R. Dejarnac
A5001184279- Magnetic confinement fusion research
- Fusion materials and technologies
- Plasma Diagnostics and Applications
- Superconducting Materials and Applications
- Ionosphere and magnetosphere dynamics
- Nuclear reactor physics and engineering
- Laser-Plasma Interactions and Diagnostics
- Nuclear Materials and Properties
- Particle accelerators and beam dynamics
- Nuclear Physics and Applications
- Atomic and Subatomic Physics Research
- Laser-induced spectroscopy and plasma
- Metal and Thin Film Mechanics
- Magnetic Field Sensors Techniques
- Metallurgical Processes and Thermodynamics
- Nuclear and radioactivity studies
- Ion-surface interactions and analysis
- Solar and Space Plasma Dynamics
- Seismic Imaging and Inversion Techniques
- Geophysics and Gravity Measurements
- High-Energy Particle Collisions Research
- Electron and X-Ray Spectroscopy Techniques
- Intermetallics and Advanced Alloy Properties
- High-Temperature Coating Behaviors
- Dust and Plasma Wave Phenomena
Czech Academy of Sciences, Institute of Plasma Physics
2016-2025
Institut de Recherche sur la Fusion par Confinement Magnétique
2024
CEA Cadarache
2003-2024
Czech Academy of Sciences
2009-2021
Charles University
2010-2021
Royal Military Academy
2020
Commissariat à l'Énergie Atomique et aux Énergies Alternatives
2003-2017
Max Planck Institute for Plasma Physics
2017
Culham Science Centre
2014
The heating of tungsten monoblocks at the ITER divertor vertical targets is calculated using heat flux predicted by three-dimensional ion orbit modelling. are beveled to a depth 0.5 mm in toroidal direction provide magnetic shadowing poloidal leading edges within range specified assembly tolerances, but this increases field incidence angle resulting reduction wetted fraction and concentration local unshadowed surfaces. This shaping solution successfully protects from inter-ELM loads, expense...
The key remaining physics design issue for the ITER tungsten (W) divertor is question of monoblock (MB) front surface shaping in high heat flux target areas actively cooled targets. Engineering tolerance specifications impose a challenging maximum radial step between toroidally adjacent MBs 0.3 mm. Assuming optical projection parallel loads, magnetic shadowing these edges required if quasi-steady state melting to be avoided under certain conditions during burning plasma operation and...
The provision of a particle and power exhaust solution which is compatible with first-wall components edge-plasma conditions key area present-day fusion research mandatory for successful operation ITER DEMO. work package plasma-facing (WP PFC) within the European programme complements laboratory experiments, i.e. in linear plasma devices, electron ion beam loading facilities, studies performed toroidally confined magnetic such as JET, ASDEX Upgrade, WEST etc. connection both groups done via...
The original goals of the JET ITER-like wall included study impact an all W divertor on plasma operation (Coenen et al 2013 Nucl. Fusion 53 073043) and fuel retention (Brezinsek 083023). ITER has recently decided to install a full-tungsten (W) from start operations. One key inputs required in support this decision was possibility melting melt splashing during transients. Damage type can lead modifications surface topology which could higher disruption frequency or compromise subsequent...
Abstract WEST is an MA class superconducting, actively cooled, full tungsten (W) tokamak, designed to operate in long pulses up 1000 s. In support of ITER operation and DEMO conceptual activities, key missions are: (i) qualification high heat flux plasma-facing components integrating both technological physics aspects relevant particle exhaust conditions, particularly for the monoblocks foreseen divertor; (ii) integrated steady-state at confinement, with a focus on power issues. During phase...
This paper summarizes the status of COMPASS tokamak, its comprehensive diagnostic equipment and plasma scenarios as a baseline for future studies. The former COMPASS-D tokamak was in operation at UKAEA Culham, UK 1992–2002. Later, device transferred to Institute Plasma Physics Academy Sciences Czech Republic (IPP AS CR), where it installed during 2006–2011. Since 2012 has been full with Type-I Type-III ELMy H-modes base scenario. enables together ITER-like shape flexible NBI heating system...
Advanced Langmuir probe techniques for evaluating the plasma potential and electron-energy distribution function (EEDF) in magnetized are reviewed. It is shown that when magnetic field applied very weak electrons reach without collisions sheath second-derivative Druyvesteyn formula can be used EEDF evaluation. At low values of field, an extended yields reliable results, while at higher first-derivative technique applicable precise evaluation EEDF. There interval intermediate both...
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,...
The inboard limiters for ITER were initially designed on the assumption that parallel heat flux density in scrape-off layer (SOL) could be approximated by a single exponential with decay length λq. This was found not to adequate 2012, when infra-red (IR) thermography measurements inner column during JET limiter discharges clearly revealed presence of narrow channel adjacent last closed surface. near-SOL occurs λq ∼ few mm, much shorter than main SOL λq, and can raise at apex factor up ∼4...
The first results of particle-in-cell simulations the electrostatic sheath and magnetic pre-sheath thermionically emitting planar tungsten surfaces in fusion plasmas are presented. Plasma conditions during edge localized modes (ELMs) inter-ELM periods have been considered for various inclinations field selected surface temperatures. All runs performed under two assumptions potential drop; fixed or floating. primary focus lies on evaluation escaping thermionic current quantification...
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...
Abstract This paper summarizes the emissivity measurements performed on plasma-facing units (PFU) of WEST lower divertor during first phase running with a mix actively cooled ITER-like PFUs made bulk tungsten (W) and inertially graphite coating tungsten. In situ assessments laboratory after removing W-coated ITER-grade from device are shown. The exhibit complex pattern strong variation as function space time mainly explained magnetic equilibrium (strike point location) well plasma...
A new system of probes was recently installed in the divertor tokamak COMPASS order to investigate ELM energy density with high spatial and temporal resolution. The consists two arrays rooftop-shaped Langmuir (LPs) used measure floating potential or ion saturation current one array Ball-pen (BPPs) plasma a resolution ~3.5 mm. combination BPPs LPs yields electron temperature microsecond We report on design probe first results profile measurements ELMy H-mode L-mode. also present comparative...
The high heat flux areas on the vertical divertor targets in ITER tokamak will consist of cuboid tungsten monoblocks bonded to copper cooling tubes. Three-dimensional ion orbit modelling is used calculate heating during ELMs at inner target, where highest surface energy densities are expected. presence thin gaps between results exposed edges onto which can be focused. ELM ions focused by their gyromotion magnetically shadowed, long toroidal monoblocks. risk monoblock edge melting greater...
Power handling experiments with a special liquid metal divertor module based on the capillary porous system technology were performed in tokamak COMPASS. The performance of two metals (Li and LiSn alloy) tested for first time under ELMy H-mode conditions. No damage mesh good exhaust capability observed both separate up to 12 MW/m2 deposited perpendicular, inter-ELM steady-state heat flux ELMs relative energy ~3% local peak fluence at ~15 kJ.m−2. droplets directly ejected from top surface...
WEST is a full W tokamak with an extensive set of diagnostics for heat load measurements especially in the lower divertor.It composed by infrared thermography, thermal measurement thermocouples and fiber Bragg grating embedded few mm below surface flush mounted Langmuir probes.A large database including different magnetic equilibrium input power investigated to compare pattern (location, amplitude peak flux decay length) on inner outer strike point regions : from first ohmic diverted plasma...
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...