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 optimized for co-injection by the foreseen high energy 510 keV NBI injector. With the previous constraints the definition and optimization of the divertor shape to maximize performance and flexibility has been done by an extensive power exhaust modelling with the 2D edge codes SOLDGE2D-EIRENE[4] and SOLPS- ITER[5]. To be relevant in terms of plasma parameters divertor optimization has been done at the maximum additional power presently foreseen for DTT (45 MW), the toroidal field and plasma current achievable for the various configurations and a density corresponding to a Greenwald fraction of about 0.5 in the SND case. Transport profiles have been validated in present experiments and tuned to provide a ?q in agreement with available scaling laws. Divertor performances have been considered not only in terms of plasma parameters but also for pumping capability to allow reliable operation in all conditions and in terms of an easy control of the configuration. Considering the technological limits imposed on divertor shape by W monoblocks, it has been found that the best performances (lowest impurity content to achieve detachment by impurity seeding) between all configurations can be obtained with a wide flat divertor using the dome as a third target and for neutral compression (fig. 1). At the maximum additional power the optimized wide flat divertor allows operation in divertor detached conditions in all magnetic configurations with a low core contamination. With the foreseen divertor pumping system gas throughput will allow in SND an easy control of deuterium and seeding gas content also in presence of strong gas-puffing requests for ELMs control or ICRH coupling, control will be possible also in the alternative configurations although in a narrow operating range.