The W7-X stellarator, presently under construction in Greifswald will be equipped from the beginning with an open island divertor designed for stationary operation. In order to optimize the W7-X cryopumping system for an efficient particle control with good confinement in long-pulse operation, simulations are performed with the 3D EMC3-EIRENE plasma edge transport code. The establishment of stationary particle balance crucially depends on the divertor pumping efficiency, which, for a given particle input, sets a lower limit to the plasma density at the separatrix. This paper presents a self-consistent study with EMC3-EIRENE code on the particle balance and pumping capability of W7-X at stationary conditions for relevant ranges of magnetic configurations (ι=5/5,5/4) and plasma parameters. Specifically, the pumping and recycling fluxes are obtained and lower limits for separatrix density are estimated for three major magnetic configurations with the required pumping parameters in various operating conditions.

With Large Helical Device (LHD) and Wendelstein 7-X (W7-X), the development of helical devices is now taking a large step forward on the path to a steady-state fusion reactor. Important issues that need to be settled in these machines are particle flux and heat control and the impact of divertors on plasma performance in future continuously burning fusion plasmas. The divertor concepts that will initially be explored in these large machines were prepared in smaller-scale devices like Heliotron E, Compact Helical System (CHS), and Wendelstein 7-AS (W7-AS). While advanced divertor scenarios relevant for W7-X were already studied in W7-AS, other smaller-scale experiments like Heliotron-J, CHS, and National Compact Stellarator Experiment will be used for the further development of divertor concepts. The two divertor configurations that are being investigated are the helical and the island divertor, as well as the local island divertor, which was successfully demonstrated on CHS and just went into operation on LHD. At present, on its route to a fully closed helical divertor, LHD operates in an open helical divertor configuration. W7-X will be equipped right from the start with an actively cooled discrete island divertor that will allow quasi-continuous operation. The divertor design is very similar to the one explored on W7-AS. For sufficiently large island sizes and not too long field line connection lengths, this divertor gives access to a partially detached quasi-steady-state operating scenario in a newly found high-density H-mode operating regime, which benefits from high energy and low impurity confinement times, with edge radiation levels of up to 90% and sufficient neutral compression in the sub-divertor region (>10) for active pumping. The basic physics of the different divertor concepts and associated implementation problems, like asymmetries due to drifts, accessibility of essential operating scenarios, toroidal asymmetries due to symmetry breaking error fields, etc., are discussed.

To minimize radiation losses of the plasma during long pulse operation, the first wall protection of WENDELSTEIN 7-X demands low-Z plasma facing materials. In addition to carbon materials on high heat flux loaded components, 300–500 μm thick vacuum plasma-sprayed (VPS) layers of boron carbide (B4C) are considered as coatings on large actively cooled stainless steel panels. In order to evaluate the behaviour relevant to the expected plasma wall interactions, an extensive material characterization and test programme was executed. Also included is the examination of the industrial manufacturing and the suitability of these coatings as 70 m2 first wall plasma facing material for W7-X. To improve the adhesion of thick B4C layers on stainless steel, different interlayers have been investigated. Based on the results, a VPS-based coating technique was identified which is suitable to manufacture the B4C protection layers on stainless steel wall panels of W7-X.

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.

The research on divertors for stellarators is at the beginning. Extensive studies are being prepared on large helical device (LHD) and W7-X. W7-AS is now being operated with an open island divertor (ID) which serves as a test bed for the W7-X divertor. The divertor enables access to a new NBI-heated, high-density operating regime with improved confinement properties. This regime-the high-density H-mode (HDH)-displays no evident mode activity, is extant above a threshold density and characterized by flat density profiles, high-energy- and low-impurity-confinement times and edge-localized radiation. Impurity accumulation, normally associated with ELM-free H-modes, is avoided. Quasi-teady-state discharges with (n) over bar (e) up to 4 x 10(20) m(-3), edge radiation levels up to 90% and plasma partial detachment at the divertor targets can be simultaneously realized. The accessibility to other improved confinement modes in W7-AS (conventional H-mode and OC-mode) is not restricted by the divertor. The results provide a promising basis for future experiments, in particular on W7-X, and recommend the ID as a serious candidate for solving the plasma exhaust problem in stellarators. (C) 2003 Elsevier Science B.V. All rights reserved.

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.

The divertor and the pumping system of Wendelstein 7-X is designed for stationary operation at various modes of operation. Maximum fluxes up to 1024 electron–ion-pairs/s are expected at the targets. Various cases of plasma–wall interaction have to be considered: at low densities convective power losses and proceeding to higher densities a high recycling mode, including detachment dominated by radiation. External fluxes (NBI, gas puffing and pellet injection) up to 1022 particles/s have to be pumped. The geometry of the target plates, the installation of additional baffle plates and the position of the pumping gap were optimised on the basis of numerical studies with the 3D neutral particle code EIRENE. For particle exhaust turbomolecular pumps (TMPs) and cryo-panels will be installed. Taking in account the capabilities of the TMP in the magnetic stray field an effective pumping speed for H2 of 4200 l/s at one divertor box will be reached. Including the cryo-panels integrated inside the divertor boxes a total pumping speed of up to 200 000 l/s will be provided.

