A5011776509- Magnetic confinement fusion research
- Fusion materials and technologies
- Particle accelerators and beam dynamics
- Ionosphere and magnetosphere dynamics
- Superconducting Materials and Applications
- Plasma Diagnostics and Applications
- Laser-Plasma Interactions and Diagnostics
- Nuclear reactor physics and engineering
- Nuclear Physics and Applications
- Solar and Space Plasma Dynamics
- Atomic and Subatomic Physics Research
- Nuclear Materials and Properties
- GNSS positioning and interference
- Quantum chaos and dynamical systems
- Magnetic Field Sensors Techniques
- Dust and Plasma Wave Phenomena
- Advanced Data Storage Technologies
- Diamond and Carbon-based Materials Research
- Meteorological Phenomena and Simulations
- Physics of Superconductivity and Magnetism
- Catalytic Processes in Materials Science
- Theoretical and Computational Physics
- Particle Detector Development and Performance
- Laser-induced spectroscopy and plasma
- Distributed systems and fault tolerance
Czech Academy of Sciences, Institute of Plasma Physics
2016-2025
Culham Science Centre
2021
Czech Technical University in Prague
2021
Charles University
2021
Czech Academy of Sciences
2006-2021
Moscow Engineering Physics Institute
2020
CEA Cadarache
2008
Prague Institute of Planning and Development
2004
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 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...
As in many of today's tokamaks, plasma start-up ITER will be performed limiter configuration on either the inner or outer midplane first wall (FW). The massive, beryllium armored FW panels are toroidally shaped to protect panel-to-panel misalignments, increasing deposited power flux density compared with a purely cylindrical surface. chosen shaping should thus optimized for given radial profile parallel heat flux, scrape-off layer (SOL) ensure optimal spreading. For plasmas limited this is...
Asymmetrical disruptions may occur during ITER operation and they be accompanied by large sideways forces rotation of the asymmetry. This is particular concern because resonance rotating asymmetry with natural frequencies vacuum vessel (and other in-vessel components) could lead to dynamic amplification forces. A significant fraction non-mitigated JET have toroidally asymmetric currents that flow partially inside plasma surrounding ('wall'). The toroidal asymmetries (otherwise known as...
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...
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...
Plasma-facing components receive enormous heat fluxes under steady state and especially during transient conditions that can even lead to tungsten (W) melting. Under these conditions, the unimpeded thermionic current density emitted from W surfaces exceed incident plasma densities by several orders of magnitude triggering a replacement which drives melt layer motion via force. However, in tokamaks, is suppressed space-charge effects prompt re-deposition due gyro-rotation. We present...
The first-derivative probe technique was applied to derive data for plasma parameters from the IV Langmuir characteristics measured in boundary region COMPASS tokamak and TJ-II stellarator. It is shown that vicinity of last closed flux surface (LCFS) electron energy distribution function (EEDF) bi-Maxwellian with low-temperature fraction predominating over higher temperature one, whereas far scrape-off layer (SOL) EEDF Maxwellian. In stellarator during NBI heated confined close LCFS while...
The ball-pen probe (BPP) technique is used successfully to make profile measurements of the electron temperature on ASDEX Upgrade (Axially Symmetric Divertor Experiment), COMPASS (COMPact ASSembly), and ISTTOK (Instituto Superior Tecnico TOKamak) tokamak. provided by a combination BPP potential (ΦBPP) floating (Vfl) Langmuir (LP), which compared with Thomson scattering diagnostic COMPASS. Excellent agreement between two diagnostics obtained for circular diverted plasmas different heating...
Remobilization is one of the most prominent unresolved fusion dust-relevant issues, strongly related to lifetime dust in plasma-wetted regions, survivability on hot plasma-facing surfaces and formation accumulation sites. A systematic cross-machine study has been initiated investigate remobilization tungsten micron-size from implementing a newly developed technique based controlled pre-adhesion by gas dynamics methods. It utilized number devices provided new insights under steady-state...
In fusion devices strong interest is deserved to the edge filament transport both related turbulent and ELMy structures, because they are believed provide important interaction with plasma facing components divertor plates. Among their features also electromagnetic (EM) properties expected play an role in particular for high beta regimes future devices. The presence of ELM inter-ELM filaments were detected during H-mode discharges COMPASS device Scrape-Off Layer, where a new probe head was...
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...
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>...