A5101705566- Magnetic confinement fusion research
- Fusion materials and technologies
- Superconducting Materials and Applications
- Particle accelerators and beam dynamics
- Laser-Plasma Interactions and Diagnostics
- Plasma Diagnostics and Applications
- Nuclear reactor physics and engineering
- Ionosphere and magnetosphere dynamics
- Physics of Superconductivity and Magnetism
- Nuclear Materials and Properties
- Quantum Computing Algorithms and Architecture
- Laser-induced spectroscopy and plasma
- Advanced ceramic materials synthesis
- Nuclear Physics and Applications
- Magnetic Properties of Alloys
- Spectroscopy and Chemometric Analyses
- Magnetic and transport properties of perovskites and related materials
- Microstructure and Mechanical Properties of Steels
- Synthetic Aperture Radar (SAR) Applications and Techniques
- Solar and Space Plasma Dynamics
- Metallurgy and Material Forming
- Metal and Thin Film Mechanics
- Atmospheric and Environmental Gas Dynamics
- Superconductivity in MgB2 and Alloys
- Soil Moisture and Remote Sensing
Kyushu University
2023
Tokyo University of Agriculture
2022
Tokyo University of Agriculture and Technology
2019-2021
Fusion (United States)
1979-2020
Fusion Academy
1979-2020
National Institutes for Quantum Science and Technology
2017-2020
National Agency for New Technologies, Energy and Sustainable Economic Development
2017
National Institute for Fusion Science
1989-2017
Japan Atomic Energy Agency
1990-2016
ITER
2007-2015
Progress in the area of MHD stability and disruptions, since publication 1999 ITER Physics Basis document (1999 Nucl. Fusion 39 2137–2664), is reviewed. Recent theoretical experimental research has made important advances both understanding control tokamak plasmas. Sawteeth are anticipated baseline ELMy H-mode scenario, but tools exist to avoid or them through localized current drive fast ion generation. Active other instabilities will most likely be also required ITER. Extrapolation from...
The 'Progress in the ITER Physics Basis' (PIPB) document is an update of 'ITER (IPB), which was published 1999 [1]. IPB provided methodologies for projecting performance burning plasmas, developed largely through coordinated experimental, modelling and theoretical activities carried out on today's large tokamaks (ITER R&D). In IPB, projections (1998 Design) were also presented. pointed some outstanding issues. These issues have been addressed by Participant Teams (the European Union, Japan,...
Analysis of Type I ELMs from ongoing experiments shows that ELM energy losses are correlated with the density and temperature pedestal plasma before crash. The loss normalized to is found correlate across collisionality (ν*ped), decreasing increasing ν*ped. Other parameters affect size, such as edge magnetic shear, etc, which influence volume affected by ELMs. particle influenced this weakly dependent on other parameters. In JET DIII-D, under some conditions, can be observed (`minimum'...
As part of the ITER Design Review and in response to issues identified by Science Technology Advisory Committee, physics requirements were reviewed as appropriate updated. The focus this paper will be on recent work affecting design with special emphasis topics near-term procurement arrangements. This describe results on: sensitivity studies, poloidal field coil requirements, vertical stability, effect toroidal ripple thermal confinement, material choice heat load for plasma-facing...
The impacts of plasma disruptions on ITER have been investigated in detail to confirm the robustness design machine potential consequential loads. loads include both electro-magnetic (EM) and heat in-vessel components vacuum vessel. Several representative disruption scenarios are specified based newly derived physics guidelines for shortest current quench time as well maximum product halo fraction toroidal peaking factor arising from ITER. Disruption simulations with DINA code EM load...
It is observed that the waves launched from a phased array antenna of four waveguides couple effectively with electrons under condition $\frac{{\ensuremath{\omega}}_{0}}{{\ensuremath{\omega}}_{1h}(0)}\ensuremath{\gtrsim}2.0$. This coupling generates rf-driven current, rather than heating bulk electrons, and current/rf-power ratio 110 A/kW was obtained rf power 125 kW radiated into plasma which included appreciable suprathermal electrons.
The ITER plasma control system has the same functional scope as systems in present tokamaks. These are operation scenario sequencing, basic (magnetic and kinetic), advanced (control of RWMs, NTMs, ELMs, error fields, etc) fast shutdown. This chapter considers only initiation control. describes progress achieved these areas tokamak experiments since Physics Basis (1999 Nucl. Fusion 39 2577) was written results assessment to provide done for design (15?MA machine) at a more detailed level than...
Physics knowledge (theory and experiment) in energetic particles relevant to design of a reactor scale tokamak is reviewed, projections for ITER are provided this Chapter the Basis. The review includes single particle effects such as classical alpha heating toroidal field ripple loss, well collective instabilities that might be generated plasmas by particles. overall conclusion fusion expected provide an efficient plasma ignition sustained burn next step device. major concern localized heat...
