H. Greuner

Current status of the JET ITER-like Wall Project

This paper presents an overview of the status and relevant technical issues for the ITER-like Wall Project with emphasis on progress since the 11th International Workshop on Plasma-Facing Materials and Components for Fusion Applications.

Investigation of Tungsten Coatings on Graphite and CFC

In the frame of JETs ITER-like wall (ILW) project tungsten coatings on carbon fibre reinforced carbon substrates will be used in the divertor and highly loaded areas in the main chamber. Fourteen different types of samples were produced by physical or chemical vapour deposition and vacuum plasma spray (VPS) with coating thickness of 4, 10 and 200 μm. Similarly, three different VPS W coatings (200 μm) on two different graphite substrates, were produced for use at the strike-point regions of ASDEX Upgrade. All coatings were subjected to thermal screening and thermal cycling tests in the ion beam facility GLADIS. Additionally, the coatings intended for the ILW project were exposed to edge localized mode (ELM)-like thermal loads in the electron beam facility JUDITH. A general failure mode with the CFC substrate is crack formation upon cool-down, whereas the coatings on graphite do not show any crack formation. Additionally, metallographic investigations, x-ray diffraction measurements, adhesion testing as well as measurements on the contents of light impurities were performed.

Final design of W7-X divertor plasma facing components – tests and thermo-mechanical analysis of baffle prototypes

Abstract The plasma facing components (PFCs) of the W7-X are designed in detail. The current design of the target plates, baffle plates and wall protection is presented which takes into account the requirements of the plasma heating, diagnostic systems and mounting. Prototypes of baffle elements were tested with heat loading to investigate the long term behaviour. The experimental results are compared with finite element calculations of the temperature and stress distributions in the elements. Based on these activities, the fabrication of the W7-X divertor PFCs and the graphite covered wall protection for W7-X can be initiated.

Tungsten and beryllium armour development for the JET ITER-like wall project

For the ITER-like wall project at JET the present main chamber CFC tiles will be exchanged with Be tiles and in parallel a fully tungsten-clad divertor will be prepared. Therefore three R&D programmes were initiated: Be coatings on Inconel as well as Be erosion markers were developed for the first wall of the main chamber. High heat flux screening and cyclic loading tests carried out on the Be coatings on Inconel showed excellent performance, above the required power and energy density. For the divertor a conceptual design for a bulk W horizontal target plate was investigated, with the emphasis on minimizing electromagnetic forces. The design consisted of stacks of W lamellae of 6 mm width that were insulated in the toroidal direction. High heat flux tests of a test module were performed with an electron beam at an absorbed power density up to 9 MW m −2 for more than 150 pulses and finally with increasing power loads leading to surface temperatures in excess of 3000 ◦ C. No macroscopic failure occurred during the test while SEM showed the development of micro-cracks on the loaded surface. For all other divertor parts R&D was performed to provide the technology to coat the 2-directional CFC material used at JET with thin tungsten coatings. The W-coated CFC tiles were subjected to heat loads with power densities ranging up to 23.5 MW m −2 and exposed to cyclic heat loading for 200 pulses at 10.5 MW m −2 . All coatings developed cracks perpendicular to the CFC fibres due to the stronger contraction of the coating upon cool-down after the heat pulses.

Qualification of tungsten coatings on plasma-facing components for JET

This contribution summarizes the work that has been performed to establish the industrial production of tungsten coatings on carbon fibre composite (CFC) for application within the ITER-like Wall Project at JET. This comprises the investigation of vacuum plasma-sprayed coatings, physical vapour deposited tungsten/rhenium multilayers, as well as coatings deposited by combined magnetron-sputtering and ion implantation. A variety of analysis tools were applied to investigate failures and oxide and carbide formation in these systems.

