Electric Potential in Core Plasma
Development of the Fusion Core Plasma Electric Potential Diagnostics
Tech Area / Field
- FUS-MCS/Magnetic Confinement Systems/Fusion
- FUS-PLA/Plasma Physics/Fusion
- INS-MEA/Measuring Instruments/Instrumentation
3 Approved without Funding
Kurchatov Research Center, Russia, Moscow
- Thermonuclear Controlled Fusion International Projects Coordinating Centre, Russia, Moscow
- European Fusion Development Agreement, UK, Abingdon\nUniversity of Saskatchewan / Department of Physics and Engineering Physics, Canada, SK, Saskatoon\nEURATOM-Ciemat, Spain, Madrid\nMax-Plank-Institut für Plasmaphysik / IPP Greifswald, Germany, Greifswald\nUniversiteit Gent / Department of Applied Physics, Belgium, Gent\nInstituut voor Plasmafysica, The Netherlands, Neiuwegein\nNational Institute for Fusion Science, Japan, Gifu-ken\nOak Ridge National Laboratory / Fusion Energy Division, USA, TN, Oak Ridge
Project summaryGoal of the project. Development of the diagnostic for the electric fields and turbulence characteristics in fusion plasmas. The presented Project is a constituent part of joint work with Ukrainian group from UNSC “Kharkov Physical and Technical Institute” leaded by L.I. Krupnik (Application STCU #4439).
Status of researches. After many years of experiments on magnetic fusion devices, the humanity closely came to creation of the experimental fusion reactor tokamak ITER; the agreement about ITER creation was signed on the state level at 2006. However, plasma parameters in the ITER project were estimated using empirical scalings instead of physical laws, although these scalings do not allow ones to understand the physical processes, which determine the particles and energy transport in fusion plasmas.
Experiments have shown that the energy and particle transport may be pided into the classical (collisional) and anomalous (turbulent) parts. When the plasma parameters approach to fusion ones, the classical (neoclassical) transport ought to decrease due to collisionality decrease, however, predictions of turbulent transport are difficult, therefore its investigations become more important. Well-known that the turbulence is ubiquitous attribute of MHD systems; it is observed both in space and laboratory plasmas. Recent experiments clearly show that the heat transport in fusion plasmas is mainly determined by excitation of turbulent fluctuations of density, electric and magnetic fields. The interplay of transport with turbulence explicitly developed during transitions to improved confinement, when the external (Н-mode) and internal transport barriers (ITB) are formed and the turbulence is suppressed. Experiment of fusion devices: tokamaks and stellarators have shown that transitions to improved confinement are accompanied by change of radial electric field Er. The possible mechanism of the confinement improvement is the suppression of turbulence by the sheared rotation driven by Er. So, understanding of physical mechanisms, which determine both the transport, and role of Er in plasma confinement in tokamaks and stellarators, is one of main issues of modern plasma physics and controlled fusion.
Although the transport is mainly turbulent, the effects of self-organization, establishment of some optimal profiles and the turbulence stabilization were observed. During last years the mechanisms of the turbulence stabilization and self-stabilization, in particular, zonal flows (ZF) and their high-frequency branch, the geodesic acoustic modes (GAM) are intensively studied. In the fundamental review P. Diamond et al., “Zonal flows in plasma - a review”, Plasma Phys. Control. Fusion, 47 R35 (2005), the heavy ion beam probing (HIBP) is adopted as the first diagnostic for electric fields and ZF study.
The role of Project in progress of researches. Realization of the main goal of Project, measurement of electric field and parameters of turbulence allows us to study them quantitatively, and open the possibility of their control. Fulfillment of the project allows us to give recommendations, how to obtain regimes with improved plasma confinement in tokamaks and stellarators. In perspective, it improves efficiency and economics of the future reactors ITER and DEMO.
Expected results. Project is directed to creation of the valid physical picture of heat and particle transport with account of electric field and turbulent mechanisms. New experimental and calculated data for substantiation the theoretical models of neoclassical transport and electrostatic turbulence will be obtained.
We will measure the electric potential and investigate its link with the tokamak plasma confinement. This allows us to estimate the limits of validity of the theoretical models. As a result, we obtain new data for clarifying the role of electric field in the energy and particle transport of fusion plasmas.
We plan to upgrade the diagnostic complex of heavy ion probing on T-10, to expand its energy range, to develop the new tools for measurements: the precision multi-slit analyzer of ion energy and the positional-sensitive matrix detector for simultaneous correlational measurement of electric potential and density fluctuations, and turbulent flux in the plasma core. This gives us the opportunity to experimentally verify the paradigm of turbulence suppression by the sheared flow of poloidal plasma rotation.
Cooperation with Ukrainian colleagues noticeably alleviates fulfillment of Project. Their experience with high-voltage power supplies and with high-current ionic emitters and their drawings of the ion source and emitter will be used in our Project to increase the accelerating voltage up to 300 kV and to measure the plasma potential and its fluctuation. On the other hand, the drawings of multi-slits analyzer and methods of measurements, developed in our Project, will be used in the Ukrainian project of plasma probing by heavy ion beam in the Uragan -2M stellarator. Our program package will be used for calculations of trajectories in the complicated magnetic geometry of Uragan -2M.
