Design Of The Divertor For The DTT Facility Optimized For Power Exhaust Experiments

The divertor tokamak test facility (DTT) [1] is a high field (Bt=6 T) high plasma current (Imax=5.5 MA) medium size (R/a=2.19/0.70 m) long pulse (tpulse?100 s) superconducting device presently under construction to study power exhaust solutions in regimes as close as possible to those foreseen in a fusion reactor in terms of power crossing the separatrix, Psep/R, collisionality and heat flux decay length, ?q. It is well know that the achievement of good core energy confinement by the development of a high pedestal pressure at the separatrix in the standard H-mode positive triangularity Single Null Divertor (SND) configuration provides two main issues in terms of power and particle exhaust: the short ?q gives a small radiating volume and a small wetted divertor area to dissipate the high level of power that must flow through the separatrix to allow the H-mode conditions; the unstable equilibrium between the continue rise of pressure and the destabilization of peeling-ballooning modes produces huge transient energy release during the type-I ELMs. The combined result of previous phenomena indicates that divertor targets made by tungsten monoblocks can be unable to provide a solution scalable towards the realization of the fusion reactor: due to the possible tungsten core contamination in presence of target erosion by the high temperature of the plasma at the targets or in the extreme case by melting of monoblocks. To provide a solution to the power exhaust problem many different solutions have been considered, they can be summarized along three different paths: the development of plasma regimes for SND characterized by low or negligible ELMs amplitude; the development of alternative divertor magnetic configurations (ADCs) able to dissipate and spread the exhaust on a wide volume and divertor area such as to provide a negligible interaction between plasma and divertor material; the use of liquid metal divertors (like thin). More specifically in terms of alternative SND regimes it has been considered [2]: the H-mode Resonant Magnetic Perturbations (RMP), high radiation ELMs Buffering, EDA (Enhanced D- Alpha H-mode), QCE (Quasi Continuous Exhaust), QH-Mode (Quiescent H-Mode) and the I- mode and Negative Triangularity (NT) which prevent the rising edge pedestal and consequent ELMs providing in a different way a good core energy confinement. On the other sides in terms of magnetic configurations alternative to the SND it has been considered [3]: the flux flaring towards the target (X divertor), the increasing of the outer target radius (Super-X divertor) and the movement of a secondary x-point inside the vessel (X-point target) as well as the entire range of Snowflake (SFD+/SFD-) configurations and the presently reconsidered double null (DND) one. Previous list of possible solutions for the power exhaust clearly shows the large number of plasma regimes and magnetic configurations that the divertor tokamak test facility (DTT) in principle should be able to explore to contribute to the selection of a solution for the future fusion reactor. To provide as soon as possible useful results easily exploitable to the design of the DEMO reactor a few preliminary decision have been taken for the first DTT divertor: it will use the ITER-like technology based on full tungsten monoblocks bonded on CuCrZr cooling tubes; the divertor has to able to test in good conditions the standard SND and with high priority some of the most promising ADCs, like the X divertor (XD) and the "hybrid Super-X/long leg SN" but not excluding the possibility to test also the SnowFlake (SF) one. Additionally, the negative triangularity (NT) operation is considered important as well the possibility to study I-mode plasma without reversing the sign of plasma current to be opti