Effect of toroidal field ripple on plasma rotation in JET

Dedicated experiments on TF ripple effects on the performance of tokamak plasmas have been carried out at JET. The TF ripple was found to have a profound effect on the plasma rotation. The central Mach number, M, defined as the ratio of the rotation velocity and the thermal velocity, was found to drop as a function of TF ripple amplitude (3) from an average value of M = 0.40-0.55 for operations at the standard JET ripple of 6 = 0.08% to M = 0.25-0.40 for 6 = 0.5% and M = 0.1-0.3 for delta = 1%. TF ripple effects should be considered when estimating the plasma rotation in ITER. With standard co-current injection of neutral beam injection (NBI), plasmas were found to rotate in the co-current direction. However, for higher TF ripple amplitudes (delta similar to 1%) an area of counter rotation developed at the edge of the plasma, while the core kept its co-rotation. The edge counter rotation was found to depend, besides on the TF ripple amplitude, on the edge temperature. The observed reduction of toroidal plasma rotation with increasing TF ripple could partly be explained by TF ripple induced losses of energetic ions, injected by NBI. However, the calculated torque due to these losses was insufficient to explain the observed counter rotation and its scaling with edge parameters. It is suggested that additional TF ripple induced losses of thermal ions contribute to this effect

Effect of the tangential NBI current drive on the stability of pressure and energetic particle driven MHD modes in LHD plasma

The aim of the present study is to analyze the stability of the pressure gradient driven modes (PM) and Alfvén eigenmodes (AE) in the large helical device (LHD) plasma if the rotational transform profile is modified by the current drive of the tangential neutral beam injectors (NBI). This study forms a basic search for optimized operation scenarios with reduced mode activity. The analysis is performed using the code FAR3d which solves the reduced MHD equations describing the linear evolution of the poloidal flux and the toroidal component of the vorticity in a full 3D system, coupled with equations for density and parallel velocity moments of the energetic particle (EP) species, including the effect of the acoustic modes. The Landau damping and resonant destabilization effects are added via the closure relation. On-axis and off-axis NBI current drive modifies the rotational transform which becomes strongly distorted as the intensity of the neutral beam current drive (NBCD) increases, leading to wider continuum gaps and modifying the magnetic shear. The simulations with on-axis NBI injection show that a counter (ctr-) NBCD in inward shifted and default configurations leads to a lower growth rate of the PM, although strong n  =  1 and 2 AEs can be destabilized. For the outward shifted configurations, a co-NBCD improves the AEs stability but the PM are further destabilized if the co-NBCD intensity is 30 kA T−1. If the NBI injection is off-axis, the plasma stability is not significantly improved due to the further destabilization of the AE and energetic particle modes (EPM) in the middle and outer plasma region.This work is supported in part by NIFS under contract NIFS07KLPH004

Transport Simulasion in a Burning Tokamak Plasma

A one-dimensional tokamak transport code (TASK/TR) has been developed to analyze the evolution of a burning plasma accompanied with fusion reaction. This code deals with the electrons, deuterons, tritons, thermalized α particles, fast α particles and beam ions, separately, in order to describe the dependence of the reaction rate on the ion mixture ratio. As an energy transport model, the drift wave turbulence mode is employed. The heating and current drive by the neutral beam injection as well as the pellet injection for fuelling are also included. This code is applied to a reactor-grade plasma aimed at in the ITER project. The cases of an ignited plasma and a current-driven plasma are examined. The required power for full current drive is estimated. The effect of pellet injection, both fuel and impurity ions, is also studied

