J. Gérardin
A5075838446- Magnetic confinement fusion research
- Fusion materials and technologies
- Superconducting Materials and Applications
- Nuclear reactor physics and engineering
- Nuclear Materials and Properties
- Radiative Heat Transfer Studies
- Nuclear Physics and Applications
- Laser-Plasma Interactions and Diagnostics
- Plasma Diagnostics and Applications
- Pickering emulsions and particle stabilization
- Graphite, nuclear technology, radiation studies
- Ionosphere and magnetosphere dynamics
- Thermal Radiation and Cooling Technologies
- Particle accelerators and beam dynamics
- Fire dynamics and safety research
- Nuclear and radioactivity studies
- Glass properties and applications
- Gas Dynamics and Kinetic Theory
- Luminescence Properties of Advanced Materials
- Combustion and flame dynamics
- Nuclear Engineering Thermal-Hydraulics
- Catalysis and Hydrodesulfurization Studies
- Infrared Target Detection Methodologies
- Catalytic Processes in Materials Science
- Nuclear materials and radiation effects
CEA Cadarache
2018-2024
Commissariat à l'Énergie Atomique et aux Énergies Alternatives
2018-2023
Czech Academy of Sciences, Institute of Plasma Physics
2020-2023
Institut de Recherche sur la Fusion par Confinement Magnétique
2019-2023
Institute of Nuclear Physics, Polish Academy of Sciences
2021
Centre National de la Recherche Scientifique
2011-2020
Institut Universitaire des Systèmes Thermiques Industriels
2016-2020
Aix-Marseille Université
2020
Laboratoire de Génie des Procédés Catalytiques
2016-2017
Laboratoire Énergies et Mécanique Théorique et Appliquée
2010-2016
COMPASS Upgrade is a new medium size, high magnetic field tokamak (R = 0.9 m, Bt 5 T, Ip 2 MA) currently under design in the Czech Republic. It will provide unique capabilities for addressing some of key challenges plasma exhaust physics, advanced confinement modes and configurations as well testing facing materials liquid metal divertor concepts. This paper contains an overview preliminary engineering main systems (vacuum vessel, central solenoid poloidal coils, toroidal support structure,...
This work presents an overview of the integrated strategy developed, as part DEMO Key Design Integration Issue 1 (KDII1), to protect EU-DEMO first wall (FW) from planned and unplanned plasma transients by employing discrete limiters. The present Breeding Blanket (BB) FW design, which aims at minimizing loss neutrons while travelling breeding zone, is able withstand steady state heat fluxes up ≈1-1.5 MW/m² [1], not sufficient guarantee its integrity for most plasma-FW direct contact....
In the present work, role of plasma facing components protection in driving EU-DEMO design will be reviewed, focusing on steady-state and, especially, transients. This work encompasses both first wall (FW) as well divertor. fact, while ITER divertor heat removal technology has been adopted, FW concept shown past years to inadequate for EU-DEMO. is due higher foreseen irradiation damage level, which requires structural materials (like Eurofer) able withstand more than 5 dpa neutron damage....
Abstract To achieve their goals, future thermonuclear reactors such as ITER and DEMO are expected to operate plasmas with a high magnetic field, triangularity confinement. address the corresponding challenges, concept of high-field ( <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" overflow="scroll"> <mml:mrow> <mml:msub> <mml:mi>B</mml:mi> <mml:mtext>T</mml:mtext> </mml:mrow> </mml:msub> <mml:mtext>⩽</mml:mtext> </mml:math> 5 T), high-current <mml:mi>I</mml:mi>...
Abstract The protection of ITER in-vessel components and the plasma-wall interaction studies will be based on a large network infrared (IR) cameras covering 70% tokamak. surface temperature measurement from IR images remains challenging due to presence metallic targets, with changes in thermo-radiative properties (emissivity) multiple reflections. paper provides an overview major progress improve interpretation image get more reliable synthetic diagnostics. presents latest development (1)...
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...
Infra-red (IR) thermography is a widely used tool in fusion devices to monitor and protect the plasma-facing component (PFC) from excessive heat loads. However, with use of all-metal walls devices, deriving surface temperature IR measurements has become more challenging. In this paper, an overview infra-red metallic tokamaks WEST ASDEX Upgrade (AUG) reported techniques carried out modeling experimental fields deal radiative fully reflective environment are presented. Experimental...
Assessing the performance of ITER design for tungsten (W) divertor Plasma Facing Units (PFUs) in a tokamak environment is high priority issue to ensure efficient plasma operation. This paper reviews most recent results derived from experiments and post-mortem analysis ITER-grade PFUs exposed WEST associated modelling, with focus on understanding heat loading damage evolution. Several shaping options, sharp or chamfered leading edge (LE), unshaped shaped blocks toroidal bevel as foreseen...
Twelve ITER-like plasma-facing units made of tungsten were exposed in the WEST tokamak divertor, with three (PFUs) significantly overexposed to plasma heat flux: one sharp-edged PFU (vertical misalignment h = 0.8 mm) and two chamfered PFUs (h =0.6 mm 0.3 mm, respectively). This paper describes first temperature analysis obtained a very high spatial resolution infrared camera (pixel size ∼ 0.1 on misaligned edges shows consistency parallel flux derived from these measurements. The is focused...
Two small liquid metal targets based on the capillary porous structure were exposed to divertor plasma of tokamak COMPASS. The first target was wetted by pure lithium and second one a lithium-tin alloy, both releasing mainly atoms (sputtering evaporation) when plasma. Due poorly conductive material steep surface inclination (implying surface-perpendicular heat flux 12–17 MW/m2) for 0.1–0.2 s, LiSn has reached 900 °C under ELMy H-mode. A model conduction is developed serves evaluate...
Following ELMy H-mode experiments with liquid metal divertor target on the COMPASS tokamak, we predict behavior of a similar Upgrade, where it will be exposed to surface heat fluxes even higher than those expected in future EU DEMO attached divertor.We simulate conduction, sputtering, evaporation, excitation and radiation lithium tin area.Measured high-resolution data from tokamak were rescaled towards Upgrade based many established scalings.Our simulation then yields amount released which...
The consequences of tungsten (W) damaging processes, such as cracking and melting, on divertor lifetime plasma operation are high priority issues for ITER. A sustained melting experiment was conducted in WEST using a 2 mm deep groove geometry the upstream mono-block (MB) to overexpose sharp leading edge (LE) downstream MB. W-cracking has been evidenced first time with very spatial resolution infrared camera before reached. These cracks develop when monoblock temperature is about 2600°C, thus...
Abstract COMPASS addressed several physical processes that may explain the behaviour of important phenomena. This paper presents results related to main fields research obtained in recent two years, including studies turbulence, L–H transition, plasma material interaction, runaway electron, and disruption physics: Tomographic reconstruction edge/SOL turbulence observed by a fast visible camera allowed visualize turbulent structures without perturbing plasma. Dependence power threshold on...