Heat flux analysis of Type-I ELM impact on a sloped, protruding surface in the JET bulk tungsten divertor

Tungsten (W) melting due to transient power loads, for example those delivered by edge localised modes (ELMs), is a major concern for next step fusion devices. A series of experiments has been performed on JET to investigate the dynamics of Type-I ELM-induced transient melting. Following initial exposures in 2013 of a W-lamella with sharp leading edge in the bulk W outer divertor, new experiments have been performed in 2016-2017 on a protruding W-lamella with a 15 degrees slope, allowing direct and spatially resolved (0.85 mm/pixel) observation of the top surface using the IR thermography system viewing from the top of the poloidal cross-section. Thermal and IR analysis have already been conducted assuming the geometrical projection of the parallel heat flux on the W-lamellas, thus ignoring the gyro-radius orbit of plasma particles. Although it is well justified during L-mode or inter-ELM period, the hypothesis becomes questionable during ELM when the ion Larmor radius is larger. The goal of this paper is to extend the previous analysis based on the forward approach to the H-mode discharges and investigate in particular the gyro-radius effect during the Type-I ELMs, those used to achieve transient melting on the slope of the protruding W-lamella. Surface temperatures measured by the IR camera are compared with reconstructed synthetic data from 3D thermal modelling using heat loads derived from optical projection of the parallel heat flux (ignoring the gyro-radius orbit), 2D gyro-radius orbit and particle-in-cell (PIC) simulations describing the influence of finite Larmor-radius effects and electrical potential on the deposited power flux. Results show that the ELM power deposition behaves differently than the optical projection of the parallel heat flux, contrary to the L-mode observations, and may thus be due to the much larger gyro-orbits of the energetic ELM ions in comparison to L-mode or inter-ELM conditions.EURATOM 63305

Modeling of COMPASS tokamak divertor liquid metal experiments

Two small liquid metal targets based on the capillary porous structure were exposed to the divertor plasma of the tokamak COMPASS. The first target was wetted by pure lithium and the second one by a lithium-tin alloy, both releasing mainly lithium atoms (sputtering and evaporation) when exposed to plasma. Due to poorly conductive target material and steep surface inclination (implying the surface-perpendicular plasma heat flux 12-17 MW/m(2)) for 0.1-0.2 s, the LiSn target has reached 900 degrees C under ELMy H-mode. A model of heat conduction is developed and serves to evaluate the lithium sputtering and evaporation and, thus, the surface cooling by the released lithium and consequent radiative shielding. In these conditions, cooling of the surface by the latent heat of vapor did not exceed 1 MW/m(2). About 10(19) lithium atoms were evaporated (comparable to the COMPASS 1 m(3) plasma deuterium content), local Li pressure exceeded the deuterium plasma pressure. Since the radiating Li vapor cloud spreads over a sphere much larger than the hot spot, its cooling effect is negligible (0.2 MW/m(2)). We also predict zero lithium prompt redeposition, consistent with our observation.

Overview of power exhaust experiments in the COMPASS divertor with liquid metals

Power handling experiments with a special liquid metal divertor module based on the capillary porous system technology were performed in the tokamak COMPASS. The performance of two metals (Li and LiSn alloy) were tested for the first time in a divertor under ELMy H-mode conditions. No damage of the capillary mesh and a good exhaust capability were observed for both metals in two separate experiments with up to 12 MW/m(2) of deposited perpendicular, inter-ELM steady-state heat flux and with ELMs of relative energy similar to 3% and a local peak energy fluence at the module similar to 15 kJ.m(-2). No droplets were directly ejected from the mesh top surface and for the LiSn experiment, no contamination of the core and SOL plasmas by Sn was observed. The elemental depth profile analysis of 14 stainless-steel samples located around the vacuum vessel for each experiment provides information about the migration of evaporated/redeposited liquid elements

Dynamics of the edge transport barrier at plasma biasing on the CASTOR tokamak

A clear and reproducible transition to a regime with an improved particle confinement is routinely observed on the CASTOR tokamak, if the biasing electrode is inserted deep enough into the plasma (r/a ~ 0.5) and biased up to +250V. The steepening of the radial profiles of the plasma density and potential demonstrate the formation of a transport barrier just inside the last closed flux surface. Fast relaxations of the edge plasma parameters, with a frequency of about 10 kHz, are observed when the average radial electric field within the barrier prevails values of about 20 kV/m. A detailed analysis of the spatial-temporal behavior of these relaxations is presented.На токамаке CASTOR устойчиво наблюдается отчетливый и воспроизводимый переход в режим с улучшенным удержанием частиц при достаточно глубоком введении в плазму (r/a - 0,5) электрода, на который подаётся смещение до +250 В. У кручение радиальных профилей плотности плазмы и потенциала свидетельствуют об образовании транспортного барьера сразу внутри последней замкнутой магнитной поверхности. Наблюдаются быстрые релаксации параметров краевой плазмы с частотой около 10 кГц, когда среднее радиальное электрическое поле внутри барьера превышает значение около 20 кВ/м. Приводится обстоятельный анализ пространственно-временного поведения этих релаксаций.На токамаці CASTOR стійко спостерігається чіткий та відтворений перехід до режиму з поліпшеним утриманням частинок при достатньо глибокому введенні у плазму (r/a ~ 0,5) електрода, на який подано зміщення до +250 В. Радіальні профілі густини плазми та потенціалу стають більш крутими, що свідчить про створення потенціального бар'єру зразу усередині останньої замкненої магнітної поверхні. Спостерігаються швидкі релаксації параметрів крайової плазми з частотою порядку 10 кГц, коли середнє радіальне електричне поле усередині бар'єра перевищує значення приблизно 20 кВ/м. Надається докладний аналіз просторово-часової поведінки цих релаксацій

Modelling of the effect of ELMs on fuel retention at the bulk W divertor of JET

Effect of ELMs on fuel retention at the bulk W target of JET ITER-Like Wall was studied with multi-scale calculations. Plasma input parameters were taken from ELMy H-mode plasma experiment. The energetic intra-ELM fuel particles get implanted and create near-surface defects up to depths of few tens of nm, which act as the main fuel trapping sites during ELMs. Clustering of implantation-induced vacancies were found to take place. The incoming flux of inter-ELM plasma particles increases the different filling levels of trapped fuel in defects. The temperature increase of the W target during the pulse increases the fuel detrapping rate. The inter-ELM fuel particle flux refills the partially emptied trapping sites and fills new sites. This leads to a competing effect on the retention and release rates of the implanted particles. At high temperatures the main retention appeared in larger vacancy clusters due to increased clustering rate