A5026144134- Fusion materials and technologies
- Magnetic confinement fusion research
- Nuclear Materials and Properties
- Plasma Diagnostics and Applications
- Nuclear reactor physics and engineering
- Superconducting Materials and Applications
- Nuclear Physics and Applications
- Hydrogen Storage and Materials
- Laser-Plasma Interactions and Diagnostics
- Metal and Thin Film Mechanics
- Ion-surface interactions and analysis
- Particle accelerators and beam dynamics
- Advanced Materials Characterization Techniques
- Hydrogen embrittlement and corrosion behaviors in metals
- Ionosphere and magnetosphere dynamics
- Laser-induced spectroscopy and plasma
- Semiconductor materials and devices
- Cold Fusion and Nuclear Reactions
- Agriculture and Biological Studies
- Advanced Chemical Physics Studies
- Nuclear materials and radiation effects
- Food Industry and Aquatic Biology
- Surfactants and Colloidal Systems
- Muon and positron interactions and applications
- Graphite, nuclear technology, radiation studies
Forschungszentrum Jülich
2013-2024
Voronezh State University of Engineering Technologies
2019
Culham Science Centre
2014-2016
Ghent University
2009-2014
University of Lisbon
2013
Moscow Engineering Physics Institute
2007-2010
Institute of Engineering Physics
2007
Martin Luther University Halle-Wittenberg
2004
Abstract After the second Deuterium–Tritium Campaign (DTE2) in JET tokamak with ITER-Like Wall (ILW) and full tritium campaigns that preceded followed after DTE2, a sequence of fuel recovery methods was applied to promote removal from wall components. The started several days baking main chamber walls at 240 °C 320 °C. Subsequently, superimposed Ion-Cyclotron Conditioning (ICWC) Glow Discharge (GDC) cleaning cycles deuterium. Diverted plasma operation deuterium different strike point...
Abstract JET returned to deuterium-tritium operations in 2023 (DTE3 campaign), approximately two years after DTE2. DTE3 was designed as an extension of JET's 2022-2023 deuterium campaigns, which focused on developing scenarios for ITER and DEMO, integrating in-depth physics understanding control schemes. These were evaluated with mixed D-T fuel, using the only remaining tritium-capable tokamak until its closure 2023. A core-edge-SOL integrated H-mode scenario developed tested D-T, showing...
Mater.415 S165-S169) carried out with the LIM code of ITER first wall (FW) on beryllium (Be) erosion, re-deposition and tritium retention by co-deposition under steady-state burning plasma conditions have shown that, depending input parameter assumptions sputtering yields, erosion lifetime fuel some parts FW can be a serious concern.The importance issue is such that benchmark this previous work sought has been provided ERO (Pitts et al 2011 J. Nucl.Mater.415 S957-S964) simulations described...
DEMO is the name for first stage prototype fusion reactor considered to be next step after ITER. For realization of energy especially materials questions pose a significant challenge already today. Advanced solution are under discussion in order allow operation conditions [1] and development used devices. Apart from issues related material properties such as strength, ductility, resistance against melting cracking one major tackled interaction with plasma. tungsten (W) discussed below do not...
Since 2011 the JET tokamak has been operated with a metal ITER-like wall (JET-ILW) including castellated beryllium limiters and lamellae-type bulk tungsten tiles in divertor.This allowed for large scale test of plasma-facing components (PFC).Procedures sectioning into single blocks castellation have developed.This facilitated morphology studies surfaces inside grooves after experimental campaigns 2011-2012 2013-2014.The deposition 0.4-0.5 mm wide is 'shallow'.It reaches 1-2 12 deep...
Abstract A sequence of fuel recovery methods was tested in JET, equipped with the ITER-like beryllium main chamber wall and tungsten divertor, to reduce plasma deuterium concentration less than 1% preparation for operation tritium. This also a key activity regard refining clean-up strategy be implemented at end 2nd DT campaign JET (DTE2) assess tools that are envisaged mitigate tritium inventory build-up ITER. The began 4 days baking 320 °C, followed by further which Ion Cyclotron Wall...
