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In the European Roadmap towards thermonuclear fusion power production, studying the controlled exhaust of energy and particles from a fusion reactor is a top priority research item. This is the main goal of the Divertor Tokamak Test (DTT) facility, a D-shaped superconducting tokamak (R = 2.19 m, a = 0.70 m, BT

Fusion-fission hybrid reactors (FFHR), which consist of a neutron-producing fusion core surrounded by a fission blanket, keep a sustained interest because of their potential to address energy production before the availability of pure fusion reactors and also because of their capability of fuel supply and waste management. The U.S. Department of Energy (DOE) sponsored a workshop in 2009 in order to assess the potentiality of fusion-fission reactors and, among the others alternative to the Tokamak as fusion core, the Reversed Field Pinch (RFP) was considered, even if not on the same timeline of the Tokamak. On the other hand, the interest in the RFP as fusion core resides, inter alia, in its possibility to reach ignition in a pure ohmic way with a "light" toroidal field winding, which simplifies significantly the hybrid reactor layout and operation and strongly decreases costs with respect to other fusion core solutions. New studies are now in progress in order to revisit the status and the potentiality of the RFP as fusion core in FFHR, taking into account (i) the recent progress in RFP physics brought by the results of the RFX-mod experiment (R=2, a=0.46, Ip=2MA), mainly about MHD modes control, (ii) the expected performance improvements resulting from the ongoing upgrade of the machine. Starting from RFX-mod results and scaling laws, and considering the estimated improvements of the performances derived from the machine enhancements underway, the possibility to realize a preliminary pilot experiment in which a RFP is the neutron fusion source is analysed. The research needs for this advanced solution are updated with respect to the 2009 workshop.

The main mission of the Divertor Test Tokamak (DTT) is to explore viable solutions to the power exhaust issues in a fusion reactor. The ultimate goal will be to qualify and control in various divertor configurations DEMO relevant heat flux densities to the wall while preserving the integrity of both the plasma facing components and the plasma performance. Experiments will involve tailored magnetic topologies, highly radiative regimes and advanced materials. In this contribution, we describe the package of diagnostic systems that will be deployed to allow DTT accomplishing its tasks. Diagnostics and feedback control are particularly functional to the need of maintaining the plasma close to equilibrium in situations prone to instabilities where the plasma wall interaction is optimized. Focusing on the divertor diagnostics, a particular effort will consist in obtaining space resolved measurements of density, temperature, impurity densities and ionization front that can be compared with the 2D results of model simulations. This will be possible using mainly optical diagnostics based on filtered cameras and spectrometers deployed in imaging mode, complemented by local measurements obtained by a space resolved Thomson scattering, Langmuir probes and gas puffing imaging systems. Diagnostics of the main plasma will assure a full qualification of the core and of the pedestal in terms of thermal contents, equilibrium, fast particles densities, impurities, reactivity levels and turbulence. Strongly oriented to the exploration of control methods suitable for DEMO, DTT will also address the study of control systems based on physics and engineering models, which in DEMO are expected to take over the role of the diagnostics deemed to be incompatible with the harsh environment of a fusion reactor.

A hybrid fusion-Mssion (HFF) reactor based on a Reversed Field Pinch (RFP) conMguration looks as an attractive option from both a technical (simple design, easy machine assembly and maintenance) as well as economic perspective (low investment costs due the absence of large Heating and Current Drive systems and superconductive toroidal Meld coils). The hybrid reactor studied here has a RFP fusion core generating 40MWth fusion power via D-T reactions. The reactor size (R=6 m, a =1), its components design and plasma performances have been extrapolated from the results of RFX-mod, the largest RFP experiment built so far. With a plasma temperature of 9.6 keV and an ohmically driven plasma current of 20 MA, the fusion reactions produce 2.2E+19 n/s which are used to both breed tritium and induce Mssion reactions. A breeding blanket is located between the superconductive central solenoid and the torus. Being composed of a lithium-lead eutectic mixture, it also acts as neutron shield. Instead, in the conMguration studied here, the circular area around the torus (16 cm width) is half dedicated to further breed tritium while the rest is occupied by the Mssion blanket. The three sectors (~3.5 m radial extension) containing Pu+MA (60%)-Zr (40%) rods embedded in solid lead for nuclear transmutation (keff~0.94, P~200 MWth) are spaced out by two intermediate regions (~1.75 m radial extension) that can be used for inducing fast and slow transmutations. The results corroborate the interest for the RFP as a HFF reactor due to the simple design of the fusion reactor, the high accessibility to the blankets for maintenance and the multiple applications of the neutron `ux (transmutation of plutonium, minor actinides and nuclear Mssion products, radiopharmaceuticals production) favoured by the blanket modularity.

