Overview of the first Wendelstein 7-X long pulse campaign with fully water-cooled plasma facing components

After a long device enhancement phase, scientific operation resumed in 2022. The main new device components are the water cooling of all plasma facing components and the new water-cooled high heat flux divertor units. Water cooling allowed for the first long-pulse operation campaign. A maximum discharge length of 8 min was achieved with a total heating energy of 1.3 GJ. Safe divertor operation was demonstrated in attached and detached mode. Stable detachment is readily achieved in some magnetic configurations but requires impurity seeding in configurations with small magnetic pitch angle within the edge islands. Progress was made in the characterization of transport mechanisms across edge magnetic islands: Measurement of the potential distribution and flow pattern reveals that the islands are associated with a strong poloidal drift, which leads to rapid convection of energy and particles from the last closed flux surface into the scrape-off layer. Using the upgraded plasma heating systems, advanced heating scenarios were developed, which provide improved energy confinement comparable to the scenario, in which the record triple product for stellarators was achieved in the previous operation campaign. However, a magnetic configuration-dependent critical heating power limit of the electron cyclotron resonance heating was observed. Exceeding the respective power limit leads to a degradation of the confinement

Simulation study of divertor geometry for COMPASS Upgrade tokamak

Simulation study of divertor geometry for COMPASS Upgrade tokamakI. Borodkina1,2, A. Kukushkin2,3, S. Wiesen4, D. Boeyaert4, M. Imrisek1,L. Kripner1, M. Peterka1, R. Dejarnac1, M. Komm11Institute of Plasma Physics of the CAS, Za Slovankou 3, 182 00 Prague 8, Czech Republic2National Research Nuclear University “MEPhI”, Moscow, Russia3National Research Center “Kurchatov Institute”, Moscow, Russia4Forschungszentrum Jülich GmbH, Institut für Energie-und Klimaforschung –Plasmaphysik, Partner of the Trilateral Euregio Cluster (TEC), 52425 Jülich, [email protected] development of divertor design with a reliable solution for the power and impurity particle exhaust is one of the important challenge towards the realization of COMPASS Upgrade (COMPASS-U) project. COMPASS-U with its high plasma and neutral density is of particular interest for ITER in terms of similar divertor plasma and neutral parameters, as well as predicted power decay length and peak power loads to the divertor targets [1].It is essential to efficiently dissipate power in the divertor, probably necessitating operation in a detached regime,to ensure the maximum steady-state power load at the divertor target below 10 ∼15 MW/m2and to maintain a low electron temperature at the target plates~ 5-10eVto suppress erosion.In this contributionwe report on the first systematic modelling examination ofthe effect of different outer and inner divertor target angles, adivertor closure and pumplocations on the main parameters of the COMPASS-U divertor plasma. The simulations are carried outby using the 2D edge plasma code packages SOLPS4.3[2] and SOLPS-ITER [3] for pure D plasma with the fixed anomalous cross-fieldcoefficients corresponded to the predicted power decay lengths. The single-null and the asymmetric double-null magnetic configurations are used with the magnetic equilibriums provided by the FIESTA [4] code for some COMPASS-U scenarios elaborated by the scaling-law based METIS code[5].Several divertorconfigurationswithvarioustargetinclinationanglesaremodelledtostudytheeffectofdivertor closure on detachment, the divertor radiated power and thereforethe peak heat flux density at the divertor target.Impurity seeded cases withneon as the radiating impurity are also considered. The effect ofNe seeding,resulting in anincrease of the total radiative fractionby about two times,on the plasma parameters and the fluxes in the divertor for different divertor configurations is presented and discussed.The neutral gas pressure in the divertor is often considered to be a key control parameter for divertor conditions as neutral gas in divertor helps reaching low plasma temperatures and reduce peak heat fluxes trough power spreading at the divertor targets. SOLPS simulations for COMPASS-U baseline scenario with Ip=2MA and Bt=5T are carried out with the simulation grid includedthe sub-divertor module for more realistic account for neutral gas pressure. The effect of cryo-pump position and widths of gaps between the divertor tiles on neutral gas pressure in divertor is investigated.[1] R. Panek, et al., Fusion Eng. Des.123 (2017)11–16; [2] A.S. Kukushkin, H.D. Pacher, V. Kotov et al., Fusion Eng. Des. 86, 2865 (2011); [3] X. Bonnin et al., Plasma Fusion Res. 11, 1403102 (2016); [4] G. Cunningham, Fusion Eng. Des. 88 (2013) 3238–3247;[5] J.F. Artaud, V. Basiuk, F. Imbeaux, M. Schneider, J. Garcia, et al., Nucl. Fusion 50 (2010)04300

