https://iopscience.iop.org/article/10.1088/1361-6587/aaaf62/pdf

The influence of plasma-surface interaction on the performance of tungsten at the ITER divertor vertical targets

Behavior of tungsten under irradiation and plasma interaction

Challenges for plasma-facing components in nuclear fusion

ASDEX Upgrade has recently finished its transition towards an all-W divertor tokamak, by the exchange of the last remaining graphite tiles to W-coated ones. The plasma start-up was performed without prior boronization. It was found that the large He content in the plasma, resulting from DC glow discharges for conditioning, leads to a confinement reduction. After the change to D glow for inter-shot conditioning, the He content quickly dropped and, in parallel, the usual H-Mode confinement with H factors close to one was achieved. After the initial conditioning phase, oxygen concentrations similar to that in previous campaigns with boronizations could be achieved. Despite the removal of all macroscopic carbon sources, no strong change in C influxes and C content could be observed so far. The W concentrations are similar to the ones measured previously in discharges with old boronization and only partial coverage of the surfaces with W. Concomitantly it is found that although the W erosion flux in the divertor is larger than the W sources in the main chamber in most of the scenarios, it plays only a minor role for the W content in the main plasma. For large antenna distances and strong gas puffing, ICRH power coupling could be optimized to reduce the W influxes. This allowed a similar increase of stored energy as yielded with comparable beam power. However, a strong increase of radiated power and a loss of H-Mode was observed for conditions with high temperature edge plasma close to the antennas. The use of ECRH allowed keeping the central peaking of the W concentration low and even phases of improved H-modes have already been achieved.

Plasma wall interaction and its implication in an all tungsten divertor tokamak. Invited Paper

Overview on plasma operation with a full tungsten wall in ASDEX Upgrade

ASDEX Upgrade became a full tungsten experiment in 2007 by coating its graphite plasma facing components with tungsten. In 2013 a redesigned solid tungsten divertor, Div-III, was installed and came into operation in 2014. The redesign of the outer divertor geometry provided the opportunity to increase the pumping efficiency in the lower divertor by increasing the gap between divertor and vessel. In parallel, a by-pass was installed into the cryo-pump in the divertor region allowing adapting of the pumping speed to the required edge density.Safe divertor operation and heat removal becomes more and more significant for future fusion devices. This requires developing 'tools' for divertor heat load control and to optimize the divertor design. The new divertor manipulator, DIM-II, allows retracting a relevant part of the outer divertor into a target exchange box without venting ASDEX Upgrade. Different front-ends can be installed and exposed to the plasma. At present, front-ends for probe exposition, gas puffing, electrical probes and actively cooled prototype targets are under construction.The installation of solid tungsten, the control of the pumping speed and the flexibility for testing divertor modifications on a weekly base is a unique feature of ASDEX Upgrade and offers together with the extended set of diagnostics the possibility to investigate dedicated questions for a future divertor design.

/pdf/solid-tungsten-divertor-iii-for-asdex-upgrade-and-jv76yawsls.pdf

Solid tungsten Divertor-III for ASDEX Upgrade and contributions to ITER

A large divertor manipulator for ASDEX Upgrade

The stellarator machine WENDELSTEIN 7-X is in the construction phase at IPP Greifswald, Germany. The elements of the mechanical structure supporting the superconducting coil system are critical components subject to very high forces and moments. The design of such components has been optimised with the objective of minimising the deformations of the coil winding packages and to limit the stresses within acceptable values. An extensive R&D program is in progress to qualify the adopted solution. The paper reports main design features of the support elements of the mechanical structure and the preliminary results of the tests campaign.

The WENDELSTEIN 7-X mechanical structure support elements: Design and tests

The Max Planck Institute of Plasmaphysics is building up the stellarator fusion experiment Wendelstein 7-X (W7-X) at the branch institute in Greifswald. W7-X continues the line of stellarator experiments at IPP. To allow for steady state operation W7-X has a superconducting coil system with 50 non-planar and 20 planar coils. The coil system is grouped in five equal modules, each consisting of two mirror symmetric half modules. The half modules are assembled from five different non-planar coils, two planar coils and a sector of the coil support structure. All cryogenic parts are enclosed in a cryostat to protect them from ambient temperature. The magnet system was ordered from the European industry. The production of superconductor, winding packs and encasings are under way. The main focus of this contribution aims on the fabrication state of the coil system.

Fabrication of the superconducting coils for Wendelstein 7-X

https://pure.mpg.de/pubman/item/item_2145342_1/component/file_2145341/Herrmann_Design.pdf

N. Jaksic

Papers

Solid tungsten Divertor-III for ASDEX Upgrade and contributions to ITER

A large divertor manipulator for ASDEX Upgrade

The WENDELSTEIN 7-X mechanical structure support elements: Design and tests

Fabrication of the superconducting coils for Wendelstein 7-X

Design and concept validation of the new solid tungsten divertor for ASDEX Upgrade