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.

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In the four-stage approach of the new ITER Research Plan, the first Pre-Fusion Power Operation (PFPO) phase will only have a limited power available from external Heating and Current Drive (H&CD) systems: 20 to 30 MW, provided by the Electron Cyclotron Resonance Heating system (ECRH). Accessing the H-mode confinement regime at such low auxiliary power requires operating at lower magnetic field, plasma current and density, i.e. 1.8 T and 5 MA for a density between 40% and 50% of the Greenwald density. H-mode plasmas at 5 MA / 1.8 T are also considered for the second PFPO phase when ITER will have its full installed H&CD capabilities, i.e. 20−30 MW of ECRH, 20 MW of Ion Cyclotron Resonance Heating (ICRH) and 33 MW of Neutral Beam Injection (NBI). The present paper describes the operational conditions of such scenarios in hydrogen and helium plasmas and the H&CD capabilities for these plasmas, to assess the viability of such scenarios and the issues that will be possible to address with them. The modelling results show that 5 MA / 1.8 T scenarios are viable and will allow the exploration of the H-mode physics and control issues foreseen in the ITER Research Programme in the PFPO phases.

Within the framework of the Adaptive Plasma Experiment (APEX) conceptual project, a trap with closed magnetic field lines, the Experimental Pseudo-Symmetric Closed Trap (EPSILON), is examined. The APEX project is aimed at theoretical and experimental development of the physical foundations for a steady state thermonuclear reactor designed on the basis of an alternative magnetic trap with tokamak-like large β plasma confinement. A discussion is given of the fundamental principle of pseudo-symmetry, which a magnetic configuration with tokamak-like plasma confinement should satisfy. Examples are given of calculations in the paraxial approximation of pseudo-symmetric curvilinear elements with a poloidal modulus B isoline. The EPSILON trap, consisting of two direct axisymmetric mirrors linked by two curvilinear pseudo-symmetric elements, is considered. To increase the equilibrium β, the plasma currents are short-circuited within curvilinear equilibrium elements. An untraditional scheme of MHD stabilization for a trap with closed field lines by use of axisymmetric mirrors with a divertor is analysed. The experimental installation EPSILON-One Mirror Element (OME), which is under construction for experimental investigation of stabilization by divertor, is discussed. The opportunity for applying the ECR method of plasma production in EPSILON-OME in conditions of high density and low magnetic field is examined.

An assessment of neutron production during the pre-fusion power-operation (PFPO) phase has been carried out for a representative set of plasma scenarios predicted by the ITER Research Plan. A range of heating systems, namely neutral beam injection (NBI) (hydrogen), electron cyclotron resonance heating (ECRH), and ion cyclotron resonance heating (ICRH) are planned to be used for PFPO studies in helium, hydrogen, and mixed hydrogen–helium plasmas. Fast ions (protons and 3He) originating from NBI and ICRH systems can increase neutron production in PFPO plasmas by directly interacting with intrinsic Be impurities or through secondary processes, as also evidenced at JET. The generation of fast ions in ITER PFPO scenarios has been modelled using the ASTRA-NBI and TORIC-SSFPQL codes. A significant impact of the synergy between hydrogen NBI and hydrogen-minority ICRH on neutron production in helium plasmas is reported. In addition, the stability of the toroidicity-induced Alfvén eigenmodes (TAE) is analyzed for PFPO plasmas with a high pressure of suprathermal ions and a weak reversed shear. The possible impact of sawtooth oscillations and TAEs on neutron production is discussed, based on a linear stability analysis.