Assessment of plasma edge transport in Neon seeded plasmas in disconnected double null configuration in EAST with SOLPS-ITER

Assessment of plasma edge transport in Neon seeded plasmas in disconnected double null configuration in EAST with SOLPS-ITERD. Boeyaert1,3, S. Wiesen1, M. Wischmeier2, W. Dekeyser3, S. Carli3, L. Wang4, F. Ding4, K. Li4, Y. Liang1,4, M. Baelmans3, and the EAST-teama1Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, D-52425 Jülich, Germany2Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching, Germany3KU Leuven, Department of Mechanical Engineering, Celestijnenlaan 300, 3001 Leuven, Belgium4Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, People’s Republic of China aSee appendix of Wan B.N. et al 2019 Nucl. Fusion 59 112003 [email protected] and particle exhaust is essential for future nuclear fusion reactors [1]. This exhaust is determined by the perpendicular/radial transport inside the Scrape-Off Layer (SOL) which include drifts and currents, neutral kinetics, radiation and (radial) anomalous transport. Under high power conditions in future all-metal fusion devices like ITER or DEMO, extrinsic impurity seeding is required to induce divertor detachment through impurity radiation. Due to the lack of surface chemistry, noble gases like neon (Ne) are key candidates as main radiator. Besides JET [2], EAST is the only tokamak that currently handles stable H-modes with Neseeding in a metallic environment as the main (upper) divertor is made out of tungsten. This device has the flexibility to do both upper single null (USN) and double null (DN) configurations with the latter showing good prospects to handle exhaust with good core confinement [3]. USN discharges in EAST are setup as disconnected DN (DDN) with a large separation between separatrices upstream (drsep) of about ~ 2cm. This causes a remaining influx of eroded C impurities from the lower (non-active) divertor. This contribution analyzes Ne seeded and unseeded DDN deuterium discharges at EAST with decreasing drsep, both with experimental data from EAST and SOLPS-ITER simulations [4] . Ne seeded discharges in H-mode from the 2019 EAST campaign are studied (heating power Pheat = 2.5 MW, plasma current Ip = 0.4 MA and toroidal field Bt = 2.4 T). For the first time, a DDN configuration with divertor Te-feedback for the Ne puff strength was attempted to achieve steady divertor conditions. A radiative fraction Prad/Pheat of up to 30% was achieved with Ne seeding while 10% is achieved without. Ne in all cases was injected from the upper (active) outer target and a significant target temperature drop was identified from the Langmuir probes. The effect of a DDN configuration however is limited as the upstream power scrape-off width (~0.5cm [5]) is significantly smaller compared to the achieved separation between the separatrices (drsep ~ 1.5 cm). SOLPS-ITER simulations are being carried out to interpret the experimental results to understand the influence of the different transport types in the SOL for a dissipative DDN divertor geometry with and without Ne seeding and C erosion. The model includes for the first time for EAST fluid drifts and edge currents in the SOL. SOLPS-ITER will be used to predict the performance of a Ne seeded EAST DN divertor (with drsep = 0) being closest to one of theconsidered DEMO divertor geometries. [1] M. Wischmeier, et al., J. Nucl. Mater. 22-29 (2015) 463; [2] S. Glöggler, et al., Nucl. Fusion. 126031 (2019) 59 ; [3] H. Meyer, et al., Nucl. Fusion. 64-72 (2006) 46; [4] S. Wiesen et al., J. Nucl. Mat. 480-484 (2015) 463 ; [5] T. Eich, et al., Nucl. Fusion 093031 (2013) 5

Towards assessment of plasma edge transport in Neon seeded plasmas in disconnected double null configuration in EAST with SOLPS-ITER

Energy dissipation in the plasma edge is key for future tokamaks. The potential of neon as radiating seeding species in disconnected double null (DDN) configuration is assessed in EAST discharges in high confinement mode (H-mode). As the separation between the two separatrices in the studied DDN discharges is minimum 1.5 cm, the configuration is effectively a single null configuration, and the benefits of the double null topology are minimal. Neon seeding, on the other hand, has a favourable effect: both the target heat flux and the divertor temperature decrease more than five-fold with increased seeding rate in high-recycling conditions. Interpretive edge plasma simulations with SOLPS-ITER in support of ongoing transport analysis are presented. For the unseeded case the numerical results agree with the experimental data within a factor two for the target temperature conditions and measured neutral pressures in the active divertor. The key for achieving good agreement is a suitable selection of coefficients for anomalous transport and neutral conductances between the upper cryopump and the main chamber