G. Vieider

Development of tungsten armor and bonding to copper for plasma-interactive components

Abstract For the highest sputtering threshold of all possible candidates, tungsten will be the most likely armor material in highly loaded plasma-interactive components of commercially relevant fusion reactors. The development of new materials, as well as joining and coating techniques are needed to find the best balance in plasma compatibility, lifetime, reliability, neutron irradiation resistance, and safety. Further important issues for selection are availability, costs of machining and production, etc. Tungsten doped with lanthanum oxide is a commercially available W grade for electrodes, designed for low electron work function, higher recrystallization temperature, reduced secondary grain growth, and machinability at relatively low costs. W–Re and related tungsten base alloys are preferred for application at high temperatures, when high strength, high thermal shock and recrystallization resistance are required. Due to the high costs and limited global availability of Re, however, the amount of such alloys in a commercial reactor should be kept low. Newly measured material properties up to high temperatures are presented for lanthanated and W–Re alloys, and the impact on fusion application is discussed. Recently developed coatings of chemical vapor deposited tungsten (CVD-W) on copper substrates have proven to be resistant to repeated thermal and shock loading. Layers of more than 5 mm, as required for the International Thermonuclear Experimental Reactor (ITER), became available. Vacuum plasma sprayed tungsten (VPS-W) in particular is attractive for its lower costs, and the potential of in situ repair. However, the advantage of sacrificial plasma-interactive tungsten coatings in long-term fusion devices has yet to be demonstrated. A durable and reliable joining of bulk tungsten to copper is needed to achieve an acceptable component lifetime in a fusion environment. The material properties of the copper alloys proposed for ITER, and their impact on the quality of bonding to tungsten is discussed. Future materials R&D should concern issues such as plasma compatibility, and above all neutron irradiation damage of promising tungsten–copper joints.

Armor and heat sink materials joining technologies development for ITER plasma facing components

An extensive program on the development of the joining technologies between armor (beryllium, tungsten and carbon fibre composites) and copper alloys heat sink materials for ITER plasma facing components (PFCs) has been carried out by ITER home teams. A brief review of this R&D program is presented in this paper. The critical problems related to these joints are described. Based on the results of this program and new requirements on the reduction the manufacturing cost of ITER PFC, reference technologies for use in ITER have been selected and recommended for further development.

ITER divertor, design issues and research and development ☆

Abstract Over the period of the ITER Engineering Design Activity (EDA) the results from physics experiments, modelling, engineering analyses and R&D, have been brought together to provide a design for an ITER divertor. The design satisfies all necessary requirements for steady state and transient heat flux, nuclear shielding, pumping, tritium inventory, impurity control, armour lifetime, electromagnetic loads, diagnostics, and remote maintenance. The design consists of 60 cassettes each comprising a cassette body onto which the plasma facing components (PFCs) are mounted. Each cassette is supported by toroidal rails which are attached to the vacuum vessel. For the PFCs the final armour choice is carbon-fibre-composite (CfC) for the strike point regions and tungsten in all remaining areas. R&D has demonstrated that CfC monoblocks can routinely withstand heat loads up to 20 MW m−2 and tungsten armour >10 MW m−2. Analysis and experiment show that a CfC armour thickness of ∼20 mm will provide sufficient lifetime for at least 1000 full power pulses. The thickness of the cassette body is sufficient to shield the vacuum vessel, so that, if necessary, rewelding is possible, and also provides sufficient stiffness against electromagnetically generated loads. The cassette design provides efficient and proven remote maintenance which should allow exchange of a complete divertor within ∼6 months.

ITER plasma facing components, design and development

Abstract The paper summarizes the collaborative effort of the ITER Conceptual Design Activity (CDA) on Plasma Facing Components (PFC) which focused on the following main tasks: (a) the definition of basic design concepts for the First Wall (FW) and Divertor Plates (DP), (b) the analysis of the performance and likely lifetime of these PFC designs including the identification of major critical issues, (c) the start of R&D work giving already first results, and the definition of the required further R&D program to support the contemplated ITER Engineering Design Activity (EDA). From the ITER CDA effort on PFC it is mainly concluded that: (a) the expected PFC operating conditions lead to design solutions at the limit of present technology in particular for the divertor, which may constrain the overall machine performance, (b) the development of convincing PFC designs requires an intensified R&D effort both on PFC technology and plasma physics.

