Lorenzo Figini

The EC Upper Launcher (UL) and Ex-vessel waveguide (EW) are integral components of the ITER Electron Cyclotron (EC) system [1], which plays a key role in achieving and maintaining the high temperatures required for nuclear fusion within the tokamak, providing plasma heating and current drive, and magnetohydrodynamics (MHD) stabilization, which are crucial for ITER’s operational success. The EC system is increasingly becoming the reference mechanism for plasma heating in both present and future devices. Notably, ITER is placing significant reliance on the EC system, with the recent re-baseline proposal initially doubling the EC installed power from 24 MW to 48 MW, ultimately reaching a final configuration of up to 80 MW. The F4E contribution to the current ITER baseline configuration comprises four Upper Launchers and five sets of ex-vessel waveguides. The UL and EW systems, currently in its Final Design phase under the frame of a Procurement Agreement between Europe and ITER Organization, have released the Final Design Documentation Package, which is here described, as input to the Final Design Gate Review that took place in July 2025. The full set of documents released is composed by: 1) a Design Description Dossier composed by a UL & EW Design Description (DD) document and supported by a set of 3D models, drawings and diagrams; 2) a Design Justification Dossier encompassing a set of end-to-end engineering analyses verifying the functionality and structural integrity of the components according to the applicable requirements, and the RAMI analysis of both systems; and 3) a Product lifecycle Dossier covering documentation for subsequent phases including manufacturing, assembly, installation, commissioning, operations, maintenance and decommissioning. Furthermore, the validation of the design is supported by currently on-going activities of prototyping and testing of the most critical components.

The Divertor Tokamak Test facility (DTT) is a device presently under construction at the ENEA site in Frascati (Italy) in the framework of a joint public/private partnership. It has been designed as a superconducting tokamak with breakeven class performance, with the main objective of developing credible solutions for heat and particle exhaust, a key challenge in view of future fusion reactors. This needs to be addressed in a core-edge integrated approach, to assess the compatibility of exhaust solutions with reactor relevant core performance. In this paper, an overview is provided of the DTT research plan, recently developed by an international team. It covers the DTT programmatic objectives, research strategy and expected scientific contributions connected with the device characteristics, not only in the key area of heat exhaust and edge plasma physics, but also on other subjects of high fusion relevance, such as MHD stability in high performance scenarios, transport and turbulence, energetic particle physics, validation of advanced theoretical developments, as well as tests of technological solutions for reactor relevant components.

The spherical tokamak for energy production (STEP) is the UK’s flagship program to design a prototype fusion power plant capable of delivering net electricity to the grid. Due to geometric constraints in spherical tokamaks, STEP is being developed for fully non-inductive operation, relying primarily on microwave-based electron cyclotron (EC) and electron Bernstein wave systems for plasma heating and current drive (HCD). A key requirement for achieving this operation is the optimization of these systems to ensure high wall-plug efficiency and reliable current profile control. This paper presents a self-consistent modeling approach using the JINTRAC framework, where the JETTO transport code is coupled with the GRAY quasi-optical beam-tracing code. Previous scenario development relied on simplified, prescribed EC current drive (ECCD) profiles. In this work, these were replaced with the profiles derived from detailed parametric GRAY scans, optimized for launcher geometry, frequency, and polarization. These realistic ECCD profiles are integrated into a fully self-consistent JETTO/GRAY simulation, capturing the dynamic interplay between plasma transport, heating, and current evolution. The methodology developed in this study offers a more accurate representation of HCD dynamics and is applicable to future STEP scenario development.

Future thermonuclear fusion reactor with magnetic confinement will require safe and reliable control system for the stabilization and/or suppression of Neo-classical Tearing Modes (NTM). The use of diagnostics sensitive to electron cyclotron emission (ECE) in different control methods and strategies is discussed considering the reliability and resolution that such diagnostics can achieve in a EU-DEMO experimental reactor.

In the initial stages of ITER operation, ELM mitigation systems need to be commissioned. This requires controlled flat-top operation in type-I ELMy H-mode regimes. Hydrogen or helium plasma discharges are used exclusively in these stages to ensure negligible production of neutrons from fusion reactions. With the expected higher L–H power threshold of hydrogen and helium plasmas compared to corresponding D and D/T plasmas, it is uncertain whether available auxiliary power systems are sufficient to operate in stable type-I ELMy H-mode. This has been investigated using integrated core and edge/SOL/divertor modelling with JINTRAC. Assuming that the L–H power threshold is well captured by the Martin08 scaling law, the presented simulations have found that 30 MW of ECRH power is likely required for the investigated hydrogen plasma scenarios, rather than the originally planned 20 MW in the 2016 Staged Approach ITER Baseline. However, past experiments have shown that a small helium fraction (∼10%) can considerably reduce the hydrogen plasma L–H power threshold. Assuming that these results extrapolate to ITER operation regimes, the 7.5 MA/2.65 T hydrogen plasma scenario is likely to access stable type-I ELMy H-mode operation also at 20 MW of ECRH.

