Testing of ITER-grade plasma facing units in the WEST tokamak: Progress in understanding heat loading and damage mechanisms

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In future fusion reactors, extended melt pools in combination with strong plasma-induced accelerations, suggest that metallic melt could reach the gaps between castellated plasma-facing components, potentially accompanied by profound changes in their mechanical response. The first results of a combined experimental and modelling effort to elucidate the physics of melt transport across gaps are presented. Transient melting of specially designed tungsten samples featuring toroidal gaps has been achieved in ASDEX Upgrade providing direct evidence of gap bridging. Detailed modelling with the MEMENTO melt dynamics code is reported. Empirical evidence and simulations reveal that the presence of gaps can be safely ignored in macroscopic melt motion predictions as well as that the re-solidification limited melt spreading facilitates gap bridging and leads to poor melt attachment. The findings are discussed in the context of ITER and DEMO.

Section: Introductionmentioning

confidence: 99%

Metallic melt transport across castellated tiles

Ratynskaia,

Paschalidis,

Krieger

et al. 2024

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WEST full tungsten operation with an ITER grade divertor

Bucalossi,

Ekedahl,

Achard

et al. 2024

The mission of WEST (tungsten-W Environment in Steady-state Tokamak) is to explore long pulse operation in a full tungsten (W) environment for preparing next-step fusion devices (ITER and DEMO) with a focus on testing the ITER actively cooled W divertor in tokamak conditions. Following the successful completion of phase 1 (2016-2021), phase 2 started in December 2022 with the lower divertor made entirely of actively cooled ITER-grade tungsten mono-blocks. A boronization prior the first plasma attempt allowed for a smooth startup with the new divertor. Despite the reduced operating window due to tungsten, rapid progress has been made in long pulse operation, resulting in discharges with a pulse length of 100 s and an injected energy of around 300 MJ per discharge. Plasma startup studies were carried out with equatorial boron nitride limiters to compare them with tungsten limiters, while ICRH assisted startup was attempted. High fluence operation in attached regime, which was the main thrust of the first campaigns, already showed the progressive build up of deposits and appearance of dust, impacting the plasma operation as the plasma fluence increased. In total, the cumulated injected energy during the first campaigns reached 43 GJ and the cumulated plasma time exceeded 5 hours. Demonstration of controlled X-Point Radiator regime is also reported, opening a promising route for investigating plasma exhaust and plasma-wall interaction issues in more detached regime. This paper summarises the lessons learned from the manufacturing and the first operation of the ITER-grade divertor, describing the progress achieved in optimising operation in a full W environment with a focus on long pulse operation and plasma wall interaction.

Evolution of pre-damaged ITER grade plasma facing components under WEST plasma exposure: smoothing and tungsten sources

Richou,

Corre,

Diez

et al. 2024

The evaluation of the impact of plasma-facing units (PFUs) damage on subsequent plasma operation is an important issue for ITER. This is one of the topic addressed by the operation of dedicated PFUs in WEST. From the start of WEST in 2017, few ITER grade divertor plasma-facing units (PFUs) were tested in the lower divertor for short plasma exposure (experimental campaigns C1-C5). A PFU is an assembly of tungsten blocks on CuCrZr tube via pure copper interlayer. Since 2022, WEST operates with a lower divertor fully equipped with ITER grade PFUs (experimental campaign C6-C7) with long duration plasma exposure capabilities. To test the ageing of the ITER-grade PFUs on an accelerated basis, well controlled damage was implemented on purpose with electron beam loading on the plasma facing surface of two different PFUs (pre-damage#1 and #2 featuring different level of damage). Damage tackled in this study is induced by the loading with numerous transient heat loading events to mimic the ELM impact on the ITER grade PFU. This paper presents the damage before and after WEST plasma exposure and shows the tungsten sources and temperature of the block measured during the operation. On damaged area, surface roughening is heterogeneous. With WEST campaigns (C3, C4 and C7) respectively ended in (2018, 2019 and 2023) a global smoothing of the pre-damaged surfaces, in particular those located in the high loaded erosion area, is observed. No evolution of the crack is observed in the electron beam loaded area, while new cracking are observed on non pre-damaged area on the tungsten top surfaces in highly loaded areas (outer strike point) after significant plasma duration and cumulated energy (first observation in C4 and tends to be confirmed with preliminary observations after the C7 campaign). Smoothing is less pronounced after the 2018 campaign, which may be due to the related short plasma duration. The surface smoothing of blocks also depends on their poloidal position. For PFU which has no shaping, toroidal positioning within a block has no influence on the surface smoothing. After 2019 and 2023 campaigns, equivalent relative smoothing is obtained (between 19 and 32%). For the explored plasma conditions, damage has no influence on the maximum brightness of W.