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Corre, Y.

Publications and source records attributed to Corre, Y..

7 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Survey of tungsten gross erosion from main plasma facing components in WEST during a L-mode high fluence campaign

An initial high fluence campaign was performed in WEST, in 2023, on the newly installed actively cooled tungsten divertor composed of ITER-grade monoblocks. The campaign consisted in the repetition of a 60 s long Deuterium L-mode pulse in attached divertor conditions, cumulating over 10000s of plasma exposure. A maximum deuterium fluence of approximately 5⋅1⁢026 m−2 was reached in the outer strike point region, representative of a few high performance ITER pulses. Gross tungsten erosion inferred from visible spectroscopy shows that the most eroded plasma facing component is the inner divertor target with rates ten times larger than on the outer divertor target. The outer midplane tungsten bumpers, located a few centimeters from the plasma, show gross erosion rates two times lower than at the outer divertor. We conclude that the outer midplane bumpers have a negligible contribution to the long range tungsten migration and deposition onto the lower divertor. The cumulated gross erosion rate on the inner divertor translates in an effective gross erosion thickness of about 20μ⁢m, while it is about 2μ⁢m for the outer divertor. Strikingly, these orderings coincide with the thickness of deposits found locally on the divertor: the exposed surfaces of high field side monoblocks are covered with several tens of μ⁢m tungsten deposits, while on the lower field side, few μ⁢m thin tungsten deposits are only found on the magnetically shadowed parts of monoblocks. The strong impact of those deposits on WEST operation, namely perturbation of surface temperature measurement with infra-red thermography, and the emission of flakes causing radiative perturbation of the confined plasma, calls for anticipating similar issues in ITER. In particular, the start of research operation shall consider the definition of a divertor erosion budget in order to anticipate the formation of deleterious deposits.

Fedorczak, N.↗
DOE OSTI2024

WEST full tungsten operation with an ITER grade divertor

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 Ion Cyclotron Resonance Heating 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 h. 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.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

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.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Experimental confirmation of efficient island divertor operation and successful neoclassical transport optimization in Wendelstein 7-X

We present recent highlights from the most recent operation phases of Wendelstein 7-X, the most advanced stellarator in the world. Stable detachment with good particle exhaust, low impurity content, and energy confinement times exceeding 100 ms, have been maintained for tens of seconds. Pellet fueling allows for plasma phases with reduced ion-temperature-gradient turbulence, and during such phases, the overall confinement is so good (energy confinement times often exceeding 200 ms) that the attained density and temperature profiles would not have been possible in less optimized devices, since they would have had neoclassical transport losses exceeding the heating applied in W7-X. This provides proof that the reduction of neoclassical transport through magnetic field optimization is successful. W7-X plasmas generally show good impurity screening and high plasma purity, but there is evidence of longer impurity confinement times during turbulence-suppressed phases.

neoclassical optimization↗
DOE OSTI2022

Operating a full tungsten actively cooled tokamak: overview of WEST first phase of operation

WEST is an MA class superconducting, actively cooled, full tungsten (W) tokamak, designed to operate in long pulses up to 1000 s. In support of ITER operation and DEMO conceptual activities, key missions of WEST are: (i) qualification of high heat flux plasma-facing components in integrating both technological and physics aspects in relevant heat and particle exhaust conditions, particularly for the tungsten monoblocks foreseen in ITER divertor; (ii) integrated steady-state operation at high confinement, with a focus on power exhaust issues. During the phase 1 of operation (2017–2020), a set of actively cooled ITER-grade plasma facing unit prototypes was integrated into the inertially cooled W coated startup lower divertor. Up to 8.8 MW of RF power has been coupled to the plasma and divertor heat flux of up to 6 MW m –2 were reached. Long pulse operation was started, using the upper actively cooled divertor, with a discharge of about 1 min achieved. This paper gives an overview of the results achieved in phase 1. Perspectives for phase 2, operating with the full capability of the device with the complete ITER-grade actively cooled lower divertor, are also described.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Interpretative transport modeling of the WEST boundary plasma: main plasma and light impurities

Understanding impurity transport in tokamak plasmas is crucial to control radiative losses and material migration in future magnetic fusion reactors. In this work we deploy the SolEdge2D-EIRENE code to model the boundary plasma in a WEST discharge, satisfactorily reproducing measurements of both upstream and divertor plasma conditions. Additionally, the spatial distribution of oxygen, studied here as a representative light impurity, is compared to vacuum ultraviolet spectroscopy measurements acquired with an oscillating line of sight. The simulation captures a key feature of the experiment, namely a factor of sime2 higher oxygen brightness in the inner divertor region compared to the outer one. This spatial asymmetry in oxygen concentration is interpreted by analyzing the balance of friction forces and thermal gradient forces that the light impurity exchanges with the main plasma.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