More and more attention is directed towards thermonuclear fusion as a possible future energy source. Major advantages of this energy conversion technology are the almost inexhaustible resources and the option to produce energy without COremissions. However, in the most advanced field of magnetic plasma confinement a number of technological challenges have to be met. In particular high-temperature resistant and plasma compatible materials have to be developed and qualified which are able to withstand the extreme environments in a commercial thermonuclear power reactor. The plasma facing materials (PFMs) and components (PFCs) in such fusion devices, i.e. the first wall (FW), the limiters and the divertor, are strongly affected by the plasma wall interaction processes and the applied intense thermal loads during plasma operation [1]. On the one hand, these mechanisms have a strong influence on the plasma performance; on the other hand, they have major impact on the lifetime of the plasma facing armour.

Fine grained oxide dispersion strengthened (ODS) W based materials with grain size from the sub-micron range to several microns have been fabricated by mechanical alloying and spark plasma sintering. The emphasis is put on the effect of alloying elements on the microstructure, basic mechanical properties and transient high heat loading performance. The micro-hardness of the fabricated ODS-W based materials are between 400HV and 1050HV, while the bending strength are between 700MPa and 1300MPa. The addition of a small amount of Ti will promote the densification and decrease the grain size of W–Y2O3 composite significantly, while the addition of Mo shows less function. However, the Mo alloyed ODS-W sample shows a much better transient high heat loading performance than the Ti alloyed ODS-W sample.

Metallic mirrors will be essential components of all optical spectroscopy and imaging systems for plasma diagnosis that will be used on the next-step magnetic fusion experiment, ITER. Any change of the mirror performance, in particular reflectivity, will influence the quality and reliability of detected signals. On the request of the ITER Design Team, a First Mirror Test (FMT) has been carried out at JET during campaigns in 2005-2007 and 2008-2009. To date, it has been the most comprehensive test performed with a large number of test mirrors exposed in an environment containing both carbon and beryllium; the total plasma time (in 20052007 period) over 35 h including 27 h of X-point operation. 32 stainless steel and polycrystalline molybdenum flat-front and 45 angled mirrors were installed in separate channels of cassettes on the outer wall and in the MkII HD divertor: inner leg, outer leg and base plate under the load bearing tile. Post exposure studies comprised reflectivity measurements and surface analyses with microscopy, secondary ion mass spectrometry, ion beam analysis and energy dispersive X-ray spectroscopy.. The essential results are: (i) on the outer wall high reflectivity (~90%) is maintained for mirrors close to the channel entrance but it is degraded by 30-40 % deeper in the channel (ii) reflectivity loss by 70-90% is measured for mirrors placed in the divertor: outer, inner and base; (iii) deuterium and carbon are the main elements detected on all mirror surfaces and the presence of beryllium is also found; (iv) thick deposits show rough columnar structure and thickness is 1-20 μm; (v) bubblelike structures are detected in deposits; (vi) the deposition in channels in the divertor cassettes is pronounced at the very entrance; (vii) photonic cleaning with laser removes deposits but the surface is damaged by laser pulses. In summary, reflectivity of all tested mirrors is degraded either by erosion with CX neutrals or by the formation of thick deposits. The implications of results obtained for first mirrors in next-step device are discussed and critical assessment of various methods for in-situ cleaning of mirrors is presented. The conclusion is that engineering solutions should be developed in order to install shutters or to implement a cassette with mirrors to replace periodically the degraded ones

Ultra-fine grained tungsten specimens with a grain size from the submicron range to several microns were fabricated by resistance sintering under ultra-high pressure. Transient heat loads were applied on the tungsten specimens at room temperature for a pulsed duration of 5ms at different power density of 0.22, 0.33, 0.44 and 0.55GWm−2, respectively, by using an electron beam facility. The crack formations and surface melting behaviors under transient heat load were investigated.

The plasma facing materials and components in existing and future fusion devices are strongly affected by plasma wall interaction processes. These components, in particular the first wall (FW), the limiters and the divertor are subject to intense quasi-stationary thermal loads during plasma operation. While the resulting thermal loads to the first wall will remain below 1 MW·m−2, special attention has to be paid to high heat flux components like limiters and the divertor. Here the expected power densities will be at least one order of magnitude above the ones at the FW, with expected peak heat fluxes of up to 20 MW·m−2 for future magnetic confinement devices. Beside quasi-stationary heat loads, short transient thermal pulses with deposited energy densities up to several tens of MJ·m−2 are another serious concern for next step tokamak devices, in particular for ITER. The most serious events are plasma disruptions, vertical displacement events, and Edge Localized Modes (so-called ELMs). These requirements make high demands on the selection of qualified materials and reliable fabrication processes for actively cooled plasma facing components. High heat flux test facilities based on intense electron and ion beams have been utilized successfully to assess the efficiency and the fatigue life time of different material solutions and design concepts. Modeling and experiments with both normal operation scenarios and transient events, are being performed to evaluate and to quantify the resulting material erosion or damage and thus to assess the life time of the components. Additional research activities are focused on the degradation of materials and joints due to energetic neutrons. In order to investigate irradiation induced property changes, materials samples and actively cooled plasma facing components have been irradiated in fission reactors and tested in thermal load tests. The technical solutions which are considered today are mainly based on beryllium, carbon materials or tungsten as armor materials and copper alloys or stainless steel for the heat sink. Furthermore, the needs for extensive quality control methods and non-destructive analyses during the procurement phase will be highlighted.