Carbon fiber composites application in ITER plasma facing components

Abstract Carbon Fiber Composites (CFCs) are one of the candidate armour materials for the plasma facing components of the International Thermonuclear Experimental Reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R&D needs are critically discussed.

Overview of the EU Small Scale Mock-up Tests for ITER High Heat Flux Components

Abstract This task within the EU R&D for ITER was aimed at the development of basic manufacturing solutions for the high heat flux plasma facing components such as the divertor targets, the baffles and limiters. More than 50 representative small-scale mock-ups have been manufactured with beryllium, carbon and tungsten armour using various joining technologies. High heat flux testing of 20 of these mock-ups showed the carbon mono-blocks to be the most robust solution, surviving 2000 cycles at absorbed heat fluxes of up to 24 MW m−2. With flat armour tiles rapid joint failures occurred at 5–16 MW m−2 depending on joining technology and armour material. These test results serve as a basis for the selection of manufacturing options and materials for the prototypes now being ordered.

Neutron irradiation effects on the properties of carbon materials

Abstract Carbon-based materials are being widely used in present tokamaks, because of their favourable properties, especially with respect to their thermal shock resistance. However, the significant effects resulting from neutron irradiation on the thermal-physical, mechanical properties and on the tritium trapping capacity could preclude the utilization of carbon as a plasma facing material in forthcoming devices. Therefore, the effects of neutron irradiation are of primary concern in the use of carbon materials in the next step devices. This paper presents the latest results of experimental investigations on the effects of neutron irradiation on the thermal-physical, mechanical properties and tritium retention behaviour of various graphites and carbon/carbon fiber composites for the doses of 10 −3 –30 dpa and irradiation temperatures of 200–1500°C. The state of experimental investigation of the behaviour of plasma hydrogen isotopes in carbon materials is given. In particular, the issue of neutron-induced changes in tritium behaviour is critically analysed.

Reference testing of actively cooled mock-ups for the neutron-irradiation experiments PARIDE 3 and 4

A neutron irradiation campaign has been performed with new designs of high heat flux components. In parallel to this irradiation, reference tests have been carried out with un-irradiated samples of the same type. This paper reports on the testing of un-irradiated divertor mock-ups from tungsten and CFC attached to water-cooled heat sinks from CuCrZr.

European development of prototypes for ITER high heat flux components

The extensive EU research and development, on international thermonuclear experimental reactor (ITER) high heat flux (HHF) components aims at the demonstration of prototypes for the divertor and baffle with challenging operating requirements. The recent progress of this development is summarised in the paper, particularly concerning the manufacture and testing of mock-ups and prototypes. The available results demonstrate the feasibility of robust solutions with carbon and tungsten armour.

EU Results on Neutron Effects on PFC Materials

Among the low-Z materials, carbon and beryllium are primary candidates for use as plasma facing materials for the International Thermonuclear Experimental Reactor (ITER), because of extensive experience in their application for first wall and divertor plate protection in existing tokamaks. In addition, their excellent plasma performance has been demonstrated. Carbon based materials have been chosen for protection of high heat flux components, whilst beryllium has been proposed as the first wall material for ITER. However, as next generation D/T plasma devices, i.e. ITER, will produce intense neutron fluxes, substantial R&D is needed to elucidate the effects of neutron-induced damage on the microstructure and critical properties of these materials, e.g. thermal conductivity, swelling, and tritium trapping, because they could limit the use of these materials in the next generation fusion devices. Neutron induced changes in thermal conductivity, dimensional stability, mechanical properties as well as behaviour of tritium interaction are crucial problems which need to be better understood. The assessed neutron flux of ITER will be around 3.5-9.0 × 10 14 cm - 2 s - 1 for the first wall, whilst the neutron flux for the divertor is around 1-3 × 10 14 cm - 2 s - for which leads to a damage of around 10-20 dpa for the first wall and 3-6 dpa for the divertor for 1 full power year of operation. In the framework of European fusion R&D programs, an extensive effort on neutron effects on plasma facing component (PFC) materials is being undertaken. This paper presents the recent results of experiments performed to investigate the effects of neutron doses and irradiation temperature on the thermal conductivity, mechanical properties, dimensional stability and tritium inventory of various carbon based materials as well as beryllium. The consequences are discussed.