Fuel retention monitoring in tokamak walls requires the development of remote composition analysis methods such as laser-induced breakdown spectroscopy (LIBS). The present study investigates the ...feasibility of the LIBS method to analyse the composition and fuel retention in three samples from WEST divertor erosion marker tiles after the experimental campaign C3. The investigated samples originated from tile regions outside of strong erosion and deposition regions, where the variation of thin deposit layers is relatively small and facilitates cross-comparison between different analysis methods. The depth profiles of main constituents W, Mo and C were consistent with depth profiles determined by other composition analysis methods, such as glow-discharge optical emission spectroscopy (GDOES) and secondary ion mass spectrometry (SIMS). The average LIBS depth resolution determined from depth profiles was 100 nm/shot. The averaging of the spectra collected from multiple spots of a same sample allowed us to improve the signal-to-noise ratio, investigate the presence of fuel D and trace impurities such as O and B. In the investigated tile regions with negligible erosion and deposition, these impurities were clearly detectable during the first laser shot, while the signal decreased to noise level after a few subsequent laser shots at the same spot. LIBS investigation of samples originating from the deposition regions of tiles may further clarify LIBS’ ability to investigate trace impurities.
•Infrared quantitative thermography.•Damaged PFC.•ITER like tungsten PFC.•heat transfer modelling.
The consequences of tungsten (W) cracking on divertor lifetime and plasma operation are high ...priority issues for ITER. One actively cooled ITER-like plasma facing unit (PFU) has been pre-damaged in a High Heat Flux (HHF) facility before its installation in WEST in order to assess the damage evolution after tokamak plasma exposure. The resulting pre-damage exhibits micrometer-size crack network and high roughness on the tungsten monoblock (MB) top surface. A total of 10 MBs, equally distributed on the low and high field sides of the lower divertor, have been pre-damaged among the 35 radially aligned MBs characteristic of the WEST PFU. Subsequent plasma exposure was carried out, from the first breakdown achieved in WEST (in 2017) until the removal of the damaged PFU three years later (2020). On top of the whole WEST plasma exposure (covering C1-C4 experimental campaigns), a dedicated experiment has also been performed in the frame of the EU work program to maximize the power and energy loads on one of the damaged MBs featuring a “crack network” pattern. The MB top surface, including both “crack network” damage and “healthy” (undamaged) areas, was monitored with a high spatial resolution IR camera to detect any potential evolution of the damage pulse after pulse. This paper describes the full plasma exposure achieved in the WEST tokamak (including large number of steady-state and transient heat loading cycles), the dedicated “damaged PFU exposure” experiment together with the experimental results (heat loading on the damaged MBs). Post-mortem measurement reveals significant broadening of the cracks and new cracks in the electron beam loaded area only.
Abstract Within the 9th European Framework programme, since 2021 EUROfusion is operating five tokamaks under the auspices of a single Task Force called "Tokamak Exploitation”. The goal is to benefit ...from the complementary capabilities of each machine in a coordinated way and help in developing a scientific output scalable to future largre machines. The programme of this Task Force ensures that ASDEX Upgrade, MAST-U, TCV, WEST and JET (since 2022) work together to achieve the objectives of Missions 1 and 2 of the EUROfusion Roadmap: i) demonstrate plasma scenarios that increase the success margin of ITER and satisfy the requirements of DEMO and, ii) demonstrate an integrated approach that can handle the large power leaving ITER and DEMO plasmas. The Tokamak Exploitation task force has therefore organized experiments on these two missions with the goal to strengthen the physics and operational basis for the ITER baseline scenario and for exploiting the recent plasma exhaust enhancements in all four devices (PEX: Plasma EXhaust) for exploring the solution for handling heat and particle exhaust in ITER and develop the conceptual solutions for DEMO. The ITER Baseline scenario has been developed in a similar way in ASDEX Upgrade, TCV and JET. Key risks for ITER such as disruptions and run-aways have been also investigated in TCV, ASDEX Upgrade and JET. Experiments have explored successfully different divertor configurations (standard, super-X, snowflakes) in MAST-U and TCV and studied tungsten melting in WEST and ASDEX Upgrade. The input from the smaller devices to JET has also been proven successful to set-up novel control schemes on disruption avoidance and detachment.
Abstract The paper presents the energy balance of 602 pulses from four different experimental campaigns of the WEST tokamak. Different magnetic configuration has been studied with lower single null ...and upper single null configuration with deuterium or helium plasmas. The energy balance is closed with an imbalance of about 5% of the total injected energy for most of the campaigns and for the different magnetic configurations. The distribution over the whole machine is shown with the outer first wall receiving the most part of the energy due to its large surface area with about 30% of the total heat load, and the divertor with 25% due to the heat loads deposited by the convected power in the scrape off layer. Finally, the tomography inversion of the bolometry measurement allows to disentangle the contribution of the radiated and convected power in the energy absorbed by each type of PFC. Showing that in upper single null configuration about 63% of the available energy in the SOL is deposited in the upper divertor through convected heat loads, while in lower single null this value is spread over the lower divertor with 45% and the baffle and upper divertor with about 10% for both.
In future thermo-nuclear fusion devices, such as ITER (International Thermonuclear Experimental Reactor), the interaction of the plasma with surrounding materials in the vacuum vessel constitutes one ...of the main remaining engineering problems. The choice of materials is a crucial point, which will determine issues such as the plasma facing components lifetime before refurbishment or the tritium inventory build up in the vessel, which should be limited for safety reasons. In order to tackle these issues, the European Task Force on Plasma–Wall Interaction has been implemented in the frame of EFDA (European Fusion Agreement) in the fall 2002 with the aim “to provide ITER with information concerning lifetime-expectations of the divertor target plates and tritium inventory build-up rates in the foreseen starting configuration and to suggest improvements, including material changes, which could be implemented at an appropriate stage.”
The EU-PWI-TF brings together the efforts of 24 European associations in the following fields of investigation:•Material erosion and transport in tokamaks.•Tritium inventory and removal.•Transient heat loads on plasma facing components.•Dust production and removal.•Associated modelling and diagnostic development.
This paper will present the organisation of the EU-PWI-TF. It will provide examples for the multitude of surface processes in Plasma–Wall Interaction and present the status of knowledge concerning material erosion and hydrogen retention for the choice of ITER materials (Beryllium, Carbon and Tungsten).
► The WEST programme is a unique opportunity to experience the industrial scale manufacture of tungsten plasma-facing components similar to the ITER divertor ones. ► In Tore Supra, it will bring ...important know how for actively cooled W divertor operation. ► This can be done by a reasonable modification of the Tore Supra tokamak. ► A fast implementation of the project would make this information available in due time. ► This allows a significant contribution to the W ITER divertor risk minimization in its manufacturing and operation phase.
The WEST programme consists in transforming the Tore Supra tokamak into an X point divertor device, while taking advantage of its long discharge capability. This is obtained by inserting in vessel coils to create the X point while adapting the in-vessel elements to this new geometry. This will allow the full tungsten divertor technology to be used on ITER to be tested in anticipation of its use on ITER under relevant heat loading conditions and pulse duration. The early manufacturing of a significant industrial series of ITER-similar W plasma-facing units will contribute to the ITER divertor manufacturing risk mitigation and to that associated with early W divertor plasma operation on ITER.
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