The authors' data indicate that the L-mode to H-mode transition in DIII-D tokamak is associated with sudden reduction anomalous, fluctuation-connected transport across outer midplane of plasma. In addition edge density and magnetic fluctuations observed at transition, radial electric field becomes more negative after transition. They have determined scaling power threshold various plasma parameters; roughly linear increase toroidal are particularly significant. Control ELM frequency duration...
ITER is planned to be the first fusion experimental reactor in world operating for research physics and engineering. The ten years of operation will devoted primarily issues at low neutron fluence following engineering testing higher fluence. can accommodate various plasma configurations modes, such as inductive high Q long pulse hybrid modes non-inductive steady state with large ranges current, density, beta power, heating current drive methods. This flexibility provide an advantage coping...
Tokamak discharges using the expanded boundary divertor in DIII-D device exhibit H-mode confinement. With neutral-beam power up to 6 MW, energy confinement remains comparable Ohmic value at a plasma current of 1 MA. Confinement is also independent density and toroidal field. increases with current, but exact functional dependence is, as yet, uncertain. These results show that H mode can be achieved reactor-compatible open configuration.
Abstract The International Symposium on Liquid Metals Applications for Fusion (ISLA) aims to assemble scientists and engineers engaged in research lithium liquid metal applications fusion devices, facilitating discussions recent advancements challenges an open forum support the development of viable reactors. 8th (ISLA-8) was organized by Institute Plasma Physics, Chinese Academy Sciences, form September 8-12, 2024, Hefei, China. symposium attended over 70 participants, marking one highest...
The helium ash exhaust function of a divertor has been experimentally demonstrated. Helium atoms accumulate in the region as electron density main plasma increases. With concentration \ensuremath{\sim} 1.6% plasma, neutral pressure at is high 1.0\ifmmode\times\else\texttimes\fi{}${10}^{\ensuremath{-}4}$ Torr. This experiment indicates possibility an $\ensuremath{\alpha}$-particle-heated diverted tokamak with use pumping ducts practical size.
Observation of the intensity recycling particle flux at main plasma edge for various limiter and divertor discharges indicates that gross energy confinement beam-heated is closely related to flux. In discharges, global time τE show many similarities: 1) linear Ip dependence < 600 kA, 2) no BT dependence, 3) deterioration against injection power. Improvement by increasing Ip, example, associated with high temperatures region accompanied reduced recycling. – Divertor low around better than...
The successful operation of a single-null poloidal divertor in Doublet-III has demonstrated several new advantages diverted tokamak addition to the suppression impurity influx as DIVA: 1) contamination and radiation loss main plasma been reduced by an open geometry, i.e. without chamber; 2) radiative cooling formation dense cold (ne≥5 × l013 cm−3, Te≤7 eV, torr) have observed. – Up 50% Ohmic input power is radiated region, thus front plate down eV. This remote greatly reduces heat load on...
Neutral beam heated DIII-D expanded boundary divertor discharges have exhibited ASDEX-like H-mode behaviour over a wide parameter range. The deuterium energy confinement of 120 ms remained near the Ohmic value for up to 6 MW neutral heating, where it was 2-2.5 times higher than L-mode at plasma current 1 MA. hydrogen and helium were similar substantially below time. decreased with increasing power only 30% better 5 MW. In an mixture ([H]/[H+D] ≅ 40%), time in between values obtained pure...
The use of actively convected liquid metals with j × B force is proposed to facilitate heat handling by the divertor, a challenging issue associated magnetic fusion experiments such as ITER. This will be aggravated even more for DEMO and power reactors because divertor load significantly higher yet copper would not allowed sink material. Instead, reduced activation ferritic/martensitic steel alloys conductivities substantially lower than that copper, used structural materials. present...
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...
The plasma profile and parallel flow in the scrape-off layer (SOL) were systematically measured using reciprocating Mach probes installed at outer midplane near divertor magnetic null (x point) JT-60U tokamak with a single divertor. For ion vertical drift due to toroidal field gradient (ion nablaB drift) directed towards divertor, SOL along lines away from ("flow reversal") was discovered far A quantitative evaluation of "Pfirsch-Schluter flow," wherein is naturally produced plasma,...
A decaborane-based boronization system has been installed in the JT-60U tokamak order to reduce influx of impurities during plasma discharges. Boronization performed under a glow discharge using helium-decaborane gas mixture. The properties boron films deposited through and effects on discharges were investigated. It was found that deposition layer with high purity achieved few other than hydrogen boronization, present toroidally nonuniform film. also resulted good performance similar...
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...