Significant progress has been made in the area of advanced modes of operation that are candidates for achieving steady state conditions in a fusion reactor. The corresponding parameters, domain of operation, scenarios and integration issues of advanced scenarios are discussed in this chapter. A review of the presently developed scenarios, including discussions on operational space, is given. Significant progress has been made in the domain of heating and current drive in recent years, especially in the domain of off-axis current drive, which is essential for the achievement of the required current profile. The actuators for heating and current drive that are necessary to produce and control the advanced tokamak discharges are discussed, including modelling and predictions for ITER. The specific control issues for steady state operation are discussed, including the already existing experimental results as well as the various strategies and needs (qψ profile control and temperature gradients). Achievable parameters for the ITER steady state and hybrid scenarios with foreseen heating and current drive systems are discussed using modelling including actuators, allowing an assessment of achievable current profiles. Finally, a summary is given in the last section including outstanding issues and recommendations for further research and development.
IAEA-CN-116/EX/2-1 _______________________________________________________________________________________ This is a preprint of a paper intended for presentation at a scientific meeting. Because of the provisional nature of its content and since changes of substance or detail may have to be made before publication, the preprint is made available on the understanding that it will not be cited in the literature or in any way be reproduced in its present form. The views expressed and the statements made remain the responsibility of the named author(s); the views do not necessarily reflect those of the government of the designating Member State(s) or of the designating organization(s). In particular, neither the IAEA nor any other organization or body sponsoring this meeting can be held responsible for any material reproduced in this preprint.
Characteristics of internal transport barrier (ITB) structure are studied and the active ITB control has been developed in JT-60U reversed shear plasmas. The following results are found. Outward propagation of the ITB with steep T i gradient is limited to the minimum safety factor location (ρ qmin). However the ITB with reduced T i gradient can move to the outside of ρ qmin. Lower boundary of ITB width is proportional to the ion poloidal gyroradius at the ITB center. Furthermore the demonstration of the active control of the ITB strength based on the modification of the radial electric field shear profile is successfully performed by the toroidal momentum injection in different directions or the increase of heating power by neutral beams.
A tearing mode with m = 3 and n = 2, destabilized in the steady state high-β p H-mode discharges with edge localized mode (ELM), was completely stabilized by local heating and current drive using the 110 GHz first harmonic O-mode electron cyclotron (EC) wave. Here, m and n are poloidal and toroidal mode numbers, respectively. The optimum EC wave injection angle was determined by identifying the mode location from an electron temperature perturbation profile and a safety factor profile. The optimum injection angle was also determined by scanning a steerable mirror during a discharge. In a typical discharge where the tearing mode is completely stabilized, the ratio of the electron cyclotron heating power to the total heating power is 0.17, and the ratio of the EC driven current to the total plasma current is 0.02. Stored energy and neutron emission rate were higher for the case with EC wave injection than that without EC wave injection, which suggests that the reduction of the stored energy and the neutron emission rate was recovered by the tearing mode stabilization.
This paper reports results on the progress in steady-state high-βp ELMy H-mode discharges in JT-60U. A fusion triple product, nD(0)τETi(0), of 3.1 × 1020 m−3 s keV under full non-inductive current drive has been achieved at Ip = 1.8 MA, which extends the record value of the fusion triple product under full non-inductive current drive by 50%. A high-beta plasma with βN ∼ 2.7 has been sustained for 7.4 s (∼60τE), with the duration determined only by the facility limits, such as the capability of the poloidal field coils and the upper limit on the duration of injection of neutral beams. Destabilization of neoclassical tearing modes (NTMs) has been avoided with good reproducibility by tailoring the current and pressure profiles. On the other hand, a real-time NTM stabilization system has been developed where detection of the centre of the magnetic island and optimization of the injection angle of the electron cyclotron wave are done in real time. By applying this system, a 3/2 NTM has been completely stabilized in a high-beta region (βp ∼ 1.2, βN ∼ 1.5), and the beta value and confinement enhancement factor have been improved by the stabilization.
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