Analysis of neutral particle transport is of importance for the successful steady-state divertor operation proposed for the stellarator W7-X. In addition to the range of magnetic configurations and the incident energy flux, the divertor design for W7-X is also being optimized for an additional control over the plasma using the particle sources in the divertor region [1]. The present work aims at characterizing the plasma and neutral particle transport processes in the plasma edge region using the 3D Monte Carlo code EMC3-EIRENE [2, 3]. The self-consistent analysis of plasma and neutral transport is done in order to optimize the divertor operation by effectively controlling the exposure of neutral particles in subdivertor volume to the edge plasma. The study involves simulation of the plasma transport to determine the density, temperature and flow velocity profiles

In the TEXTOR tokamak the five top and five bottom poloidal carbon limiter blocks have been replaced by inertially cooled copper blocks coated with a 170 μm VPS-B4C layer. Similar limiter blocks have been inserted through lock systems, extensively diagnosed in situ as well as ex situ. During the thermal load by the plasma, the surface temperature rose and decayed extremely fast which can be explained by a different thermal conductivity and heat capacity of the coating. For heat loads below 8 MWm−2 no severe cracking or delamination of the B4C-coating were observed. Due to the insulating behaviour of the layer, distinct craters developed on both limiter types, which reached down to the copper surface and are assumed to be caused by electrical arcs. An oscillation of the evolution of the surface temperature has been observed under certain conditions, which is clearly correlated to the use of the coated test limiter. Particle fluxes as well as hydrogen inventory turned out to be very similar to those from a low-Z surface in a carbon surrounding. No significant impact of the plasma on the coating and vice versa was observed.

