower limit on the power crossing the separatrix (e.g. because of the L–H transition) is present. [ABSTRACT FROM AUTHOR]

A finite radial temperature gradient has been observed to be maintained within magnetic islands in gyro-kinetic turbulence simulations despite the fast motion along the field, and is related to the trapped particles which do not move along the field around the island. Recent calculations of the interaction of drift-wave turbulence with magnetic islands have shown that turbulence can exist within the separatrix, which in turn allows only the partial flattening of electron temperature profiles. Here we calculate, using a minimal drift-kinetic model, the effect on the bootstrap current in a tokamak. Consequences for the stability of the neoclassical tearing mode are discussed. [ABSTRACT FROM AUTHOR]