Demonstrating improved confinement of energetic ions is one of the key goals of the Wendelstein 7-X (W7-X) stellarator. In the past campaigns, measuring confined fast ions has proven to be challenging. Future deuterium campaigns would open up the option of using fusion-produced neutrons to indirectly observe confined fast ions. There are two neutron populations: 2.45 MeV neutrons from thermonuclear and beam-target fusion, and 14.1 MeV neutrons from DT reactions between tritium fusion products and bulk deuterium. The 14.1 MeV neutron signal can be measured using a scintillating fiber neutron detector, whereas the overall neutron rate is monitored by common radiation safety detectors, for instance fission chambers. The fusion rates are dependent on the slowing-down distribution of the deuterium and tritium ions, which in turn depend on the magnetic configuration via fast ion orbits. In this work, we investigate the effect of magnetic configuration on neutron production rates in W7-X. The neutral beam injection, beam and triton slowing-down distributions, and the fusion reactivity are simulated with the ASCOT suite of codes. The results indicate that the magnetic configuration has only a small effect on the production of 2.45 MeV neutrons from DD fusion and, particularly, on the 14.1 MeV neutron production rates. Despite triton losses of up to 50 %, the amount of 14.1 MeV neutrons produced might be sufficient for a time-resolved detection using a scintillating fiber detector, although only in high-performance discharges.

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.

Density limit investigations on ASDEX have been performed under a variety of conditions: ohmically heated and neutral injection heated plasmas in H2, D2 and He have been studied in different divertor configurations, after various wall coating procedures, with gas puff and pellet fuelling, and in different confinement regimes with their characteristically different density profiles. A detailed description of the parametric dependence of the density limit, which in all cases is a disruptive limit, is given. This limit is shown to be a limit to the density at the plasma edge. Therefore, the highest densities corresponding to neRqa/Bt>30*1019 m-2.T-1 are obtained with centrally peaked ne profiles. Radiation from the main plasma at the density limit is always significantly below the total input power. The plasma disruption is due to an m=2 instability which for medium and high qa is preceded by one or more minor disruptions. In this range of qa, the disruptive instability is initiated by the occurrence of a Marfe on the high field side as a consequence of strong plasma cooling in this region. The duration of the Marfe increases with increasing distance between the plasma edge and the q=2 surface. After penetrating onto closed flux surfaces the Marfe leads to a current contraction and a subsequent destabilization of the m = 2 mode. In helium plasmas a strongly radiating, poloidally symmetric shell is observed before the density limit instead of a Marfe. An instantaneous destabilization of this mode is observed at low qa. Detailed measurements of plasma edge and divertor parameters close to the density limit indicate the development of a cold, dense divertor plasma before the disruption. Models describing the scrape-off layer and the divertor region predict an upper limit to the edge density at low divertor temperatures according to power balance considerations. Their relations to the experimental findings, especially the low field side cooling, ar

The fundamental behavior of the W7-X island divertor under detached conditions, which has been theoretically predicted with the EMC3-Eirene code, is re-examined here under the experimental conditions achieved so far and compared with the first experimental results. Both simulations and experiments cover a range of divertor configurations and plasma parameters, and show the following common trends: (1) with rising impurity radiation, the target heat load decreases ‘uniformly’ over the entire target surface in the sense that both the peak and average heat loads can drop by an order of magnitude. Impurity radiation (mainly from intrinsic carbon) occurs primarily at the plasma edge and the resulting negative impact on the stored energy is less than 10%. (2) When the total radiation exceeds a critical level, the target particle flux (the recycling flux Γrecy) begins to fall and can drop by a factor of 3–5 at high radiation levels without an obvious indication of significant volume recombination. (3) While Γrecy decreases, the divertor neutral pressure continues to build up and reaches a maximum, at which point Γrecy has declined significantly. (4) During detachment, the electron temperature at the last closed flux surface falls in a way that is not quantitatively understandable from parallel classical heat conduction processes. This paper presents a physical explanation of the numerical/experimental results described above. Furthermore, using the EMC3-Eirene code as a diagnostic tool, we are able, apparently for the first time, to provide a full quantitative analysis of each transport channel in the island divertor, aiming to clarify how the island divertor plasma self-regulates to maintain particle, energy, and momentum balance under detached conditions.

In the previous divertor campaign, the Wendelstein 7-X (W7-X) device injected 3.6 MW of neutral beam heating power allowing for the achievement of densities approaching 2 × 10 20 m −3 , and providing the first initial assessment of fast ion confinement in a drift optimized stellarator. The neutral beam injection (NBI) system on W7-X is comprised of two beam boxes with space for four radio frequency sources each. The 3.6 MW of heating reported in this work was achieved with two sources in the NI21 beam box. The effect of combined electron-cyclotron resonance heating (ECRH) and NBI was explored through a series of discharges varying both NBI and ECRH power. Discharges without ECRH saw a linear increase in the line-integrated plasma density, and strong peaking of the core density, over the