Component development for operation in a large-scale fusion device requires thorough testing and qualification for the intended operational conditions. In particular environments are necessary which ...are comparable to the real operation conditions, allowing at the same time for in situ/in vacuo diagnostics and flexible operation, even beyond design limits during the testing. Various electron and neutral particle devices provide the capabilities for high heat load tests, suited for material samples and components from lab-scale dimensions up to full-size parts, containing toxic materials like beryllium, and being activated by neutron irradiation. To simulate the conditions specific to a fusion plasma both at the first wall and in the divertor of fusion devices, linear plasma devices allow for a test of erosion and hydrogen isotope recycling behavior under well-defined and controlled conditions. Finally, the complex conditions in a fusion device (including the effects caused by magnetic fields) are exploited for component and material tests by exposing test mock-ups or material samples to a fusion plasma by manipulator systems. They allow for easy exchange of test pieces in a tokamak or stellarator device, without opening the vessel. Such a chain of test devices and qualification procedures is required for the development of plasma-facing components which then can be successfully operated in future fusion power devices. The various available as well as newly planned devices and test stands, together with their specific capabilities, are presented in this manuscript. Results from experimental programs on test facilities illustrate their significance for the qualification of plasma-facing materials and components. An extended set of references provides access to the current status of material and component testing capabilities in the international fusion programs.
Tungsten heavy alloy (97 wt% W, 2 wt% Ni, 1 wt% Fe) was investigated as an alternative for tungsten (W) as plasma facing material. It is produced commercially by several companies and compared to ...bulk W it is readily machinable and considerably cheaper. In order to qualify the material for use in the divertor of the mid-size tokamak ASDEX Upgrade (AUG) dedicated laboratory investigations as well as high heat flux tests in the neutral beam facility GLADIS were performed. These investigations revealed that the thermal conductivity at high temperature is close to that of W, the magnetisation is small and saturates already at low magnetic field and the hydrogen retention is similarly low as that of W. In high heat flux tests at power densities up to 20 MWm−2 no failure was observed up to the melting temperature (≈1500∘C) of the binder phase. Even at surface temperatures of up to 2200 °C the mechanical integrity was sustained. Mechanical tests confirm the ductile behaviour of the W heavy alloy at room temperature and finite element analyses using the aforementioned data suggest a lower tendency for cracking. The increase of the long term dose-rate resulting from the activation of Ni under neutron irradiations appears to be moderate. During the 2017 campaign more than one fifth of the AUG divertor tiles consisted of W heavy alloy. Under nominal operation conditions the tiles showed no macroscopic failure and no increased Fe/Ni influx into the plasma was detected. Even though a few tiles showed strong melting at the edges due to accidental misalignment no failure due to cracking was observed.
Divertor target is one of the most critical in-vessel components in a fusion power plant being in charge of particle and power exhaust. The targets are exposed to severe thermal loads produced by ...steady bombardment of impinging plasma flux. Since 2014, integrated R&D efforts have been continued aiming at developing a design concept and high-heat-flux (HHF) technologies for divertor targets of the European DEMO reactor. Recently, the second round (2017–2019) of the R&D program has been concluded. As in the first R&D round, five water-cooled target design concepts were further developed and evaluated. Fabrication technologies were improved reaching a consolidated production quality. Extensive HHF tests were conducted using small-scale mock-ups for extended loading regimes (heat flux: 20–32MW/m²). Comparative studies were performed to investigate effects of copper interlayer thickness (0.1–1 mm) and different tungsten armor materials. In the present paper, the final results of the second round HHF testing campaign are reported. The HHF performance of each design variant is discussed based on in-situ diagnostic data (infrared thermography), ultrasonic inspection images and postmortem metallographic micrographs. All monoblock-type design concepts passed the specified qualification criterion (≥500 pulses at 20 MW/m², coolant: 130 °C) without any failure or armor cracking. Moreover, two of them (ITER-like and composite pipe) remained fully intact even under 25 MW/m² (100 pulses) and 32 MW/m² (5 pulses).
•New design of flush-mounted Langmuir probes in the WEST divertor made of tantalum.•Great flushness of the probes within specifications and confirmed during experiments.•Probes successfully measuring ...since the WEST operation start with consistent results.•Probes still in operation after 22000s of plasma and 1.5 GJ of additional RF power.