The pre-fusion power operation (PFPO) phase of ITER, as described in the ITER research plan with Staged Approach2, includes both hydrogen (H) and helium (He) plasma operations. In preparation for PFPO, WEST JET ran He campaigns to study plasma-wall interactions a tungsten environment. included back-and-forth transition between H or deuterium (D) allowing assessment achievable content well accessible wall reservoirs respective species. changeovers tokamak pulses fixed divertor configuration....
Abstract The deposition/erosion on optical diagnostic components—mirrors—is a critical issue in reactor class devices with long-pulsed high fluence plasma operation. paper presents results of the three-dimensional Monte–Carlo code ERO2.0 for two aperture and first mirror geometries to be deployed ITER, along separate simulation study that aims replicate from an experimental first-mirror carried out JET. Promisingly, very little impurity deposition mirrors anticipated durations is found...
Abstract For its initial operational phase, ITER has until recently considered using non-nuclear hydrogen (H) or helium (He) plasmas to keep nuclear activation at low levels. To this end, the Tokamak Exploitation Task Force of EUROfusion Consortium carried out dedicated experimental campaigns in He on ASDEX Upgrade (AUG) and JET tokamaks 2022, with particular emphasis put ELMy H-mode operation plasma-wall interaction processes as well comparison H deuterium (D) plasmas. Both pure mixed +...
The Monte-Carlo neutral transport code 3D-GAPS is described. models impurity and deposition in remote areas, such as gaps between cells of castellated plasma-facing surfaces. A step-by-step investigation the interplay different processes that may influence inside gaps, namely particle reflection, elastic collisions, sources, chemical erosion plasma penetration into presented. Examples modelling results application to TEXTOR experiment with a test limiter are provided. It shown only...
This paper presents particle-in-cell simulations of the plasma behaviour in vicinity gaps castellated plasma-facing components (PFCs). The point interest was test limiter TEXTOR tokamak, a PFC designed for studies plasma–wall interactions, particular, related to impurity transport and fuel retention. Simulations were performed various conditions surface, where can be either shaped or unshaped. It observed that depending on parameters particles inside gap potential- geometry-dominated...
Estimations of the ITER first wall (FW) lifetime, previously made using three-dimensional Monte-Carlo ERO code (Borodin et al 2011 Phys. Scr. T145 014008), depend strongly on assumptions physical sputtering yield for beryllium (Be). It is importance to validate respective model and data at existing devices including JET ITER-like (ILW) as most ITER-relevant experiments. Applying same input in those used before ITER-predictions, simulations Be light intensity (using up date atomic from ADAS...
For the development of tritium monitoring system in ITER hydrogen isotope release by Laser-Induced Desorption (LID) from Be layers is studied to determine laser parameters for a high desorption efficiency while minimising dust production and surface modifications also pursued. 1 µm thickness with 25–30 at% D 3 × 1022 D/m2 comparable JET-ILW areal concentrations [1] have been produced High Power Impulse Magnetron Sputtering (HiPIMS) on grade W. Laser pulses 1, 5 10 ms duration heat layer...
Physical and chemical assisted physical sputtering were characterised by the Be I II line BeD band emission in observation chord measuring sightline integrated front of inner beryllium limiter at torus midplane. The 3D local transport plasma-surface interaction Monte-Carlo modelling (ERO code [18]) is a key for interpretation observations vicinity shaped solid limiter. plasma parameter variation (density scan) regime has provided useful material simulation benchmark. improved background...
For the in situ application of LID (Laser-Induced Desorption) as a space-resolved tritium retention diagnostic ITER, desorption behaviour co-deposited deuterium (D) from beryllium (Be) layers is studied. In particular, efficiency dependence on laser pulse parameters investigated for durations 1–20 ms and absorbed energy densities up to 5 MJ m−2. these parameter scans homogenous Be/D were produced by High Power Impulse Magnetron Sputtering, with 10 μm thickness 1.6 at% D. Almost 99% initial D...
In confined plasma magnetic fusion devices significant amounts of the hydrogen isotopes used for reaction can be stored in plasma-facing materials by implantation. The desorption this retained was seen to follow a tα law with α ≈ −0.7 tokamaks. For pulsed reactor outgassing define inter-pulse waiting time. This work presents new experimental data on dynamic ITER grade tungsten exposed under well-defined conditions PSI-2 pure and mixed D2 plasmas.