This paper describes the first vertical electron cyclotron emission measurement of non-thermal electron distributions in the Tokamak à Configuration Variable. These measurements were conducted in runaway electron scenarios and in the presence of electron cyclotron current drive. Measured intensities of linearly polarized X- and O-mode radiation from fast electrons allow the analysis of the energy distribution. The measurements were made possible through the creation of an operational regime for the diagnostic that is free of thermal background radiation, in relaxed electron density operations. This operational regime notably enables the cross-calibration of the diagnostic system, relying on thermal plasma measurements and modeling with the ray-tracing code SPECE.

Designing a robust termination scenario for a burning ITER plasma is a challenge that requires extensive core plasma and divertor modelling. The presented work consists of coupled core/edge/SOL/divertor simulations, performed with the JINTRAC code, to study the Q = 10 flat-top phase and exit phase of the ITER 15 MA/5.3 T DT scenario. The modelling utilizes the recently implemented option to treat deuterium and tritium separately in the SOL/divertor, enabling a consistent treatment of deuterium and tritium in the whole plasma volume, which is a unique capability of JINTRAC. In addition, these are the first JINTRAC simulations of this scenario that use a first-principles transport model to self-consistently model the ECRH power deposition and to include tungsten while keeping track of tungsten sputtering and accumulation. The flat-top simulations demonstrate the possibility of sustaining a steady state fusion Q of 10 using pure deuterium gas puffs together with DT mixed pellets, which is an option to make a more effective use of tritium. Simulations of the exit phase are set up sequentially, with each phase providing initial conditions for the next, starting with a density decay at full current and auxiliary power, and demonstrate the possibility of reducing the density robustly within a few seconds. Following the density decay, a subsequent auxiliary power ramp-down in H-mode is performed with a late H–L transition at low auxiliary power, which may provide an option for the optimization of the plasma termination. The final ramp-down phase consists of a current ramp-down in L-mode to 3.75 MA.

A fusion reactor based on the spherical tokamak is very likely to be completely non-inductive for the majority of the plasma ramp-up and steady-state phases, due to the limitations imposed on the central coil assemblies by the compact design. Efficiency gains from solenoid-driven current cannot be relied upon. It is also critical that an electricity-producing plant maximises the wall-plug efficiency of its heating and current drive (HCD) system, this being one of the largest consumers of recirculating power. It is therefore essential that the HCD system is well-optimised for current drive efficiency in order to meet the goal of net electricity production. The UK’s Spherical Tokamak for Energy Production (STEP) reactor design program has recently taken the decision to use exclusively microwave-based heating and current drive actuators for its reactor concepts. We present the optimisation of an electron cyclotron current drive scheme for a spherical tokamak reactor, based around the STEP concept, arriving at a solution which overcomes the limitations imposed by the spherical tokamak geometry in terms of microwave access and high trapped particle fraction. The solution uses high-field side absorption and a mix of fundamental and 2nd harmonic O mode, with overall power requirements reducing with increasing number of frequencies used. An additional fundamental frequency is also added to further boost the efficiency during non-inductive plasma ramp.

The Divertor Tokamak Test (DTT) facility is equipped with auxiliary heating systems in order to be able to load the divertor with a power flux relevant to study the power exhaust issue in a reactor relevant range of parameter. The powerful system is the Electron Cyclotron Heating (ECH) with an installed power of 32 MW in its largest extension. Together with the bulk heating of the DTT plasma, the ECH system will cover several tasks for the plasma operation. This paper summarizes the main characteristics and design choices of the DTT ECH system and the related physics studies, based on the reference DTT plasma, to develop and control the plasma, fulfilling the functional tasks, with the support of simulation activities. Dedicated studies have been carried out to investigate the capability of EC power to assist plasma start-up, stabilize MHD activity and support current ramp up/down. In addition, it has been studied how changes of the ECH power distribution can have an impact on the plasma profiles, affecting the fueling pellet effectiveness and MHD modes.

ITER is of key importance in the European fusion roadmap as it aims to prove the scientific and technological feasibility of fusion as a future energy source. The EUROfusion consortium of labs within Europe is contributing to the preparation of ITER scientific exploitation and operation and aspires to exploit ITER outcomes in view of DEMO. The paper provides an overview of the major progress obtained recently, carried out in the frame of the new (initiated in 2021) EUROfusion work-package called ‘Preparation of ITER Operation’ (PrIO). The overview paper is directly supported by the eleven EUROfusion PrIO contributions given at the 29th Fusion Energy Conference (16–21 October 2023) London, UK [www.iaea.org/events/fec2023]. The paper covers the following topics: (i) development and validation of tools in support to ITER operation (plasma breakdown/burn-through with evolving plasma volume, new infra-red synthetic diagnostic for off-line analysis and wall monitoring using Artificial Intelligence techniques, synthetic diagnostics development, development and exploitation of multi-machine databases); (ii) R&D for the radio-frequency ITER neutral beam sources leading to long duration of negative deuterium/hydrogen ions current extraction at ELISE and participation in the neutral beam test facility with progress on the ITER source SPIDER, and, the commissioning of the 1 MV high voltage accelerator (MITICA) with lessons learned for ITER; (iii) validation of neutronic tools for ITER nuclear operation following the second JET deuterium–tritium experimental campaigns carried out in 2021 and in 2023 (neutron streaming and shutdown dose rate calculation, water activation and activated corrosion products with advanced fluid dynamic simulation; irradiation of several materials under 14.1 MeV neutron flux etc).