The divertor and the pumping system of W7-X are designed for stationary operation at the full range of magnetic and plasma parameters (5/6 < designed to withstand a stationary heat flux up to 10-12 MW/m2, are shielded with CFC joined to a CuCrZr heat sink cooled by a pressurised water-flow. The final geometry was optimised to reduce the total target area without narrowing the operational range of parameters. Baffles guide the neutrals to the pumping gap into the divertor chamber and keep the neutral pressure in the main chamber low. The baffle plates are designed to withstand a maximum heat flux of 0.5 MW/m2 and consist of graphite tiles clamped on a water-cooled CuCrZr structure. External particle fluxes injected by NBI, localised gas puffing and pellet injection in a range up to 10 s will be applied for heating and controlling the density profiles. These fluxes have to be pumped out to achieve a stationary regime. The main pumping system of W7-X will consist of turbomolecular pumps (TMPs), connected to the 10 divertor chambers, with their corresponding backing pumps. Cryo-panels will be integrated inside the divertor units. The high pumping speed can be maintained up to two hours. Introduction The HELIcal Advanced Stellarator (HELIAS) WENDELSTEIN 7-X (W7-X) was designed at IPP Garching and is presently build in the institute's branch at Greifswald, Germany. The experiment aims to demonstrate the reactor potential of this stellarator line at steady-state operation close to fusion relevant parameters. This requires the use of superconducting coils and the installation of a divertor to handle the power and the particle fluxes. The magnetic configuration of the device has five field periods and was optimised with respect to plasma equilibrium, stability and reduced neo-classical transport in a wide range of parameters [1 – 3]. The chosen magnetic configuration is characterised by inherent divertor properties without the installation of additional coils [4]. The interaction of the divertor target plates with the islands at the boundary or the ergodisation produces a region with open field lines and allows to localise the divertor target plates sufficiently far from the confinement region to screen neutrals and impurities [5]. The target plates have to follow a complex 3-D geometry as the islands are winding helical around the confinement region. In this paper we discuss the engineering of the divertor structures responsible for the power handling, and the pumping systems for an effective particle exhaust. Divertor engineering and operation For W7-X an open divertor configuration was chosen as a first approach to match the whole operational range of the rotational transform and beta values. Two divertor chambers are arranged in a mirror-symmetry above and below the helical axis in each of the five field periods. In total, ten divertor units are located along the helical edge of the plasma column. These units consist of the target plates for the dissipation of energy, the baffle plates to guide the neutral particles to the pumping slit and are closed by poloidal and toroidal end plates [6]. The actively cooled wall protection shields the wall of the plasma vessel and minimises the heat transfer to the plasma vessel. These plasma facing components are coated with low-Z material (vacuum plasma sprayed boron carbide on stainless steel) or consist of graphite on a CuCrZr heat sink. They are able to withstand long pulse discharges [7,8]. The main heating system is ECRH with a max. power of 10 MW for max. 30 min duration. The experimental flexibility of the device is increased with additional heating systems such as ICRH with a max. power of 4 MW for max. 30 min duration, and NBI with a max. power of 5 MW for several 10 s intervals per 30 min discharge. Fig. 1 Overall view of a divertor chamber including the target and baffle modules in the plasma vessel The divertor concept and design was introduced in detail in [9]. The 3D-shaped surface of the target plates is approximated by a series of plane target elements (Figure 1). The decision to increase the angle of incidence of incoming particles onto the target plates by 1° allowed to reduce the area of the highest loaded targets from about 30 m to about 20 m. The new design of the target plates combines two different technological solutions. The highest loaded areas are designed to withstand stationary heat fluxes up to 10-12 MW/m. The selected design is a flat tile made of CFC Sepcarb NB31 as plasma facing material bonded by AMC and electron beam welding onto a heat sink made of CuCrZr-alloy watercooled by an inner channel equipped with a twisted tape. There are five different basic types of target elements, of a length ranging from 250 to 595 mm and of a width from 50 to 57 mm. For some types, the inboard part facing the pumping slits in the opening of the divertor is also shielded with CFC tiles. Special additional types of target elements dedicated to diagnostics have been designed. Slits of about 10 mm width in the toroidal direction allow experimental studies through the target plates to be carried out. The remaining area is designed to withstand a stationary heat flux up to about 1 MW/m (Figure 2). The technological solution consists of graphite tiles clamped on CuCrZr heat sinks which are brazed onto water-cooled stainless steel tubes [10]. The baffle plates are designed to withstand stationary heat fluxes up to 0.5 MW/m. These plates also protect the water-cooling system of the target elements. The total area of the baffle plates is about 30 m. The basic technology is the same as for the lowest loaded areas of the target plates graphite tiles clamped on CuCrZr heat sink brazed to water-cooled stainless steel tubes. Fig. 2 Combination of two different technological solutions for the target plates matched to the increased angle of incidence of the incoming particles by 1°. The highest loaded areas withstand up to about 10-12 MW/m (gray). The remaining area for up to about 1 MW/m is marked in orange. The wall protection area is about 120 m. This plasma facing component is designed to withstand stationary heat fluxes ranging from 200 to 500 kW/m depending on the location. Two different technologies have been selected. For higher loaded areas, it is the same technology as for the baffle plates. For lower loaded areas, cylindrical-shaped water-cooled panels made of stainless steel and covered with boron carbide are used. The thickness of the boron carbide layer is 300 m. This modification of the target areas also required the geometrical adaptation of other in-vessel components to this new configuration, like the baffle plates, the toroidal closing plates, the wall protections, the diagnostics, the control coils, and the cryopumps. With this new configuration, it will still be possible to operate in the full plasma and magnetic parameter space, but further detailed studies are necessary to clear up the action of the control coils and the pumping. First results of the EMC3-EIRENE code [11, 12] with the sampling grid upgraded to incorporate arbitrary magnetic structures confirm the assumptions made earlier concerning the width of the pumping gap [13, 14]. The behaviour of selfconsistently generated recycling neutrals was studied in the plasma of the scrape-off layer of W7-X chamber was calculated for a range of divertor openings in order to validate the proposed divertor geometry. Achievable pumping speeds were estimated for a range of pumping efficiencies of the proposed pumping system. An optimum pumping speed of 3.5*10 s was estimated for a neutral particle density of 1.0*10 cm in the divertor chamber (Figure 3). Fig. 3 Dependence of Hydrogen pumping speed on the width of the pumping gap (normalised to the actual chosen value A0). here varied from 2 to 10 %. The neutral density was taken to be 1.0*10 cm . Particle exhaust The vacuum pumping installation will consist of two independent systems turbomolecular pumps (TMPs) and cryo-panels. Each of the 10 divertor chambers will be equipped with two ports (diameter 400 mm) for pumping. TMPs with their corresponding backing pumps will define the basic system for pumping the plasma vessel of W7-X. They will be used for pump down from atmospheric pressure to the base level of 10 mbar, during conditioning of the vessel as well as during the plasma experiments. Whereas the ten backing pump blocks (two-stage systems consisting of 1,000 m3/h roots-pumps and 65 m3/h rotary pumps to handle high gas loads stationary will be located far enough from the experiment, the TMPs have to work in the magnetic stray field of the superconducting coil system of W7-X. Consequently the positioning of the TMPs will be a compromise with respect to the effective pumping speed in the divertor chamber and the value of the magnetic stray field. Testing the TMPs of various manufacturers in a variable magnetic field was therefore essential to design the pump lines and fix the positions of the TMPs relative to the cryostat. Including a safety margin, a magnetic field of 7 mT perpendicular to the rotor axis and up to 15 mT parallel to the rotor axis seem to be the upper limit for the use of TMPs in a stationary magnetic field, higher pulse loads seem to be acceptable. However, TMPs with magnetic suspension were sensitive to fields above 15 mT, causing the suspension to fail. For W7-X it was decided to install three TMPs to each divertor chamber with a total nominal pumping speed of 6,000 l/s, connected via approx. 3.6 m and 4.3 m long tubes of 400 mm diameter, thus reaching about 4,200 l/s pumping speed for hydrogen at the plasma vessel [14]. The cryo-panels will be located directly behind the target plates in the divertor chamber with the openings pointing towards the pumping gap [15]. The restricted space and the needs for diagnostics make it necessary to divide the panels into two sections (Figure 4). Fig. 4 Arrangement of the cryo panels behind the tar

