D. Stork

ELM control strategies and tools: status and potential for ITER

Operating ITER in the reference inductive scenario at the design values of Ip = 15 MA and QDT = 10 requires the achievement of good H-mode confinement that relies on the presence of an edge transport barrier whose pedestal pressure height is key to plasma performance. Strong gradients occur at the edge in such conditions that can drive magnetohydrodynamic instabilities resulting in edge localized modes (ELMs), which produce a rapid energy loss from the pedestal region to the plasma facing components (PFC). Without appropriate control, the heat loads on PFCs during ELMs in ITER are expected to become significant for operation in H-mode at Ip = 6–9 MA; operation at higher plasma currents would result in a very reduced life time of the PFCs. Currently, several options are being considered for the achievement of the required level of ELM control in ITER; this includes operation in plasma regimes which naturally have no or very small ELMs, decreasing the ELM energy loss by increasing their frequency by a factor of up to 30 and avoidance of ELMs by actively controlling the edge with magnetic perturbations. Small/no ELM regimes obtained by influencing the edge stability (by plasma shaping, rotational shear control, etc) have shown in present experiments a significant reduction of the ELM heat fluxes compared to type-I ELMs. However, so far they have only been observed under a limited range of pedestal conditions depending on each specific device and their extrapolation to ITER remains uncertain. ELM control by increasing their frequency relies on the controlled triggering of the edge instability leading to the ELM. This has been presently demonstrated with the injection of pellets and with plasma vertical movements; pellets having provided the results more promising for application in ITER conditions. ELM avoidance/suppression takes advantage of the fact that relatively small changes in the pedestal plasma and magnetic field parameters seem to have a large stabilizing effect on large ELMs. Application of edge magnetic field perturbation with non-axisymmetric fields is found to affect transport at the plasma edge and thus prevent the uncontrolled rise of the plasma pressure gradients and the occurrence of type-I ELMs. This paper compiles a brief overview of various ELM control approaches, summarizes their present achievements and briefly discusses the open issues regarding their application in ITER.

On the heating mix of ITER

This paper considers the heating mix of ITER for the two main scenarios. Presently, 73 MW of absorbed power are foreseen in the mix 20/33/20 for ECH, NBI and ICH. Given a sufficient edge stability, Q = 10-the goal of scenario 2-can be reached with 40MW power irrespective of the heating method but depends sensitively inter alia on the H-mode pedestal temperature, the density profile shape and on the characteristics of impurity transport. ICH preferentially heats the ions and would contribute specifically with Delta Q 0.5, and strong off-axis current drive (CD). The findings presented here are based on revised CD efficiencies gamma for ECCD and a detailed benchmark of several CD codes. With ECCD alone, the goals of scenario 4 can hardly be reached. Efficient off-axis CD is only possible with NBI. With beams, inductive discharges with f(ni) > 0.8 can be maintained for 3000 s. The conclusion of this study is that the present heating mix of ITER is appropriate. It provides the necessary actuators to induce in a flexible way the best possible scenarios. The development risks of NBI at 1 MeV can be reduced by operation at 0.85 MeV.

The use of neutral beam heating to produce high performance fusion plasmas, including the injection of tritium beams into the joint European Torus (JET)

The neutral beam injection (NBI) system of the Joint European Torus (JET) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1985), Vol. 1, p. 11] has proved to be an extremely effective and flexible heating method capable of producing high performance plasmas and performing a wide range of related physics experiments. High fusion performance deuterium plasmas have been obtained in the hot‐ion (HI) H‐mode regime, using the central particle fueling and ion heating capabilities of the NBI system in low target density plasmas, and in the pellet enhanced plasma (PEP) H‐mode regime, where the good central confinement properties of pellet fueled plasmas are exploited by additional heating and fueling as well as the transition to H mode. The HI H‐mode configuration was used for the First Tritium Experiment (FTE) in JET in which NBI was used to heat the plasma using 14 D0 beams and, for the first time, to inject T0 using the two remaining beams. These plasmas had a...

Tritium operation of the JET neutral beam systems

Neutral Beam Injection (NBI) is a very flexible auxiliary heating method for tokamak plasmas, capable of being efficiently coupled to the various plasma configurations required in the Deuterium–Tritium Experimental campaign (DTE1) undertaken in the JET device during 1997. In particular, experiments for high fusion yield and amplification factor Q require intense NBI heating, and for maximum performance and optimum fuel mixture control in Deuterium–Tritium (D–T) plasmas it was necessary to operate the JET NBI systems in both deuterium and tritium. All technical aspects of the modifications to the JET NBI systems for compatibility with tritium operation are discussed, and the associated commissioning is described, including preparatory commissioning using deuterium. Problems experienced and their resolution are highlighted. Some specific beamline physics issues relating to tritium operation are discussed in detail, in particular experimental measurements of beam-target D–T reactions occurring in beam-stopping elements and associated modelling of isotope exchange in these components. Data on NBI performance and tritium usage and recovery for the DTE1 campaign are presented.

