A5018447368- Magnetic confinement fusion research
- Fusion materials and technologies
- Superconducting Materials and Applications
- Nuclear reactor physics and engineering
- Laser-Plasma Interactions and Diagnostics
- Particle accelerators and beam dynamics
- Nuclear Materials and Properties
- Plasma Diagnostics and Applications
- Ionosphere and magnetosphere dynamics
- Laser-induced spectroscopy and plasma
- Dust and Plasma Wave Phenomena
- Nuclear Physics and Applications
- Metal and Thin Film Mechanics
- Cold Fusion and Nuclear Reactions
- Atomic and Molecular Physics
- Computational Fluid Dynamics and Aerodynamics
- Graphite, nuclear technology, radiation studies
- Gas Dynamics and Kinetic Theory
- Marine and environmental studies
- Quantum chaos and dynamical systems
- Physics of Superconductivity and Magnetism
- High-Energy Particle Collisions Research
- Nuclear Engineering Thermal-Hydraulics
Kurchatov Institute
2015-2024
Moscow Engineering Physics Institute
2015-2022
ITER
2008-2020
Plasma (Russia)
2020
Royal Military Academy
2020
Japan External Trade Organization
2014
Forschungszentrum Jülich
2010-2013
Max Planck Institute for Plasma Physics
1997-2009
Max Planck Society
1997-2009
Centre National de la Recherche Scientifique
2009
Progress, since the ITER Physics Basis publication (ITER Editors et al 1999 Nucl. Fusion 39 2137–2664), in understanding processes that will determine properties of plasma edge and its interaction with material elements is described. Experimental areas where significant progress has taken place are energy transport scrape-off layer (SOL) particular anomalous scaling, particle SOL plays a major role diverted plasmas main-chamber elements, localized mode (ELM) deposition on mechanism for ELM...
A multi-machine database for the H-mode scrape-off layer power fall-off length, λq in JET, DIII-D, ASDEX Upgrade, C-Mod, NSTX and MAST has been assembled under auspices of International Tokamak Physics Activity. Regression inside finds that most important scaling parameter is poloidal magnetic field (or equivalently plasma current), with decreasing linearly increasing Bpol. For conventional aspect ratio tokamaks, regression , yielding λq,ITER ≅ 1 mm baseline inductive burning scenario at Ip...
• Reviews the fundamental physics aspects of first ITER W divertor and defines required operational lifetime within Staged Approach. Uses SOLPS simulation database to establish target peak heat flux neutral pressure burning plasma operating domain. Assesses consequences narrow SOL channels, fluid drifts, component shaping 3D magnetic fields for ELM control. recrystallization define an budget shows that fluxes ∼50% higher than previously assumed may be acceptable. Shows Ne N should equally...
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...
Regime with the plasma detached from divertor targets (detached regime) is a natural continuation of high recycling conditions to higher density and stronger impurity radiation loss. Both theoretical considerations experimental data show clearly that increase loss volumetric recombination causes rollover flux target when increases, which manifestation detachment. Plasma-neutral friction (neutral viscosity effects), although important for sustainment density/pressure upstream providing...
The basic physics of the processes playing most important role in divertor plasma detachment is reviewed. models used two-dimensional edge transport codes that are widely to address different issues and simulate experimental data, as well numerical schemes convergence issues, described. leading ultimate detachment, transition stability detached regime, impact magnetic configuration geometry on considered. A consistent, integral physical picture a tokamak developed.
Tokamak Energy Ltd, UK, is developing spherical tokamaks using high temperature superconductor magnets as a possible route to fusion power relatively small devices. We present an overview of the development programme including details enabling technologies, key modelling methods and results, remaining challenges on path compact fusion.
Using the new version of SOLPS plasma boundary code package, SOLPS-ITER, paper presents first ever simulations ITER burning baseline H-mode edge with drifts and currents activated.Neon (Ne) seeded discharges for divertor power dissipation are considered.The results scrape-off layer (SOL) parameters without compared, both SOLPS-ITER against earlier SOLPS-4.3modelling (which did not include a drift description) constituting bulk existing simulation database.Whereas effect on equatorial...
The issue of first wall and divertor target lifetime represents one the greatest challenges facing successful demonstration integrated tokamak burning plasma operation, even in case planned next step device, ITER, which will run at a relatively low duty cycle comparison to future fusion power plants. Material erosion by continuous or transient ion neutral impact, susbsequent transport released impurities through their deposition and/or eventual re-erosion constitute process migration. Its...
The paper describes the results of a physics analysis modified divertor cassette for ITER. issues addressed are impact on operational window, effect gas leaks through broader gaps between cassettes and radiation power loading different components cassettes. shows that new design ensuring more flexibility ITER operation remains acceptable within framework usual trade-off target helium removal efficiency. load side walls structures in inter-cassette is identified as constraint not previously...
Results of a detailed study the parameter space ITER divertor with B2-Eirene code are presented. Relations between plasma parameters at separatrix, interface core and edge plasma, parametrized to provide set boundary conditions for models. The reference geometry is compared straight target option, possibility controlling density by shifting equilibrium in explored.
Recent results of divertor modelling are summarized and some incorporated in the ITER-FEAT design, such as beneficial effect V-shaped targets, importance high gas conductivity between divertors, role deep core fuelling maintaining plasma density, discussed. Differences carbon seeded impurities pointed out, helium elastic scattering is demonstrated. A number considerations for reactor-like operation additional requirements imposed on code under these conditions
This paper shows experimental results from the TCV tokamak that indicate plasma-molecule interactions involving $D_2^+$ and possibly $D^-$ play an important role as sinks of energy (through hydrogenic radiation well dissociation) particles during divertor detachment if low target temperatures ($< 3$ eV) are achieved. Both molecular activated recombination (MAR) ion source reduction due to a power limitation effect shown be in reducing flux density ramp. In contrast, electron-ion (EIR) sink...
In September 2011, the ITER Organization (IO) proposed to begin operation with a full-tungsten (W) armoured divertor, objective of taking decision on final target material (carbon fibre composite or W) by end 2013. This period 2 years would enable development full-W divertor design compatible nuclear operations, investigation further several physics R&D aspects associated use W targets and completion technology qualification. Beginning brief overview reference heat load specifications which...
Molecule-Activated Recombination (MAR) effect is re-considered in view of divertor plasma conditions. A strong isotopic demonstrated. In deuterium plasmas, the reaction chain through D2+ formation, usually considered dominant and included 2D edge models, negligible. However, this case other branch, D−, neglected modelling, becomes relatively strong. The overall share MAR recycling stays within 20%. operational parameters such as peak power loading on targets or pressure limit for partial...
ITER will be the first magnetic confinement device with burning DT plasma and fusion power of about 0.5 GW. Parameters have been predicted using methodologies summarized in Physics Basis (1999 Nucl. Fusion 39 2175). During past few years, new results obtained that substantiate confidence achieving Q ⩾ 10 inductive H-mode operation. These include achievement a good near Greenwald density at high triangularity cross section; improvements theory-based projections for core plasma, even though...