Progress in the area of MHD stability and disruptions, since the publication of the 1999 ITER Physics Basis document Nucl. Fusion 39 2137-2664, is reviewed. Recent theoretical and experimental research has made important advances in both understanding and control of MHD stability in tokamak plasmas. Sawteeth are anticipated in the ITER baseline ELMy H-mode scenario, but the tools exist to avoid or control them through localized current drive or fast ion generation. Active control of other MHD instabilities will most likely be also required in ITER. Extrapolation from existing experiments indicates that stabilization of neoclassical tearing modes by highly localized feedback-controlled current drive should be possible in ITER. Resistive wall modes are a key issue for S128 Chapter 3: MHD stability, operational limits and disruptions advanced scenarios, but again, existing experiments indicate that these modes can be stabilized by a combination of plasma rotation and direct feedback control with non-axisymmetric coils. Reduction of error fields is a requirement for avoiding non-rotating magnetic island formation and for maintaining plasma rotation to help stabilize resistive wall modes. Recent experiments have shown the feasibility of reducing error fields to an acceptable level by means of non-axisymmetric coils, possibly controlled by feedback. The MHD stability limits associated with advanced scenarios are becoming well understood theoretically, and can be extended by tailoring of the pressure and current density profiles as well as by other techniques mentioned here. There have been significant advances also in the control of disruptions, most notably by injection of massive quantities of gas, leading to reduced halo current fractions and a larger fraction of the total thermal and magnetic energy dissipated by radiation. These advances in disruption control are supported by the development of means to predict impending disruption, most notably using neural networks. In addition to these advances in means to control or ameliorate the consequences of MHD instabilities, there has been significant progress in improving physics understanding and modelling. This progress has been in areas including the mechanisms governing NTM growth and seeding, in understanding the damping controlling RWM stability and in modelling RWM feedback schemes. For disruptions there has been continued progress on the instability mechanisms that underlie various classes of disruption, on the detailed modelling of halo currents and forces and in refining predictions of quench rates and disruption power loads. Overall the studies reviewed in this chapter demonstrate that MHD instabilities can be controlled, avoided or ameliorated to the extent that they should not compromise ITER operation, though they will necessarily impose a range of constraints.

The paper discusses edge stability, beta limits and power handling issues for negative triangularity tokamaks. The edge MHD stability is the most crucial item for the power handling. For the case of negative triangularity the edge stability picture is quite different from that for conventional positive triangularity tokamaks: the 2nd stability access is closed for localized Mercier/ballooning modes due to the absence of magnetic well, and nearly internal kink modes set the pedestal height limit weakly sensitive to diamagnetic stabilization just above the margin of localized mode Mercier criterion violation. While negative triangularity tokamak is thought to have low beta limit with its magnetic hill property, it is found that plasmas with reactor relevant values of normalized beta β N > 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 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 case of peaked pressure profile leading to Mercier unstable core. With the theoretical predictions supported by experiments, a negative triangularity tokamak would become a perspective fusion energy system with other advantages including larger separatrix wetted area, more flexible divertor configuration design, wider trapped particle free SOL, lower background magnetic field for internal poloidal field coils and larger pumping conductance from the divertor room.

Tokamak scenario development requires understanding of the properties that determine the kinetic profiles in non-steady plasma phases, and of self-consistent evolution of the magnetic equilibrium. Current ramps are of particular interest since many transport-relevant parameters explore a large range of values and their impact on transport mechanisms has to be assessed. To this purpose a novel full-discharge modeling tool has been developed, which couples the transport code ASTRA [1] and the free boundary equilibrium code SPIDER [2], utilizing a specifically designed coupling scheme. The current ramp-up phase can be accurately and reliably simulated using this scheme, where a plasma shape, position, and current controller is applied, which mimics the one of ASDEX Upgrade. Transport of energy is provided by theory-based models (e.g. TGLF [3]). A recipe based on edge-relevant parameters [4] is proposed to resolve the low current phase of the current ramps, where the impact of the safety factor on microinstabilities could make quasi-linear approaches questionable in the plasma outer region. Current ramp scenarios, selected from ASDEX Upgrade discharges, are then simulated to validate both the coupling with the free-boundary evolution and the prediction of profiles. Analysis of the underlying transport mechanisms is presented, to clarify the possible physics origin of the observed L-mode empirical energy confinement scaling. The role of toroidal micro-instabilities (ITG, TEM) and of non-linear effects is discussed.

