A5003875003- Magnetic confinement fusion research
- Fusion materials and technologies
- Superconducting Materials and Applications
- Laser-Plasma Interactions and Diagnostics
- Nuclear reactor physics and engineering
- Nuclear Materials and Properties
- Plasma Diagnostics and Applications
- Ionosphere and magnetosphere dynamics
- Particle accelerators and beam dynamics
- Nuclear Physics and Applications
- Laser-induced spectroscopy and plasma
- Cyclone Separators and Fluid Dynamics
- Atomic and Molecular Physics
- Nuclear materials and radiation effects
- Particle Dynamics in Fluid Flows
- Granular flow and fluidized beds
- Advanced materials and composites
- High-Energy Particle Collisions Research
- Solar and Space Plasma Dynamics
Commissariat à l'Énergie Atomique et aux Énergies Alternatives
2013-2024
CEA Cadarache
2013-2024
Institut de Recherche sur la Fusion par Confinement Magnétique
2021-2023
Max Planck Institute for Plasma Physics
2023
Instituto Superior Técnico
2015-2020
Royal Military Academy
2020
Culham Science Centre
2013-2019
University of Lisbon
2016-2019
Instituto de Engenharia de Sistemas e Computadores Investigação e Desenvolvimento
2016
Culham Centre for Fusion Energy
2015
The impact of carbon and beryllium/tungsten as plasma-facing components on plasma radiation, divertor power particle fluxes, neutral conditions in the divertors has been assessed JET both experimentally by edge fluid code simulations for plasmas low-confinement mode. In high-recycling studies show a 30% reduction total radiation scrape-off (SOL) layer when replacing (JET-C) with beryllium main chamber tungsten (JET-ILW). Correspondingly, at low-field side (LFS) plate two-fold increase...
The erosion of tungsten (W), induced by the bombardment plasma and impurity particles, determines lifetime plasma-facing components as well impacting on performance influx W into confined region. screening divertor transport in largely content core, but source strength itself has a vital impact this process. JET tokamak experiment provides access to large set erosion-determining parameters permits detailed description closest ITER one: (i) effective sputtering yields fluxes function energy...
Burning plasmas with 500 MW of fusion power on ITER will rely partially detached divertor operation to keep target heat loads at manageable levels. Such regimes be maintained by a real-time control system using the seeding radiative impurities like nitrogen (N), neon or argon as actuator and one more diagnostic signals sensors. Recently, detachment has been successfully achieved in Type I ELMy H-mode JET-ITER-like wall discharges saturation current (Isat) measurements from Langmuir probes...
The low temperature boundary layer plasma (scrape-off or SOL) between the hot core and surrounding vessel determines level of power loading, erosion implantation material surfaces, thus viability tokamak-based fusion as an energy source. This study explores mechanisms affecting formation flattened density profiles, so-called 'density shoulders', in low-field side (LFS) SOL, which modify ion neutral fluxes to surfaces—and subsequent erosion. There is evidence against local enhancement...
Abstract Using a recently installed impurity powder dropper (IPD), boron (<150 μ m) was injected into lower single null (LSN) L-mode discharges in WEST. IPDs possibly enable real-time wall conditioning of the plasma-facing components and may help to facilitate H-mode access full-tungsten environment The this experiment featured I p = 0.5 MA, B T 3.7 T, q 95 4.3, t pulse 12–30 s, n e,0 ∼ 4 × 10 19 m −2 , P LHCD 4.5 MW. Estimates deuterium particle fluxes, derived from combination visible...
The ITER baseline scenario, with 500 MW of DT fusion power and Q = 10, will rely on a Type I ELMy H-mode, ΔW 0.7 MJ mitigated edge localized modes (ELMs). Tungsten (W) is the material now decided for divertor plasma-facing components from start plasma operations. W atoms sputtered targets during ELMs are expected to be dominant source under partially detached conditions required safe operation. impurity concentration in core can dramatically degrade its performance lead potentially damaging...
Abstract A parametrised spectral line profile model is formulated to investigate the diagnostic scope for recovering plasma parameters from hydrogenic Balmer and Paschen series spectroscopy in context of JET-ILW divertor plasmas. The separate treatment Zeeman Stark contributions tested against PPP-B code which accounts their combined influence on shape. proposed simplified does not fully reproduce Stark–Zeeman features α β transitions, but good agreement observed width wing profiles,...
The ITER baseline scenario, with 500 MW of DT fusion power and Q = 10, will rely on a Type I ELMy H-mode be achieved tungsten (W) divertor. W atoms sputtered from divertor targets during mitigated ELMs are expected to the dominant source in ITER. impurity concentration plasma core can dramatically degrade its performance lead potentially damaging disruptions. Understanding physics target due sputtering inter-ELMs is important helped by experimental measurements improved precision. It has...
Abstract This contribution aims to improve existing scalings of the L-mode power decay length <?CDATA ${\lambda_{q}^{{\rm omp}}}$?> <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" overflow="scroll"> <mml:mrow> <mml:msubsup> <mml:mi>λ</mml:mi> <mml:mi>q</mml:mi> </mml:mrow> <mml:mi mathvariant="normal">o</mml:mi> mathvariant="normal">m</mml:mi> mathvariant="normal">p</mml:mi> </mml:msubsup> </mml:math> , especially for plasma configurations with strike points at ITER-relevant...