Overview of the design of the ITER heating neutral beam injectors

The heating neutral beam injectors (HNBs) of ITER are designed to deliver 16.7MWof 1 MeVD0 or 0.87 MeVH0 to the ITER plasma for up to 3600 s. They will be the most powerful neutral beam\uf0a0(NB) injectors ever, delivering higher energy NBs to the plasma in a tokamak for longer than any previous systems have done. The design of the HNBs is based on the acceleration and neutralisation of negative ions as the efficiency of conversion of accelerated positive ions is so low at the required energy that a realistic design is not possible, whereas the neutralisation ofH 12 andD 12 remains acceptable ( 4856%). The design of a long pulse negative ion based injector is inherently more complicated than that of short pulse positive ion based injectors because: \u2022 negative ions are harder to create so that they can be extracted and accelerated from the ion source; \u2022 electrons can be co-extracted from the ion source along with the negative ions, and their acceleration must be minimised to maintain an acceptable overall accelerator efficiency; \u2022 negative ions are easily lost by collisions with the background gas in the accelerator; \u2022 electrons created in the extractor and accelerator can impinge on the extraction and acceleration grids, leading to high power loads on the grids; \u2022 positive ions are created in the accelerator by ionisation of the background gas by the accelerated negative ions and the positive ions are back-accelerated into the ion source creating a massive power load to the ion source; \u2022 electrons that are co-accelerated with the negative ions can exit the accelerator and deposit power on various downstream beamline components. The design of the ITER HNBs is further complicated because ITER is a nuclear installation which will generate very large fluxes of neutrons and gamma rays. Consequently all the injector components have to survive in that harsh environment. Additionally the beamline components and theNBcell, where the beams are housed, will be activated and all maintenance will have to be performed remotely. This paper describes the design of theHNBinjectors, but not the associated power supplies, cooling system, cryogenic system etc, or the high voltage bushingwhich separates the vacuum of the beamline fromthehighpressureSF6 of the high voltage (1MV) transmission line, through which the power, gas and coolingwater are supplied to the beam source. Also themagnetic field reduction system is not described

Power requirements for electron cyclotron current drive and ion cyclotron resonance heating for sawtooth control in ITER

13MW of electron cyclotron current drive (ECCD) power deposited inside the q = 1 surface is likely to reduce the sawtooth period in ITER baseline scenario below the level empirically predicted to trigger neo-classical tearing modes (NTMs). However, since the ECCD control scheme is solely predicated upon changing the local magnetic shear, it is prudent to plan to use a complementary scheme which directly decreases the potential energy of the kink mode in order to reduce the sawtooth period. In the event that the natural sawtooth period is longer than expected, due to enhanced alpha particle stabilisation for instance, this ancillary sawtooth control can be provided from > 10MW of ion cyclotron resonance heating (ICRH) power with a resonance just inside the q = 1 surface. Both ECCD and ICRH control schemes would benefit greatly from active feedback of the deposition with respect to the rational surface. If the q = 1 surface can be maintained closer to the magnetic axis, the efficacy of ECCD and ICRH schemes significantly increases, the negative effect on the fusion gain is reduced, and off-axis negative-ion neutral beam injection (NNBI) can also be considered for sawtooth control. Consequently, schemes to reduce the q = 1 radius are highly desirable, such as early heating to delay the current penetration and, of course, active sawtooth destabilisation to mediate small frequent sawteeth and retain a small q = 1 radius.Comment: 29 pages, 16 figure

Advances in the physics studies for the JT-60SA tokamak exploitation and research plan

JT-60SA, the largest tokamak that will operate before ITER, has been designed and built jointly by Japan and Europe, and is due to start operation in 2020. Its main missions are to support ITER exploitation and to contribute to the demonstration fusion reactor machine and scenario design. Peculiar properties of JT-60SA are its capability to produce long-pulse, high-ß, and highly shaped plasmas. The preparation of the JT-60SA Research Plan, plasma scenarios, and exploitation are producing physics results that are not only relevant to future JT-60SA experiments, but often constitute original contributions to plasma physics and fusion research. Results of this kind are presented in this paper, in particular in the areas of fast ion physics, high-beta plasma properties and control, and non-linear edge localised mode stability studies.Postprint (published version