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Interactions of Tungsten (W), Molybdenum (Mo), and Beryllium (Be) vapors with a steady-state plasma were studied by the PISCES-B liner plasma experiments as well as Particle-In-Cell (PIC) simulations for the understanding of vapor-shielding phenomena. Effective cooling of the plasma by laser-generated Be vapor was observed in PISCES-B. On the other hand, no apparent cooling was observed for W and Mo vapors. The PIC simulation explains these experimental observations of the difference between low-Z and high-Z vapors. Decrease of electron temperature due to the vapor ejection was observed in case of a simulation of the Be vapor. As for the W vapor, it was found that the plasma cooling is localized only near the wall at a higher electron density plasma (∼1019m−3). On the other hand, the appreciable plasma cooling can be observed in a lower density plasma (∼1018m−3) for the W vapor. Keywords: Vapor shielding, Plasma-vapor interaction, Particle-in-cell simulation

Electron cyclotron current drive (ECCD) has been numerically simulated in magnetic islands caused by neo-classical tearing modes. The electron drift orbits are tracked with the Coulomb collisions and the quasi-linear diffusion by EC waves, which are simulated by the Monte-Carlo method. The EC resonance region is assumed to be located around the O-point and localized in the toroidal direction. The driven current is well confined in the helical flux tube which includes the EC resonance region. The driven current channel looks like a 'snake' in real space. As the results of the ECCD with 10 MW in a plasma with ne = 3 × 1019 m[?]3 and Te = 10 keV, the current drive efficiency is about 1.6 times higher than that of an axi-symmetric plasma with no magnetic island. The driven current profile tends to peak around the O-point with increasing EC wave power. [ABSTRACT FROM AUTHOR]

To predict the heat and particle controllability in the divertor of tokamak reactors and to optimize the divertor design, comprehensive simulations by integrated modeling allowing for various physical processes are indispensable. SOL/divertor codes have been developed in the Japan Atomic Energy Agency for the interpretation and the prediction of behaviour of SOL/divertor plasmas, neutrals and impurities. The code system consists of the two-dimensional fluid code SOLDOR, the neutral Monte-Carlo (MC) code NEUT2D and the impurity MC code IMPMC. Their integration code 'SONIC' is almost completed and examined to simulate self-consistently the SOL/divertor plasmas in JT-60U. In order to establish the physics modelling used in fluid simulations, the particle simulation code PARASOL has also been developed.Simulation studies using those codes have progressed with the analysis of JT-60U experiments and the divertor designing of JT-60SA (modification program of JT-60U). The X-point multifaceted asymmetric radiation from the edge in the JT-60U experiment is simulated. It is found that the deep penetration of chemically sputtered carbon at the dome causes the large radiation peaking near the X-point. The pumping capability of JT-60SA is evaluated through the simulation. A guideline to enhance the pumping efficiency is obtained in terms of the exhaust slot width and the strike point distance. Transient behaviour of SOL/divertor plasmas after an ELM crash is characterized by the PARASOL simulation; the fast-time-scale heat transport is affected by collisions while the slow-time-scale behaviour is affected by the recycling. [ABSTRACT FROM AUTHOR]

A fluid treatment of convective-transport threshold model of neoclassical tearing modes (NTMs) in tokamaks is developed. A Grad-type system of moment equations of the drift kinetic equation with a model perpendicular transport is derived. The essence of this moment equation system is to allow for the parallel heat flux on an equal footing with pressure and temperature, what goes beyond the scope of the Braginskii approach. The suggested moment equation system is applied for analyzing the bootstrap current drive of NTMs. As a result, a threshold model of these modes is derived, which coincides qualitatively with the convective-transport threshold model initially formulated by means of intuitive considerations. © 2003 American Institute of Physics. [ABSTRACT FROM AUTHOR]

The negative triangularity tokamak (NTT) is a unique reactor concept based on ‘power-handling-first’ philosophy with the heat exhaust problem as the leading concern. The present paper exposes a reactor concept using L-mode edge based on NTT configuration, providing merits of no (or very weak) edge-localized modes, larger particle flux and large major radius for power handling. It is shown that a reasonably compact (Rp from 9 m to 7 m) NTT reactor is possible by achieving higher confinement improvement (HH  =  1.5) and/or by utilizing reasonably higher magnetic field (Bmax  =  15.5 T). Current physics basis and critical issues on its scientific and technical feasibility are discussed. [ABSTRACT FROM AUTHOR]