This contribution provides an overview of experim ental results obtained in the area of material eros ion and migration during the JET operational campaigns 2002 - 2004 (C5 - C14) with the MkIIGB-SRP divertor configuration and summarises recent modelling activities aimed at understanding these data. Post-morte m analysis of a poloidal tungsten stripe deposited pr ior the campaigns in the divertor reveals in agreem ent with previous results that the inner divertor is deposit ion dominated with maximum layer thicknesses up to 300 m whereas the outer divertor is erosion dominated in the average over the campaigns. Detailed SEM analysis shows that the outer divertor erosion occurs non-uniforml y on a scale length of 10 - 30 µm. Shot-resolved deposition/erosion measurements with a quartz microbalance (QMB) mounted in the inner divertor louver region demonstrate that the inner strike point position is the most important parameter for the deposition on the QMB with increasing deposition moving the strike point down the vertical target and largest deposition whe n the strike point is located on the base plate. Data with the s trike point on the base plate show a significantly larger deposition in H-mode than in L-mode discharges (about a factor of 20) while this difference is not cle arly pronounced with the strike point on the vertical ta rget. 13 CH4 injection through the outer divertor during identi cal H-mode discharges showed about 75% of the measured 13 C deposition on the outer divertor tiles but also a significant transport of 13 C to the inner divertor and the main chamber region .

The development of beryllium first wall (FW) plasma facing components for future magnetic confinement fusion experiments, such as the International Thermonuclear Experimental Reactor (ITER), is a topic of great importance as research into long-term energy sources increases in urgency. The FW components must be able to survive the harsh plasma environment for extended periods of time. One proposed method for initial fabrication and repair of FW components is plasma spraying. Previous plasma-sprayed beryllium mock-up FW components had coating separation from the substrate at the edges. The present work describes experiments to produce beryllium mock-up FW components by plasma spray deposition on macro-roughened substrates. Experimental parameters, high heat flux testing and characterization results from the components are presented. No separation of the coating from the substrate was observed. Results of high heat flux testing under electron beam irradiation show performance exceeding that required for ITER FW components. Differences in macro-roughening features result in changes in the threshold absorbed heat flux before damage to the coatings occurs.

Low temperature irradiation embrittlement is one of the major criteria to determine the lifetime of spallation targets. Embrittlement is especially high at low service temperatures, e.g. 250°C in liquid-mercury sources. It was the aim of the present study to investigate the effect of post-irradiation annealing on the mechanical properties of irradiated structural materials. The specimens used were obtained from spent target components of operating spallation facilities (Los Alamos Neutron Science Center, LANSCE, and the Spallation Neutron Source at Rutherford-Appleton Laboratory, ISIS). The investigated materials include a nickel-based alloy (IN718), an austenitic stainless steel (AISI 304L), a martensitic stainless steel (DIN 1.4926) and a refractory metal (Ta) which experienced 800MeV proton irradiation to fluences of several 1025p/m2. The specimens were annealed from 300°C to 700°C for 1 to 10h, respectively, and their mechanical property changes were subsequently investigated at room temperature and 250°C by tensile testing and fracture surface analysis conducted by scanning electron microscopy (SEM). The results showed that the ductility recovered to a large degree in 304L and DIN 1.4926 materials while their strength remained almost unchanged. Especially for DIN 1.4926, the ductility recovery is remarkable already at 400°C. Together with its favorable thermo-mechanical properties, this makes martensitic steel a candidate for structural materials of spallation targets.

The ITER divertor vertical target has to sustain heat fluxes up to 20MWm(-2). ne concept developed for this plasma facing component working at steady state is based on carbon fibre composite armour for the lower straight part and tungsten for the curved upper part. The main challenges involved in the use of such components include the removal of the high heat fluxes deposited and mechanically and thermally joining the armour to the metallic heat sink, despite the mismatch in the thermal expansions. Two solutions based on the use of a CuCrZr hardened copper alloy and an active metal casting (AMC (R)) process were investigated during the ITER EDA phase: the first one called 'flat tile geometry' was mainly developed for the Tore Supra pumped limiter, the second one called 'monoblock geometry' was developed by the EU Participating Team for the ITER project. This paper presents a review of these two solutions and analyses their assets and drawbacks: pressure drop, critical heat flux, surface temperature and expected behaviour during operation, risks during the manufacture, control of the armour defects during the manufacture and at the reception, and the possibility of repairing defective tiles.

