D. Naydenkova
A5076593473- Magnetic confinement fusion research
- Fusion materials and technologies
- Ionosphere and magnetosphere dynamics
- Atomic and Subatomic Physics Research
- Superconducting Materials and Applications
- Laser-Plasma Interactions and Diagnostics
- Particle accelerators and beam dynamics
- Plasma Diagnostics and Applications
- Nuclear Materials and Properties
- Nuclear Physics and Applications
- Laser-induced spectroscopy and plasma
- Spacecraft and Cryogenic Technologies
- Astronomical Observations and Instrumentation
- Magneto-Optical Properties and Applications
- Magnetic Field Sensors Techniques
- Nuclear reactor physics and engineering
- Quantum and electron transport phenomena
- GNSS positioning and interference
- Solar and Space Plasma Dynamics
Czech Academy of Sciences, Institute of Plasma Physics
2009-2025
Czech Academy of Sciences
2021
Charles University
2009-2020
École nationale supérieure d'arts et métiers
2010
Institut Polytechnique de Paris
2010
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...
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 role of the COMPASS tokamak in research generation, confinement and losses runaway electron (RE) population is presented. Recently, two major groups experiments aimed at improved understanding control REs have been pursued. First, effects massive gas injection ( Ar/Ne particles) impurity seeding were studied systematically. observed phenomena include generation post-disruption RE beam current conversion from plasma to RE. Zero loop voltage was implemented order study decay simplified...
Partial detachment is the desired regime for baseline burning plasma scenario in ITER and next-step devices, as it allows to dissipate majority of energy carried by charged particles through scrape-off-layer (SOL) thus avoids localised heat flux deposition divertor region.The COMPASS tokamak equipped with an open has a relatively short connection length, both factors being unfavourable access detachment.As such, only approach naturally detached operation at very high lineaveraged densities...
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...
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...
This paper presents two scenarios used for generation of a runaway electron (RE) beam in the COMPASS tokamak with focus on decay phase and control beam. The first scenario consists massive gas injection argon into current ramp-up phase, leading to disruption accompanied by plateau generation. In second scenario, smaller amount is order isolate RE from high-temperature plasma. performances radial vertical position feedback were experimentally studied analysed. role energy stability seems be...
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 COMPASS tokamak at IPP Prague is a small-size device with an ITER-relevant plasma geometry and operating in both the Ohmic as well neutral beam assisted H-modes since 2012. A basic set of diagnostics installed beginning operation has been gradually broadened type diagnostics, extended number detectors collected channels improved by increased data acquisition speed. In recent years, significant progress diagnostic development motivated performance broadening its scientific programme (L-H...
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...
First systematic measurements of pedestal structure during Ohmic and NBI-assisted Type I ELMy H-modes were performed on the COMPASS tokamak in two dedicated experimental campaigns 2015 2016. By adjusting NBI heating a toroidal magnetic field, electron temperature was increased from 200 eV up to 300 eV, which allowed reaching collisionality < 1 at q95 ~3. has approached conditions for Identity experiment done JET & DIII-D, complementing range scanned . The pressure successfully reproduced by...
Abstract Substantial power dissipation in the edge plasma is required for safe operation of ITER and next-step fusion reactors, otherwise unmitigated heat fluxes at divertor plasma-facing components (PFCs) would easily exceed their material limits. Traditionally, such flux mitigation linked to regime detachment, which characterised by a significant pressure gradient between upstream downstream scrape-off layer (SOL). However, physics phenomena responsible loss are distinctly different,...
Typical situations, which can be met during the process of absolute calibration, are shown in case a visible light observation system for COMPASS tokamak. Technical issues and experimental limitations measurements connected with tokamak operation discussed.