Abstract As part of ITPA-Integrated Operational Scenario activities, this contribution reviews recent experimental characterizations radio-frequency (RF)-induced scrape-off layer (SOL) modifications various tokamaks worldwide and the LArge Plasma Device at UCLA. The phenomenology, as observed using a large variety measurement techniques, is consistent with expectations from RF sheath rectification. Emphasis then put on complex three-dimensional (3D) spatial patterns RF–SOL interaction, in...
Abstract The consequences of tungsten (W) melting on divertor lifetime and plasma operation are high priority issues for ITER. Sustained controlled W-melting experiment has been achieved the first time in WEST a poloidal sharp leading edge an actively cooled ITER-like facing unit (PFU). A series dedicated power steady state discharges were performed to reach point tungsten. was exposed parallel heat flux about 100 MW.m −2 up 5 s providing melt phase 2 without noticeable impact (radiated...
The EDGE2D-EIRENE code is applied for simulation of divertor detachment during matched density ramp experiments in high triangularity, L-mode plasmas both JET-Carbon (JET-C) and JET-ITER-Like Wall (JET-ILW). runs without drifts includes either C or Be as impurity, but not W, assuming that the W targets have been coated with via main chamber migration. simulations reproduce reasonably well observed particle flux raised JET-C JET-ILW can better match experimental in-out target power asymmetry...
The dynamics and stability of divertor detachment in N 2 seeded, type-I, ELMy H-mode plasmas with dominant NBI heating the JET ITER-like wall device is studied by means an integrated analysis diagnostic data from several systems, classifying relative to ELM times.It thereby possible study response evolution control parameters (SOL input power, upstream density impurity fraction) prevailing during inter-ELM periods effect ELMs on detached divertor.A relatively comprehensive overview achieved,...
Experiments in the JET tokamak equipped with ITER-like wall (ILW) revealed that inner and outer target plate at location of strike points represent after one year operation intact tungsten (W) surfaces without any beryllium (Be) surface coverage. The dynamics near-surface retention, implantation, desorption recycling deuterium (D) divertor plasma discharges are determined by W plates. As plasma-facing components (PFCs) not actively cooled, temperature (Tsurface) is increasing exposure,...
Gross and net erosion of tungsten (W) other plasma-facing materials in the divertor region have been investigated deuterium (D) helium (He) plasmas during dedicated experiments L- H-mode on ASDEX Upgrade after full-length experimental campaigns WEST tokamak. Net was determined via post-exposure analyses plasma-exposed samples full-size wall components, we conclude that same approach is applicable to gross if marker structures with sub-millimeter dimensions are used eliminate contribution...
Radiative divertor operation in JET high confinement mode plasmas with the ITER-like wall has been experimentally investigated and simulated EDGE2D-EIRENE horizontal vertical low field side (LFS) configurations. The simulations show that LFS heat fluxes are reduced N2-injection similar fashion both configurations, qualitatively consistent experimental observations. no substantial difference between two configurations reduction of peak flux as a function radiation, nitrogen concentration, or...
The JET ITER-Like Wall experiment, with its all-metal plasma-facing components, provides a unique environment for plasma and plasma-wall interaction studies. These studies are of great importance in understanding the underlying phenomena taking place during operation future fusion reactor. Present work summarizes reports fuel retention divertor resulting from two first experimental campaigns Wall. deposition pattern after second campaign shows same trend as was observed campaign: highest...
The WEST experiment is currently operating with tungsten plasma-facing components and testing ITER-like divertor monoblocks. In order to support experiments interpretation, numerical analyses were carried out. Starting from experimental data, realistic background plasma conditions reproduced through SolEdge-EIRENE used as input for ERO2.0 simulations investigate migration. Tungsten contamination due the different was modelled under conditions, highlighting a non-negligible contribution of...
Abstract Achieving a successful plasma current ramp-up in full tungsten tokamak can be challenging due to the large core radiation (and resulting low temperature) that faced with this heavy metallic impurity if its relative concentration is too high. Nitrogen injection during of WEST discharges greatly improves temperature and magnetohydrodynamic (MHD) stability. Experimental measurements integrated simulations RAPTOR code, complemented QuaLiKiz neural network for computing turbulent...
In this paper we discuss results from the study of energy balance in JET based on calculated heating energies, radiated bolometry and tile calorimetry. Recent data enables us to be more confident numbers used exclude certain possibilities but overall imbalance which typically amounts 25% total input remains unexplained. This shows that caution is required interpreting fractional powers are commonly measure effectiveness impurity seeded scenarios at reducing divertor heat load.
The design and operation of future fusion devices relying on H-mode plasmas requires reliable modelling edge-localized modes (ELMs) for precise prediction divertor target conditions. An extensive experimental validation simple analytical predictions the time evolution plasma loads during ELMs has been carried out here in more than 70 JET-ITER-like wall experiments with a wide range Comparisons these diagnostic measurements ion flux density, power impact energy electron temperature are...
This paper presents the use of a novel modelling technique based around intermittent transport due to filament motion, interpret experimental profile and fluctuation data in scrape-off layer (SOL) JET during onset evolution density shoulder.A baseline case is established, prior shoulder formation, stochastic model shown be capable simultaneously matching time averaged measurement as well PDF shape autocorrelation function from ion-saturation current series at outer wall.Aspects are then...
Abstract Experimental data, together with interpretive modeling tools, are examined to study trends in the tungsten (W) source all-W environment of WEST tokamak, both from divertor and main chamber. In particular, a poloidal limiter protecting an ion cyclotron resonance heating (ICRH) antenna is used as proxy for chamber sourcing. The key carried out by stepping up lower hybrid current drive (LHCD) power, only auxiliary power source. Limiter W sources exhibit qualitatively similar...