The development of plasma facing components for next step fusion devices in Europe is strongly focused to ITER. Here a wide spectrum of different design options for the divertor target and the first wall have been investigated with tungsten, CFC, and beryllium armor. Electron beam simulation experiments have been used to determine the performance of high heat flux components under ITER specific thermal loads. Beside thermal fatigue loads with power density levels up to 20MWm(-2), off normal events are a serious concern for the lifetime of plasma facing components. These phenomena are expected to occur on a time scale of a few milliseconds (plasma disruptions) or several hundred milliseconds (vertical displacement events) and have been identified as a major source for the production of neutron activated metallic or tritium enriched carbon dust which is of serious importance from a safety point of view. The irradiation induced material degradation is another critical concern for future D-T-burning fusion devices. In ITER the integrated neutron fluence to the first wall and the divertor armour will remain in the order of I dpa and 0.7 dpa, respectively. This value is low compared to future commercial fusion reactors; nevertheless, a non-negligible degradation of the materials has been detected, both for mechanical and thermal properties, in particular for the thermal conductivity of carbon based materials. Beside the degradation of individual material properties, the high heat flux performance of actively cooled plasma facing components has been investigated under ITER specific thermal and neutron loads.

Several electron beam and ion beam facilities are involved in high heat flux testing of plasma-facing components for next step fusion devices. Up to a certain degree, these machines are comparable, but differences concern, e.g. beam generation, beam sweeping, calibration techniques and diagnostics. In order to get an information if tests in the different facilities are really comparable, a set of actively cooled CFC monoblocks has been heated in four electron beam and one ion beam facility at comparable power densities. The temperature response during these loadings has been registered and used as a criteria for assessment.

Tensile specimens, prepared from AISI 316L austenitic stainless steel in three conditions (solution-annealed, cold-worked and electron-beam welded) and from OPTIFER martensitic stainless steel in tempered condition, were irradiated in the Swiss spallation neutron source (SINQ) at 90–400°C to displacement doses from 3dpa to 11dpa. The mechanical properties were measured by tensile testing at room temperature and 250°C, respectively, and subsequent metallographic analysis was employed. The tensile results indicated that the strength of AISI 316L-SA is quite similar or a little higher than in 316L-EBW but elongation of SA 316L is somewhat larger than EBW for both unirradiated and irradiated samples. The cold-worked specimens revealed much higher strength but almost zero strain-to-necking after irradiation. The results from OPTIFER samples showed that irradiation hardening increases with dose, which is accompanied by a dramatic reduction of uniform elongation beginning at very low dose. The metallographic analysis showed that the samples of AISI 316L-EBW failed in the welded zone.

To evaluate the performance of plasma facing materials (PFM) and components for future thermonuclear fusion devices under the expected operation conditions, an in-depth material characterization and extensive high heat flux simulation tests are performed routinely in electron beam test facilities. These experiments cover both, thermal fatigue tests with power densities up to approx. 20 MWm-2 and thermal shock loads with deposited energy densities of several MJm-2. In addition, irradiation experiments have been performed in material test reactors to investigate the neutron induced material degradation.

With the installation of a new electron beam (EB) test facility JUDITH II, the parameter range of high heat flux simulations is extended and the urgent need of additional testing capacity is addressed. A nominal power of 200 kW combined with a beam scanning angle of ±14° enables to test larger components of up to 0.5 m × 1 m surface area. A relatively low acceleration voltage (30–60 kV) reduces volumetric heating for the benefit of a more plasma like surface heating. Due to a very flexible and individual programmable system for electron beam pattern generation, very homogeneous load distributions can be achieved during static load tests and highly realistic simulations of ITER relevant transient heat loads become possible. Furthermore, static and transient loads can be combined in one single experiment. Improved diagnostics, such as IR-analyses, a spectrometer in the visible range, a photodiode array and acoustic emission shall finally contribute to a better understanding of the effects caused by extreme thermal shock loads.

After a brief description of the rationale of the material and geometry selection for each carbon armoured plasma facing components, this paper describes the European development of the two basic geometries, namely the monoblock and the hat tile. An overview of the non-destructive inspection techniques specifically developed for these components is also presented.

An overview is given of the wide range of activities contained within recent R&D being performed within Europe on In-vessel materials for structural, heat sink and plasma facing purposes for ITER. The effect of creep-fatigue interaction on the fatigue life of CuCrZr and the effects of irradiation on over-aged CuCrZr are given. In addition the lifetime of ITER components has been further investigated by the performance of in situ experiments with both neutrons and high-energy protons. The effect of hydrogen on the crack initiation fracture toughness of Ti is reported at a range of irradiation temperatures. The irradiation induced stress relaxation of Alloy 718 used for bolting applications is being studied and initial results will be described. Work in the area of plasma facing materials and re-welding issues will also be presented.