This paper presents a methods for the intrinsic impurity concentration measurements by means of VUV and SXR diagnostics on the JET-ILW tokamak. Measurements of mid-Z impurities content were obtained by means of VUV spectra. To provide absolute concentrations a new relative calibration technique has been proposed. It's based on cross-calibration with calibrated spectrometer by using unresolved transition array of W in the relevant wavelength range. The SXR cameras were used to deduce W profiles and poloidal asymmetries. Focus is given to hybrid discharges stopped by the real-time vessel protection system due to hot-spots formation. This effect was linked to the application of ICRH power. Local D2 gas injection allows mitigating hot-spot formation and run pulses with acceptable temperature values on vessel components. Hot spot temperature analysis showed a lower maximum temperature at higher gas rate. A decrease of impurity concentration with D2 injection rate was observed. Changes in the plasma current have a strong impact on the plasma-wall interaction, both via modifications in the edge density and in the fast-ion losses. Finally it was observed that at constant gas injection rate, both the hot spot temperature and the core impurity content decrease with the separatrix density.

The planned upgrade of the RFX-mod device is a good opportunity to widen the operational space of the machine, in both RFP and tokamak configurations. Installation of a power neutral beam injector (NBI) is also envisaged and a NBI system compatible with RFX-mod is already available on site. It was previously installed in TPE-RX (Tsukuba, Japan), it has a nominal power of 1.25 MW, a nominal current of 50A and it can operate at a voltage of 25kV for 30ms (or 15 kV for 100ms). A porthole in the modified vacuum toroidal support structure is planned to be adapted to host this injector, which would operate in the equatorial plane and in a perpendicular direction with respect to the plasma column. This geometry is the only one possible due to mechanical constraints, mainly linked to toroidal field coils configuration.In this work, the METIS simulator is used to study the feasibility of TPE-RX injector integration in RFX-mod tokamak plasmas. METIS code allows the simulation of a full tokamak discharge in a time of the order of a few minutes by using relations coming from scaling laws coupled with simplified source models. The neutral beam injection in METIS is described by a decay equation applied in a simplified geometry and an analytical solution of the Fokker-Planck equation. In order to estimate a set of requirements for an NBI system to be applied to RFX-mod tokamak plasmas the beam shine-through has been carefully considered for a wide range of scenarios. Finally some indications of the physical effects to be expected from NBI-plasma coupling will be highlighted, with special attention to the beam energy absorption (both in terms of time dependent quantities and of spatial profiles) and to the possibility of inducing a transition from L to H confinement modes.

In tokamaks, magnetohydrodynamic (MHD) instabilities are frequently observed to have a strong impact on the evolution of the plasma discharge. It is, therefore, important to properly account for MHD activity when modelling the tokamak plasma behaviour. In 2011 the inner wall of JET was covered with Tungsten and Beryllium tiles, replacing the previous Carbon wall with the ITER Like Wall (ILW). In some discharges with ILW, Tungsten accumulates in the core of the plasma, leading to plasma radiation collapse and ultimately disruption. This accumulation seems exacerbated by the presence of a magnetic island [1]. Among the various MHD instabilities, a magnetic island can be detected by magnetic coils placed within the torus wall and by temperature measurements provided by Electron Cyclotron Emission (ECE) diagnostic. The island appearance time and position can be measured by exploiting the correlation between the fluctuations of coil and ECE signals. The island width can be determined by comparing the first and the second harmonic of ECE fluctuations [2]. We use JETTO (a one-and-a-half-dimensional transport code calculating the evolution of plasma parameters in a time dependent axisymmetric MHD equilibrium configuration) in interpretive mode and its impurity component SANCO in predictive mode with transport coefficients calculated by the codes NEO and GKW [1]. We model a discharge with both an initially off-axis Tungsten peak and a (3, 2) island. The island is modelled enhancing Tungsten diffusion, in a position, time instant and with a width determined by diagnostics as previously detailed [3]. The output of the simulations is studied in different ways: reconstructed SXR emissivity, line integrated emissivity profiles, time evolution of SXR channels intensity, Tungsten density and radiation. When opportune, comparison between simulated and experimental data is performed. Although not reproducing quantitatively the experimental result, the simulated emissivity at the final state reproduces qualitatively the trend of the experimental one. In particular if the island is situated in the inner, convex part of the Tungsten density gradient, the impurity is channelled toward the inner core as observed in experiments [4]. [1] Angioni C. et al., Nucl. Fusion 54, 083028 (2014) [2] Baruzzo M., et al., Plasma Phys. Control. Fusion 52 075001 (2010) [3] Marchetto C., et al., 41st EPS Conference on Plasma Physics, ECA Vol. 38, P1.018 (2014) [4] Hender T. C., et al., 41st EPS Conference on Plasma Physics, ECA Vol. 38, P1.011 (2014)

The hydrogen campaign completed at JET in 2016 has demonstrated isotope ratio control in JET-ILW using gas puffing and pellets for fuelling, Neutral Beam Injection alone or in combination, with D/H spectroscopy as a diagnostic. The plasma properties such as confinement, L-H threshold, density limit depend on the isotope composition. The L-H transition power increases with the hydrogen concentration with a wide plateau in the range 0.2