Search

After a long device enhancement phase, scientific operation resumed in 2022. The main new device components are the water cooling of all plasma facing components and the new water-cooled high heat flux divertor units. Water cooling allowed for the first long-pulse operation campaign. A maximum discharge length of 8 min was achieved with a total heating energy of 1.3 GJ. Safe divertor operation was demonstrated in attached and detached mode. Stable detachment is readily achieved in some magnetic configurations but requires impurity seeding in configurations with small magnetic pitch angle within the edge islands. Progress was made in the characterization of transport mechanisms across edge magnetic islands: Measurement of the potential distribution and flow pattern reveals that the islands are associated with a strong poloidal drift, which leads to rapid convection of energy and particles from the last closed flux surface into the scrape-off layer. Using the upgraded plasma heating systems, advanced heating scenarios were developed, which provide improved energy confinement comparable to the scenario, in which the record triple product for stellarators was achieved in the previous operation campaign. However, a magnetic configuration-dependent critical heating power limit of the electron cyclotron resonance heating was observed. Exceeding the respective power limit leads to a degradation of the confinement. more...

Combined high-fusion performance and long-pulse operation is one of the key integration challenges for fusion energy development in magnetic devices. Addressing these challenges requires an integrated vision of physics and engineering aspects with the purpose of simultaneously increasing time duration and fusion performance. Significant progress has been made in tokamaks and stellarators, including very recent achievement in duration and/or performance. This progress is reviewed by analyzing the experimental data (109 plasma pulses with a total of 3200 data points, i.e. on average 29 data per pulse) provided by ten tokamaks (in alphabetical order: Axially Symmetric Divertor Experiment Upgrade, DIII-D, Experimental Advanced Superconducting Tokamak, Joint European Torus, JT-60 Upgrade, Korea Superconducting Tokamak Advanced Research, tokamak à configuration variable, Tokamak Fusion Test Reactor, Tore Supra, W Environment in Steady-State Tokamak) and two stellarators (Large Helical Device and Wendelstein 7-X) expanding the pioneering work of Kikuchi (Kikuchi M. and Azumi M. 2015 Frontiers in Fusion Research II: Introduction to Modern Tokamak Physics (Springer)). Data have been gathered up to January 2022 and coordination has been provided by the recently created International Energy Agency-International Atomic Energy Agency international C oordination on I nternational C hallenges on L ong duration OP eration group. By exploiting the multi-machine international database, recent progress in terms of injected energies (e.g. 1730 MJ in L-mode, 425 MJ in H-mode), durations (1056 s in L-mode, 101 s in H-mode), injected powers, and sustained performance will be reviewed. Progress has been made to sustain long-pulse operation in tokamaks and stellarators with superconducting coils, actively cooled components, and/or with metallic walls. The graph of the fusion triple products as a function of duration shows a dramatic reduction of, at least two orders of magnitude when increasing the plasma duration from less than 1 s to 100 s. Indeed, long-pulse operation is usually reached in dominant electron-heating modes at reduced density (current drive optimization) but with low ion temperatures ranging from 1 to 3 keV for discharges above 100 s. Difficulties in extending the duration may arise from coupling high heating powers over long durations and the evolving plasma-wall interaction towards an unstable operational domain. Possible causes limiting the duration and critical issues to be addressed prior to ITER operation and DEMO design are reported and analyzed. more...

Early stellarator designs suffered from high particle losses, an issue that can be addressed by optimization of the coils. Here the authors measure the magnetic field lines in the Wendelstein 7-X stellarator, confirming that the complicated design of the superconducting coils has been realized successfully. more...

Nature Communications 7: Article number: 13493 (2016); Published: 30 November 2016; Updated: 14 February 2017 The original version of this Article contained errors in the spelling of authors names and their affiliations. These errors were: The affiliation of Boyd Blackwell is 10, but was incorrectlygiven as 10,11,12. more...

The Wendelstein 7-X (W7-X) optimized stellarator fusion experiment, which went into operation in 2015, has been operating since 2017 with an un-cooled modular graphite divertor. This allowed first divertor physics studies to be performed at pulse energies up to 80 MJ, as opposed to 4 MJ in the first operation phase, where five inboard limiters were installed instead of a divertor. This, and a number of other upgrades to the device capabilities, allowed extension into regimes of higher plasma density, heating power, and performance overall, e.g. setting a new stellarator world record triple product. The paper focuses on the first physics studies of how the island divertor works. The plasma heat loads arrive to a very high degree on the divertor plates, with only minor heat loads seen on other components, in particular baffle structures built in to aid neutral compression. The strike line shapes and locations change significantly from one magnetic configuration to another, in very much the same way that codes had predicted they would. Strike-line widths are as large as 10 cm, and the wetted areas also large, up to about 1.5 m2, which bodes well for future operation phases. Peak local heat loads onto the divertor were in general benign and project below the 10 MW m−2 limit of the future water-cooled divertor when operated with 10 MW of heating power, with the exception of low-density attached operation in the high-iota configuration. The most notable result was the complete (in all 10 divertor units) heat-flux detachment obtained at high-density operation in hydrogen. [ABSTRACT FROM AUTHOR] more...

The main objective of the Wendelstein 7-X (W7-X) stellarator is to demonstrate the integrated reactor potential of the optimized stellarator line. An important element of this mission is the achievement of high heating-power and high confinement in steady-state operation. Such an integrated plasma operation has not yet been demonstrated and represents the major scientific goal of W7-X. The way towards this goal is staged. In the first phase, called OP 1.1, December 2015-March 2016, a limiter configuration was used. In this paper, the preparation of the first operation phase as well as lessons learned during the first commissioning and the operation phase are discussed, while the physics results from OP 1.1 are reported elsewhere (Wolf et al 2017 Nucl. Fusion57 102020). [ABSTRACT FROM AUTHOR] more...

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. [ABSTRACT FROM AUTHOR] more...

Observations on confinement in the first experimental campaign on the optimized Stellarator Wendelstein 7-X are summarized. In this phase W7-X was equipped with five inboard limiters only and thus the discharge length restricted to avoid local overheating. Stationary plasmas are limited to low densities  <2–3 · 1019 m−3. With the available 4.3 MW ECR Heating core Te ~ 8 keV, Ti ~ 1–2 keV are achieved routinely resulting in energy confinement time τE between 80 ms to 150 ms. For these conditions the plasmas show characteristics of core electron root confinement with peaked Te-profiles and positive Er up to about half of the minor radius. Profiles and plasma currents respond to on- and off-axis heating and co- and counter ECCD respectively. [ABSTRACT FROM AUTHOR] more...

Wendelstein 7-X (W7-X) is currently under commissioning in preparation for its initial plasma operation phase, operation phase 1.1 (OP1.1). This first phase serves primarily to provide an integral commissioning of all major systems needed for plasma operation, as well as systems, such as diagnostics, that need plasma operation to verify their foreseen functions. In OP1.1, W7-X will have a reduced set of in-vessel components. In particular, five graphite limiter stripes replace the later foreseen divertor. This paper describes the expected machine capabilities in OP1.1, as well as a selection of physics topics that can be addressed in OP1.1, despite the simplified configuration and the reduced machine capabilities. Physics topics include the verification and adjustment of the magnetic topology, the testing of the foreseen plasma start-up scenarios and the feed-forward control of plasma density and temperature evolution, as well as more advanced topics such as scrape-off layer (SOL) studies at short connection lengths and transport studies. Plasma operation in OP1.1 will primarily be performed in helium, with a hydrogen plasma phase at the end. [ABSTRACT FROM AUTHOR] more...