After completing the main construction phase of Wendelstein 7-X (W7-X) and successfully commissioning the device, first plasma operation started at the end of 2015. Integral commissioning of plasma start-up and operation using electron cyclotron resonance heating (ECRH) and an extensive set of plasma diagnostics have been completed, allowing initial physics studies during the first operational campaign. Both in helium and hydrogen, plasma breakdown was easily achieved. Gaining experience with plasma vessel conditioning, discharge lengths could be extended gradually. Eventually, discharges lasted up to 6 s, reaching an injected energy of 4 MJ, which is twice the limit originally agreed for the limiter configuration employed during the first operational campaign. At power levels of 4 MW central electron densities reached 3  ×  1019 m−3, central electron temperatures reached values of 7 keV and ion temperatures reached just above 2 keV. Important physics studies during this first operational phase include a first assessment of power balance and energy confinement, ECRH power deposition experiments, 2nd harmonic O-mode ECRH using multi-pass absorption, and current drive experiments using electron cyclotron current drive. As in many plasma discharges the electron temperature exceeds the ion temperature significantly, these plasmas are governed by core electron root confinement showing a strong positive electric field in the plasma centre.

Since the installation of an ITER-like wall, the JET programme has focused on the consolidation of ITER design choices and the preparation for ITER operation, with a specific emphasis given to the bulk tungsten melt experiment, which has been crucial for the final decision on the material choice for the day-one tungsten divertor in ITER. Integrated scenarios have been progressed with the re-establishment of long-pulse, high-confinement H-modes by optimizing the magnetic configuration and the use of ICRH to avoid tungsten impurity accumulation. Stationary discharges with detached divertor conditions and small edge localized modes have been demonstrated by nitrogen seeding. The differences in confinement and pedestal behaviour before and after the ITER-like wall installation have been better characterized towards the development of high fusion yield scenarios in DT. Post-mortem analyses of the plasma-facing components have confirmed the previously reported low fuel retention obtained by gas balance and shown that the pattern of deposition within the divertor has changed significantly with respect to the JET carbon wall campaigns due to the absence of thermally activated chemical erosion of beryllium in contrast to carbon. Transport to remote areas is almost absent and two orders of magnitude less material is found in the divertor.

In the framework of the DEMOnstration fusion power plant (DEMO) design coordinated by the EUROfusion consortium, a pre-conceptual design of the superconducting magnet system has been developed. For the toroidal field coils (TFCs), three winding pack (WP) options have been proposed; exploring different winding approaches (pancakes vs. layers), and manufacturing techniques (react & wind vs. wind & react Nb 3 Sn). Thermal-hydraulic and mechanical analyses on the three WPs have produced encouraging results, with some critical issues to be solved in future studies and optimizations. The experimental tests on TF prototype short sample conductors have demonstrated a limited performance degradation with electromagnetic cycles and significantly lower effective strains than most of the large-size Nb 3 Sn conductors reported in literature. The toroidal field quench protection circuit has been studied, starting from different topologies and focusing on the most promising one. Two designs are also presented for the central solenoid magnet, with preliminary evaluations on the AC losses during the plasma breakdown. Finally, the design of a TF winding pack based on HTS conductors and the experimental tests on "fusion-relevant" HTS cables are illustrated.

The DEMO reactor is expected to be the first application of fusion for electricity generation in the near future. To this aim conceptual design activities are progressing in Europe (EU) under the lead of the EUROfusion Consortium in order to drive on the development of the major tokamak systems. In 2014 the activities carried out by the magnet system project team were focused on the Toroidal Field (TF) magnet system design and demonstrated major achievements in terms of concept proposals and of consolidated evaluations against design criteria. Several magnet system R&D activities were conducted in parallel, together with broad investigations on High Temperature Superconductor (HTS) technologies. In this paper we present the outcomes of the work conducted in two areas in the 2014 magnet work program: (1) the EU inductive reactor (called DEMO1) 2014 configuration (power plant operating under inductive regime) was the basis of conceptual design activities, including further optimizations; and (2) the HTS R&D activities building upon the consolidated knowledge acquired over the past years

Abstract-The stellarator fusion experimental device Wendelstein 7-X (W7-X) is presently under assembly at the Greifswald branch of the Max-Planck-Institut für Plasmaphysik (IPP), Germany. The superconducting magnet system consists of 50 non planar coils, 20 planar coils, a superconducting bus system and 14 current leads. It is organized in seven electrical circuits with ten coils, the bus system for the interconnection and two current leads each. The magnet system is cooled by supercritical helium. It is enclosed in a cryostat with an outer diameter of 16 meters formed by the plasma vessel and the outer vessel. There are five different types of non planar coils having dimensions of 3.5 x 2.5 x 1.5 meters in maximum and a weight of about 5.5 tons. The two different types of the planar coils are nearly circular coils with a diameter of up to 4.5 meters and a weight of about 3 tons. The superconducting bus bar system connects the coils to each other and provides the connection to the current leads inside the cryostat. All types of coils and the bus system use the W7-X superconductor, a forced flow cable-in-conduit superconductor with 243 copper stabilized NbTi strands with an outer aluminumalloy jacket. The current leads provide the transfer of the electrical current from the room temperature bus bar system outside the cryostat to the superconducting parts inside the cryostat. Their special feature is the upside down orientation with the cold end at the top.