Requirements for pellet injection parameters for plasma fuelling are assessed for ITER scenarios with enhanced particle confinement. The assessment is based on the integrated transport simulations including models of pedestal transport, reduction of helium transport and boundary conditions compatible with SOL/divertor simulations. The requirements for pellet injection for the inductive H-mode scenario (H H98y,2 = 1) are reconsidered taking account of a possible reduction of the particle loss obtained in some experiments at low collisionalities. The assessment of fuelling requirements is carried out for the hybrid and steady state scenarios with enhanced confinement with H H98y,2 > 1. A robustness of plasma performance to the variation of particle transport is demonstrated. A new type of steady state (SS) scenario is considered with neutral beam current drive (NBCD) and electron cyclotron current drive (ECCD) instead of lower hybrid current drive (LHCD) to extend the range of stable operation and to avoid the reduction of the edge LHCD efficiency caused by pellet injection.

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The operational space (I p − n) for long-pulse scenarios ( t burn ∼ 1000 s, Q 5) of ITER has been assessed by 1.5D core transport modelling with pedestal parameters predicted by the EPED1 code by a set of transport codes under a joint activity carried out by the Integrated Operational Scenario ITPA group. The analyses include the majority of transport models (CDBM, GLF23, Bohm/gyroBohm (BgB), MMM7.1, MMM95, Weiland, scaling-based) presently used for interpretation of experiments and ITER predictions. The EPED1 code was modified to take into account boundary conditions predicted by SOLPS4 for ITER. In contrast to standard EPED1 assumptions, EPED1 with the SOLPS boundary conditions predicts no degradation of the pedestal pressure as density is reduced. Lowering the plasma density to n e ∼ (5-6) × 10 19 m −3 leads to an increased plasma temperature (similar pedestal pressure), which reduces the loop voltage and increases the duration of the burn phase to t burn ∼ 1000 s with Q 5 for I p 13 MA at moderate normalized pressure (β N ∼ 2). These ITER plasmas require the same level of additional heating power as the reference Q = 10 inductive scenario at 15 MA (33 MW NBI and 17-20 MW EC heating and current drive power). However, unlike the 'hybrid' scenarios considered previously, these H-mode plasmas do not require specially shaped q profiles nor improved confinement in the core for the transport models considered in this study. Thus, these medium density H-mode plasma scenarios with I p 13 MA present an attractive alternative to hybrid scenarios to achieve ITER's long-pulse Q 5 scenario and deserve further analysis and experimental demonstration in present tokamaks.

One of the three goals of the ITER project is to demonstrate fusion gain Q ≥ 5 in steady-state operation (SSO). A reassessment of the necessary conditions for Q ≥ 5 SSO in ITER has been carried out for steady-state scenarios assuming no internal transport barrier in the core plasma and with the sole use of neutral beam injection (NBI) and electron cyclotron heating and current drive (EC H&CD) as heating and current drive sources in these scenarios. The parametric operational space for SSO in ITER has been reassessed utilizing an inverse evaluation approach that takes into account the baseline design of the NBI and EC H&CD systems with P NBI = 33 MW, P EC = 20 MW and their possible upgrade capabilities included in their design, P NBI ≤ 49.5 MW, and P EC ≤ 30 MW. The optimal operational points from this evaluation approach have been chosen for detailed 1.5-D transport modelling by ASTRA and followed-up by magnetohydrodynamic (MHD) stability analysis to demonstrate the conditions in which the Q = 5 SS goal can be achieved in ITER where the plasma current is fully accounted for by the driven current by NBI and EC H&CD and the self-driven bootstrap current. The ideal MHD stability of plasma configurations with self-consistently simulated plasma profiles and equilibrium for the chosen OPs has been analysed by the KINX code. Using this analysis, the possibility to control the MHD stability of these steady-state plasmas by tailoring the current profile with the flexibility provided by design of the systems for electron cyclotron current drive and neutral beam current drive in ITER has been demonstrated. Issues related to the realization of such scenarios from the point of view of plasma physics, experimental demonstration and design limits of the ITER systems and components is discussed.

This paper considers a fast track to non-energy applications of nuclear fusion that is associated with the ‘fusion for neutrons’ (F4N) paradigm. Being a useful product accompanying energy, fusion neutrons are more valuable than the energy released in DT reactions and they are urgently needed for research purposes and to develop and validate modern technologies. In the near future neutron yield in fusion devices will become significantly larger than that of fission and accelerator sources. This paper describes a compact tokamak fusion neutron source based on a small spherical tokamak (FNS-ST) with a MW range of DT fusion power and considers the key physics issues of this device. The major and minor radii are ∼0.5 and ∼0.3 m with magnetic field ∼1.5 T, heating power less than 15 MW and plasma current 1–2 MA. The production rate of DT neutrons of (3–10) × 1017 n s−1 and their flux at the first wall of 0.2 MW m−2 ensure that the device is capable of fusion–fission demonstration experiments. The problems of major concern are discharge initiation, current drive, plasma—fast ion beam stability and high first wall and divertor loads. The conceptual design provides solutions to these problems and suggests the feasibility of the FNS-ST.