Experiences with tungsten coatings in high heat flux tests and under plasma load in ASDEX Upgrade

ASDEX Upgrade was operated with about 6400 s plasma discharge during the scientific program in 2007/2008 exploring tungsten as a first wall material in tokamaks. In the first phase, the heating power was restricted to 10 MW. It was increased to 15 MW in the second phase. During this operational period, a delamination of the 200 μm W-VPS coating happened at 2 out of 128 tiles of the outer divertor and an unscheduled opening was required. In the third phase, ASDEX Upgrade was operated with partly predamaged tiles and up to 15 MW heating power. The target load was actively controlled by N2-seeding. This paper presents the screening test of target tiles in the high heat flux test facility GLADIS, experiences with operation and detected damages of the outer divertor as well as the heat load to the outer divertor and the reasons for the toroidal asymmetry of the divertor load.

Materials for the plasma-facing components of fusion reactors

Abstract According to current knowledge and understanding, nuclear fusion can be developed to a sustainable energy technology. Fuel is abundant and key points for fusion power production and alpha particle heating have already been demonstrated. The next step device, international thermonuclear experimental reactor (ITER), is designed to demonstrate net fusion power production and to address most of the technological issues on the way to a power reactor. There is, however, a series of materials problems related to the plasma facing components and to the structural materials which cannot be fully addressed by ITER. These developments are covered by long term development of radiation resistant low activation materials, heat sink materials, plasma facing protection materials as well as functional tritium barrier materials. A brief survey of the current status of materials development for plasma facing applications is given in the present paper. To provide materials which can sustain the severe loading conditions in a fusion environment is a key issue in developing fusion as an economic energy technology without long lived radioactive waste.

Water-cooled target modules for steady-state operation of the W7-X divertor

The stellarator WENDELSTEIN 7-X (W7-X) includes water-cooled plasma facing components (PFCs) to allow steady-state operation and to provide an efficient particle and power exhaust up to 10 MW for a maximum pulse duration of 30 min. Ten divertor units are arranged along the helical edge of the fivefold periodic plasma column. The three-dimensional shape and positioning of the target surfaces are optimized to address physics issues for a wide range of experimental parameters, which influence the topology of the boundary. The three-dimensional target surfaces are reproduced by a series of consecutive plane target elements as a set of parallel water-cooled elements positioned onto the frameworks of target modules. The design and arrangement of target modules and elements are described.

Divertor concept for the W7-X stellarator and mode of operation

A favourable property of the stellarator concept is the potential of stationary operation within a magnetic configuration maintained by a superconducting coil system. For proof of principle the stellarator Wendelstein 7-X is presently under construction at Greifswald, Germany, and the start of operation is planned for 2007. The magnetic configuration of the confinement is a non-axisymetric three-dimensional configuration with a helix-like magnetic axis and five identical magnetic field periods. As a first-step divertor design, an open divertor structure has been chosen, which benefits from the inherent divertor property of the magnetic configuration. The system will allow an effective particle and energy exhaust for a wide range of plasma and magnetic parameters. Experimental tools, e.g. localized heating, various heating schemas, gas feed and pellet injection, impurity doping and variation of the pumping speed together with appropriate diagnostics are provided. The purpose is to investigate different modes of operation for the divertor system and to evaluate an extended database for further improvement of the divertor. The main heating method will be 140 GHz ECR as a cw heat source of 10 MW. Additional heating schemes are ICRF and NBI.

A solid tungsten divertor for ASDEX Upgrade

The conceptual design of a solid tungsten divertor for ASDEX Upgrade (AUG) is presented. The Div-III design is compatible with the existing divertor structure. It re-establishes the energy and heat receiving capability of a graphite divertor and overcomes the limitations of tungsten coatings. In addition, a solid tungsten divertor allows us to investigate erosion and bulk deuterium retention as well as test castellation and target tilting. The design criteria as well as calculations of forces due to halo and eddy currents are presented. The thermal properties of the proposed sandwich structure are calculated with finite element method models. After extensive testing of a target tile in the high heat flux test facility GLADIS, two solid tungsten tiles were installed in AUG for in-situ testing.