Overview of transport, fast particle and heating and current drive physics using tritium in JET plasmas

Results are presented from the JET Trace Tritium Experimental (TTE) campaign using minority tritium (T) plasmas (n(T)/n(D) 2 MA) and monotonic q-profiles. In CH discharges the gamma-ray emission decay times are much lower than classical (tau(Ts) + tau(alpha s)), indicating alpha confinement degradation, due to the orbit losses and particle orbit drift predicted by a 3-D Fokker-Planck numerical code and modelled using TRANSP.

Tritium fast ion distribution in JET current hole plasmas

Current hole plasmas in JET are those in which the current density within r/a < 0.3 is close to zero. Tritium ions injected quasi-tangentially into such plasmas can fulfil a stagnation condition whereby their vertical drift is cancelled by the poloidal component of their parallel velocity. These ions remain trapped at approximately 0.2 m from the plasma axis and can be detected by a distortion in the neutron emission profile. Numerical modelling of the steady-state distribution reproduces the experimental results while the decay of neutron emission after the cessation of injection is found to be sensitive to small changes in the q-profile.

Effects of enhanced toroidal field ripple on JET plasmas

The JET machine is equipped with 32 toroidal field coils. In order to study the effect of TF ripple on the confinement of fast particles and, more generally, on the plasma behaviour, a series of experiments was performed using only 16 TF coils. At the position of the outer limiter, this led to an increase of the ripple, delta =(Bmax-Bmin)/(Bmax+Bmin), from 1% to 12.5%. The toroidal field was limited to 1.4 T, with plasma currents in the range between 2 and 3 MA. Additional heating power-levels and energy-input were kept low in order to avoid possible damage to some first wall components made out of Inconel. Experiments were carried out using 140 keV NBI injected deuterons, ICRF accelerated protons and deuterons ( approximately 0.5 to approximately 2 MeV) and 1 MeV tritons from DD reactions.

Confinement of fusion alpha-particles in JET hollow current equilibrium

The effect of a toroidal current hole on the first orbit (FO) loss and on the collisional loss of alpha particles in JET is investigated. Numerical results of predictive three-dimensional Fokker–Planck modelling of the distribution function of D–T fusion alphas in hollow current JET discharges are presented. If the current hole region is kept reasonably small, it induces only a moderate increase of FO losses as well as of the collisional loss of fast alphas. The current hole effect is shown to be qualitatively equivalent to a reduction of the total plasma current I. Hence, the alpha confinement degradation by the current hole profiles can be compensated by enlarging I.

European negative ion based neutral beam developments

Abstract The European negative ion based neutral beam development programme consists presently of experiments on high energy electrostatic accelerators and high current D − sources, and of neutral beam physics and system design studies. At CEA-Cadarache the source test bed MANTIS started operation last June, and the 1 MV, 0.1 A, D − acceleration project SINGAP will be operational by the end of 1994. Additionally, negative ion source physics-oriented experiments will be conducted at Dublin City University. A Pagoda source, developed for the SINGAP experiment, recently produced more than 1 A D − beams and j (D − ) of 10 mA cm −2 , with Cs seeding, at 0.6 Pa operating pressure. 1.3 A H − and 0.5 A D − beams have also been obtained without the use of Cs during the first experiments on MANTIS. Studies on neutral beam heating (NBH) for a next step machine have been conducted during 1993, mainly at JET and CEA-Cadarache. NBH physics studies based on the RLW model indicate that ignition on ITER-EDA should be reached with about 40 MW of 0.4–1 MeV D 0 beams. The outlines of an NBH system for ITER-EDA, based on SINGAP, have been studied.

Overview on Experiments On ITER‐like Antenna On JET And ICRF Antenna Design For ITER

Following an overview of the ITER Ion Cyclotron Resonance Frequency (ICRF) system, the JET ITER‐like antenna (ILA) will be described. The ILA was designed to test the following ITER issues: (a) reliable operation at power densities of order 8 MW/m2 at voltages up to 45 kV using a close‐packed array of straps; (b) powering through ELMs using an internal (in‐vacuum) conjugate‐T junction; (c) protection from arcing in a conjugate‐T configuration, using both existing and novel systems; and (d) resilience to disruption forces. ITER‐relevant results have been achieved: operation at high coupled power density; control of the antenna matching elements in the presence of high inter‐strap coupling, use of four conjugate‐T systems (as would be used in ITER, should a conjugate‐T approach be used); operation with RF voltages on the antenna structures up to 42 kV; achievement of ELM tolerance with a conjugate‐T configuration by operating at 3Ω real impedance at the conjugate‐T point; and validation of arc detection sys...