The design of the Langmuir probes for the lower divertor of the WEST tokamak is presented, in which uncooled, tungsten-coated graphite targets were installed for the first phase of operation in anticipation of full-tungsten, actively cooled, ITER-like divertor components. Due to the long plasma discharges and high heat loads, the probe tips are flush with the divertor tiles in which they are embedded, as it is foreseen and indeed required in ITER. Tantalum metal was chosen for the probe tips because of its high melting point, and ductility at room temperature which allows simple crimping to electrically connect them to cables, as well as being compatible with the full metal environment of WEST. High heat flux tests of Ta Langmuir probe prototypes confirmed the robustness of the design for energy flux densities up to 8 MW/m2 for 5.5 s, and at lower 3.5 MW/m² energy flux density for 40 s long discharges. The probes were operational from the first discharge in WEST and the good alignment of all probes with the surrounding surface was confirmed by a good match of each adjacent probe's data when the strike points were swept across the target.
The new ion beam facility GLADIS started the operation at IPP Garching. The facility is equipped with two individual 1.1 MW power ion sources for testing actively cooled plasma facing components ...under high heat fluxes. Each ion source generates heat loads between 3 and 55 MW/m2 with a beam diameter of 70 mm at the target position. These parameters allow effective testing from probes to large components up to 2 m length. The high heat flux allows the target to be installed inclined to the beam and thus increases the heated surface length up to 200 mm for a heat flux of 15 MW/m2 in the standard operating regime. Thus the facility has the potential capability for testing of full scale ITER divertor targets. Heat load tests on the WENDELSTEIN 7-X pre-series divertor targets have been successfully started. These tests will validate the design and manufacturing for the production of 950 elements.
Material issues pose a significant challenge for future fusion reactors like DEMO. When using materials in a fusion environment a highly integrated approach is required. Damage resilience, power ...exhaust, as well as oxidation resistance during accidental air ingress are driving issues when deciding for new materials. Neutron induced effects e.g. transmutation adding to embrittlement are crucial to material performance. Here advanced materials such as tungsten fibre-reinforced tungsten Wf/W and fibre-reinforced copper Wf/Cu composites could allow the step towards a fusion reactor. Recent developments in the area Wf/W mark a possible path towards a component mock-up early enough for utilisation in DEMO. High heat-flux tests show that having short fibres at the exposed surface leads to their selective erosion and melting. Initial tests in the linear plasma device PSI-2 confirm this behaviour.
•Measurement of magnetisation and thermal conductivity of W–Ni/Fe heavy alloys (D185, HPM1850).•Successful high heat flux testing of W heavy alloys with power densities of up to 20MW/m3.•Exposure of ...W-heavy alloys in the divertor of ASDEX Upgrade in discharges with up 26MW of heating power.
An alternative solution for tungsten as a plasma facing material could be the use of W heavy alloys as they are produced commercially by several companies. They consist of up to 97% W and Ni/Fe (or Ni/Cu) admixtures, they are readily machinable and considerably cheaper than bulk tungsten. Their major drawbacks in view of the application in fusion experiments are the rather low melting temperature and their magnetic properties (in case of a Ni/Fe binder phase). In a first step W heavy alloys from two manufacturers were investigated concerning their thermal and magnetic properties and subjected to screening tests and cyclic loading in the high heat flux test facility GLADIS with up to 20MWm−2 and surface temperatures of up to 2200°C, showing no macroscopic failure. SEM investigations show a segregation of Ni and Fe at the top surface after the thermal overloading, but no signs of micro-cracking. The long-term behaviour of a W–Ni/Fe tile under plasma and electromagnetic load was investigated in ASDEX Upgrade using its divertor manipulator. The tile was exposed in discharges with record values of injected energy and power. Despite the observed surface modifications (Ni/Fe segregation) the W heavy alloys seem to provide a pragmatic and cost-effective alternative to bulk W tiles in the divertor of ASDEX Upgrade.
Powder metallurgy tungsten and plasma-sprayed tungsten samples were exposed to high heat and particle fluxes in order to study changes in the surface morphology. Actively water-cooled and ...adiabatically loaded samples were irradiated with H, He, and mixed 90% H/10% He neutral beams at power densities of 2MWm−2 and 10MWm−2 with fluences up to 3×1025m−2. For He irradiation of adiabatically loaded samples a peak surface temperature of 2100°C was reached. Under this condition the growth of a coral-like surface structure modification due to bubble formation was the dominant process observed. In all other cases physical sputtering was the surface modification process. For hydrogen irradiation of actively cooled samples with much lower surface temperatures no formation of blisters was observed. Instead, we found small nano-sized pores.
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