Hence, joint realization of both Projects raises the diagnostics on the higher technical level, which is necessary for investigations in the modern fusion devices, such as TJ-II, LHD, JET and W7-X, and in perspective in ITER and DEMO reactors.
Obtained physical results will be important for the improved confinement regimes formation in operating fusion devices like stellarators TJ-II (Spain), WEGA (Germany) and LHD (Japan), European tokamak JET (UK), and in the near-term future, they will be applied to world largest constructed stellarator Wendelstein-7X (Germany) and international fusion reactor ITER (France).
Realization of ISTC goals. Fulfillment of the ISTC project allows to 10 “weapon” specialists to redirect their capabilities for peaceful activity in the field of fusion energy. Realization of the project supports fundamental and application-oriented researches, and development of devices for fusion energy. The Project will assist to deeper involvement of scientists and specialists from Kurchatov Institute into international fusion community. Project will help to solve the important technical problems of fusion reactor ITER, which was initiated and constructed with active participation of Russia.
Scope of Activity. For realization of the Project we ought to solve the following tasks:
To conduct experiments on T-10 tokamak for investigation of physical mechanisms of transport and the role of electric fields in neoclassical and turbulent transport in discharges with Ohmic and electron cyclotron (EC) heating at different plasma density and EC power. To investigate the evolution of plasma potential, electric fields and rotation velocities during formation of external (L-H transition) and internal transport barriers (ITB).
To measure the turbulent flux in the hot core and at the plasma edge. To study the external control of electric field through biasing, which allows us to reveal the causal links between modification of confinement and electric filed. To increase the beam energy up to 300 keV for measurement of the potential in the core plasma area with the sawtooth oscillations. To compare data obtained in the shearless stellarator TJ-II and in the T-10 tokamak that allows us to clarify the role of neoclassical mechanisms, magnetic shear and Er in the transport.
The Project consists of the preliminary phase, when the apparatus will be preliminary calculated, designed, manufactured and tested on the bench. During the main phase, the experimental results will be obtained in Т-10. During the final phase, the experimental T-10 data will be compared with data obtained on the collaborators machines and with the theoretical models of turbulence.
Role of collaborators. In frames of proposed Project, the regular exchange of obtained scientific information is supposed, as well as cross-check of electric potential measurements in Т-10, TJ-II, LHD and WEGA, joint symposiums and workshops (NIFS), mutual visits to Т-10, TJ-II and WEGA for joint experiments, publication of joint papers and reports.
Technical approach and methods. Main method of study will be the plasma probing by high-energy heavy ion beam. The following scientific and technical problems ought to be solved:
Modernization of high-voltage power supply for the acceleration of the heavy ions up to 300 keV. Development, manufacturing, calibration and study of performances of multi-slit energy analyzer. Development, manufacturing, calibration and study of the positional sensitive matrix beam detector (MCAD). Measurement of the plasma electric potential and its fluctuations in the core of T-10 plasma up to the center. Approbation of multi-channel analyzer in the realistic condition of T-10 experiment. Multichannel measurements of potential and convective turbulent flux in the plasma core in T-10.
The proposed method of the turbulent flux measurement is as follows. Local value of radial turbulent flux Гr, perpendicular to magnetic field B is Гr=<n~·V~r>, where n~ are density fluctuations, V~r is the radial component of fluctuation velocity, <> mean average of distribution. The flux velocity across confining magnetic field Vr is defined by drift in the crossed poloidal electric field Еp and toroidal magnetic field Вtor: V~r =[E~p * Btor]. We propose to define E~p by simultaneous measurements of potential fluctuations in two points of plasma by multi-slit analyzer. The studied points ought to be placed on the same magnetic surface with different poloidal angles. Thus the differential signal of potentials will be proportional to E~p. To determine n~, we will measure the fluctuations of the ion beam current onto the analyzer. Simultaneous detecting of the density and potential fluctuations allows us to find their mutual correlation, which is necessary to calculate Гr.
The project will be performed in the Kurchatov Institute and in the Fusion Centre (Moscow, Russia) using the diagnostic developed by authors of the project in cooperation with Ukrainian colleagues (STCU#4439), and the T-10 diagnostics and data acquisition complex, the high-voltage test bench. Authors of the project are qualified specialists. They are well known in the international fusion society. They published many experimental and theoretical papers in the field of physics, diagnostics and simulation of fusion plasmas.
Developed drawings of multi-slits analyzer and methods of measurements will be used in Ukrainian project of diagnostics for plasma probing by the heavy ion beam in the stellarator Uragan -2M. The developed software package will be used in the Ukrainian project for calculation of trajectories in the complex magnetic geometry of Uragan. In its turn, the methods of high-voltage power supply and high-current emitters development from Ukrainian project, as well as the drawings of power supply and emitters will be used in our Project. This allows us to increase the accelerating voltage up to 300 keV and to measure the plasma potential and its fluctuations. Thus, joint realization of both Projects elevates the technical level of HIBP diagnostic, which needed to study plasmas in large modern fusion devices, like TJ-II, LHD, JET and W7-X, and in the distant future in fusion reactors ITER and DEMO.
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