Detached divertor is considered a solution for the heat control in magnetic-confinement fusion reactors. Numerical simulations using the comprehensive divertor codes based on the plasma fluid modeling are indispensable for the design of the detached divertor in future reactors. Since the agreement in the results between detached-divertor experiments and simulations has been rather fair but not satisfactory, further improvement of the modeling is required. The kinetic effect is one of key issues for improving the modeling. Complete kinetic behaviors are able to be simulated by the kinetic modeling. In this paper at first, major kinetic effects in edge plasma and detached divertor are listed. One of the most powerful kinetic models, particle-in-cell (PIC) model, is described in detail. Several results of PIC simulations of edge-plasma kinetic natures are presented. Future works on PIC modeling and simulation for the deeper understanding of edge plasma and detached divertor are discussed. [ABSTRACT FROM AUTHOR]

The paper discusses edge stability, beta limits and power handling issues for negative triangularity tokamaks. The edge magnetohydrodynamic stability is the most crucial item for power handling. For the case of negative triangularity the edge stability picture is quite different from that for conventional positive triangularity tokamaks: the second stability access is closed for localized Mercier/ballooning modes due to the absence of a magnetic well, and nearly internal kink modes set the pedestal height limit to be weakly sensitive to diamagnetic stabilization just above the margin of the localized mode Mercier criterion violation. While a negative triangularity tokamak is thought to have a low beta limit with its magnetic hill property, it is found that plasmas with reactor-relevant values of normalized beta βN > 3 can be stable to global kink modes without wall stabilization with appropriate core pressure profile optimization against localized mode stability, and also with increased magnetic shear in the outer half-radius. The beta limit is set by the n = 1 mode for the resulting flat pressure profile. The wall stabilization is very inefficient due to strong coupling between external and internal modes. The n > 1 modes are increasingly internal when approaching the localized mode limit, and set a lower beta in the case of the peaked pressure profile leading to a Mercier unstable core. With the theoretical predictions supported by experiments, a negative triangularity tokamak would become a prospective fusion energy system with other advantages including a larger separatrix wetted area, more flexible divertor configuration design, wider trapped particle-free scrape-off layer, lower background magnetic field for internal poloidal field coils, and larger pumping conductance from the divertor room. [ABSTRACT FROM AUTHOR]

The tokamak hybrid scenario, a combination of the inductive current drive and the non-inductive current drive, aims at high neutron fluence and a long pulse length for the ITER engineering test towards DEMO. It has been reported experimentally that the concentration of current density, J, peculiar to inductive discharges, was much smaller than that of a simple estimation. We propose a new model to explain the above observations: the central core plasma column can be non-axisymmetric due to the m/n = 1/1 MHD-mode deformation (m and n are poloidal and toroidal mode numbers, respectively). The neo-classical resistivity is enhanced, and the diagnostics are affected. An internal kink mode is a candidate of this instability. Compared with the JT-60U experiment, the reconstructed J profile agrees well with the calculated J profile including the enhanced neo-classical resistivity. [ABSTRACT FROM AUTHOR]

The dependence of the ion-temperature-gradient scale length on the hydrogen isotope mass was examined in conventional H-mode plasmas in JT-60U tokamak. While identical profiles for density and temperature were obtained for hydrogen and deuterium plasmas, the ion conductive heat flux necessary for hydrogen to sustain the same ion temperature profile was two times that required for deuterium, resulting in a clearly higher ion heat diffusivity for hydrogen at the same ion-temperature-gradient scale length. On the other hand, the ion-temperature-gradient scale length for deuterium is less than that for hydrogen at a given ion heat diffusivity.

Nondimensional parameter dependence of heat transport between edge localized modes (ELMs) is examined for H mode plasmas. The electron heat diffusivity between ELMs is reduced to the level of ion neoclassical transport in the plasma edge region which is affected by ELM burst. At lower edge collisionality, the heat flux assigned to the heat transport between ELMs is reduced and the ELM loss power is enhanced. During the inter-ELM phase, the energy confinement time becomes larger with decreasing the edge collisionality and poloidal Larmor radius.