Two significiant problems that need to be solved for any future fusion device are heat removal and particle control. A very promising method to attack these problems in tokamaks and helical devices is, the use of a divertor, providing a controlled interaction zone between plasma and wall. By carefully designing a divertor, conditions can be created in front of the divertor targets, which lead to a. sufficient reduction of the power load on the targets by strong radiation redistribution. Any solution of course I needs to allow for an energy confinement which is at least sufficient for the realization of a fusion reactor. Since energy confinement has been found to be strongly related to edge anomalous transport and edge plasma profiles,the ultimate aim is to find an integral solution which is optimum with respect to exhaust, heat load and energy confinement.Two different types of divertors are presently being investigated in helical devices: the 'helical divertor' and the 'island divertor'. So far divertor' concepts have been investigated only in a few helical devices. Theoretical and experimental efforts have mainly concentrated on the suitability of divertor magnetic field structures, While detailed studies of the divertor plasma properties for the two types of divertor configurations have only recently begun. In the course of this exploration, a promising new high-density H-mode (HDH) plasma operational regime has been discovered on the Wendelstein stellarator W7-AS. It benefits from high-energy (up to twice the value of the International Stellarator Scaling ISS95) and. low impurity confinement times, complemented by edge radiated power fractions of up to 90% in detached regimes. This allowed quasi-steady-state operation for up to 50 energy confinement times and so far was only constrained by machine operability.

The stellarator WENDELSTEIN 7-X is under construction in Greifswald. For energy and particle exhaust under stationary conditions ten divertor units are arranged along the helical edge of the five-fold symmetric plasma column. In respect of the boundary variations in a first step an “open” divertor was chosen. The geometry and the specifications of the in-vessel components reflecting the 3D topology of the boundary are defined in accordance with results of various theoretical studies. The theoretical methods for the characterisation and the proposed technical solutions for the main components including the instrumentation target, baffle, wall protection, and control coils will be described.

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.

For operation within a wide range of plasma and magnetic parameters an open divertor structure of the Wendelstein 7-X stellarator (optimised HELIAS configuration with superconducting coils, main parameters: R = 5.5 m, a = 0.55 m, B = 3 T, rotational transform variable between 5/6 and 5/4) has been developed. The most critical components of the divertor are ten 3D-shaped target surfaces, located along the “helical edge” of the five-fold rotational symmetry magnetic configuration of the machine. The plasma outflow concentrates on the target areas, designed to withstand a maximum stationary heat flux of 10 MW/m2. To prevent excessive radiation losses all plasma facing components will be manufactured using low-Z materials, e.g. CFC, B4C. The paper presents the optimised design of the target plates, which takes into account the latest results of the physics, the requirements of the heating and diagnostic systems, the manufacturing costs and the mounting operations in order to build a technical reliable system.

For the actively cooled first wall panels of the W7-X stellarator under construction at Greifswald, Germany, boron carbide (B4C) coatings with a thickness up to 500 µm are being developed as plasma facing surface on the stainless steel component. Different coating technologies (vacuum plasma spray (VPS), atmospherical plasma spray (APS)) are investigated. Characterisation of the coatings comprises metallographic and SEM investigations, impurity determination and measurement of the thermal conductivity, characterisation of the sputtering behaviour and the behaviour under plasma exposure. First in-pile-tests in the tokamak TEXTOR-94 at the Research Centre Juelich have been performed. 10 poloidal limiter elements and 5 test limiters made of copper with a 170 µm thick B4C VPS coating were exposed to plasma discharges with ohmic and additional heating. By varying the limiter positions and plasma density, different loading conditions were imposed. After exposure, the test limiters were removed for further analyses. First results showed the formation of small craters in the coating, probably due to arcing. Although melting occurred within a limited zone around the craters, the coatings did not detach from the substrate or show severe cracking outside the molten area.