Speakers
Monday 21 September 9:40 am
ITER is moving on a decisive phase, marked by significant technical progress, an increasingly well defined schedule, and strong performance against the project’s new baseline. Two years after its definition, ITER is now operating with a schedule performance index above 1, reflecting the project’s steady recovery, improved planning discipline, and increased predictability across all major work packages. This presentation will provide an up to date overview of the status of construction, manufacturing, and assembly across key ITER systems—magnets, vacuum vessel sectors, cryogenics, plasma heating, and diagnostics. It will highlight recent milestones achieved by the ITER Organization and the Domestic Agencies, as well as the remaining challenges associated with integration, commissioning, and the path to first plasma.
Beyond the project’s internal progress, ITER’s broader role within a rapidly evolving fusion ecosystem has never been more relevant. As private fusion initiatives accelerate and national fusion programs expand, ITER remains a key scientific and engineering anchor point for validating burning plasma physics, demonstrating integrated fusion technologies, and establishing safety and regulatory frameworks. The talk will discuss how ITER’s experimental mission complements the ambitions of next generation pilot plants, how it supports the fusion supply chain, and how it contributes to the international knowledge base that future commercial reactors will depend on.
By situating ITER within this global context, the presentation will clarify how the project’s outcomes—scientific, technological, and collaborative—will shape the fusion landscape for decades to come, ensuring that ITER remains central to the world’s pursuit of fusion energy.
Monday 21 September 10:20 am
Tuesday 22 September 8:30 am
IFMIF-DONES (International Fusion Materials Irradiation Facility, DEMO-Oriented NEutron Source) is a cutting-edge neutron irradiation facility designed to simulate the high-energy neutron environment of future fusion reactors. It also contributes to the development of tritium and breeding blanket technologies. As part of the European roadmap to fusion electricity, its primary objective is to create a comprehensive database of material properties under intense neutron irradiation conditions, similar to those encountered in a fusion reactor. The neutron source is generated by an accelerated deuteron beam striking a liquid lithium curtain, producing neutrons with an energy spectrum and flux comparable to those experienced by the first wall of a fusion reactor.
The IFMIF-DONES facility construction phase, part of the DONES Programme, has progressed from preliminary to detailed design, with some parts already in manufacturing process. Construction officially started after the first DONES Steering Committee met on 16 March 2023, facilitating the transfer of design authority to the DONES Programme Team. The Multilateral International DONES Agreement, signed on 21 November 2025, secured contributions for the construction phase. This paper reviews the current progress of the DONES Programme and the IFMIF-DONES Facility.
This contribution will focus on the design status of the Facility, the progress of facility construction, anticipated performance, including a summary of its experimental features and programs.
Friday 25 September 11:50 am
The Laser Mégajoule (LMJ) [1] is currently the second most powerful laser in operation in the world after the National Ignition Facility (NIF) in the USA. It currently performs experiments with 120 nanosecond laser beams and 450 kJ of laser energy at 3ω (351 nm). Once all beamlines in operation in early 2027, it will ultimately have 176 laser beams, and the laser energy will be ramped up to 1.3 MJ. As of 2026, 25 plasma diagnostics are operational around the target chamber. I will describe this exhaustive diagnostics suite ranging from optical diagnostics to time-resolved hard X-ray microscopes (200 ps temporal resolution, 10 microns spatial resolution) for central hotspot characterization. Notably, the VISAR system (1ω, 2ω) has seen its probe duration extended to 100 ns, with an improved fringe contrast. An absolutely-calibrated, time-resolved NBI (near-backscattering- imaging) diagnostic is also actively used for laser-plasma interaction (LPI) studies in indirect drive experiments, as well as in direct drive experiments involving cross-beam energy transfer (CBET).
The LMJ has been built to produce experimental data for the Simulation program [2]. An additional laser, PETAL (PETAWATT Aquitaine Laser) [4], operating in short pulses (ps) and with kJ-class energy is coupled to the same target chamber. PETAL exceeded the PW focused at the center of the chamber in 2025. This laser, used by the academic community, could generate intense sources of particles (protons, electrons, neutrons) to probe hot, dense plasmas [4,5,6]. I will present the most recent experiments carried out with LMJ-PETAL, striving to describe the various physics themes necessary for mastering Inertial Confinement Fusion in indirect drive [7]. I will also detail the future use of LMJ for the benefit of academic and industrial research on Inertial Fusion Energy, in the context of France 2030 initiative and in collaboration with the start-up GenF.
[1] A. Casner et al., High Energy Density Physics 17, 2-11 (2015).[2] E. Lefebvre et al., Nuclear Fusion 59, 032010 (2019).[3] N. Blanchot et al., Opt. Express 25, 16957 (2017).[4] D. Raffestin et al., Matter and Radiation at Extremes 6, 056901 (2021).[5] F. Brun et al., Matter and Radiation at Extremes 9, 057203 (2024).[6] A.F.A. Bott et al., Phys. Rev. Lett. 127, 175002 (2021).[7] S. Liberatore et al., Phys. Plasmas 30? 122707 (2023).
Wednesday 23 September 9:10 am
The emergence of a fusion industry is an exciting time. The UK Spherical Tokamak for Energy Production (STEP) programme aims to demonstrate not only steady state net electric power from fusion, tritium self-sufficiency and practical maintenance but also to develop a supply chain capable of delivering the fusion power plants of the future. The programme leverages years of UK investment in skills and the development of critical technologies.
The STEP prototype design exploits the high self-driven (bootstrap) plasma current and high ratio of plasma pressure to magnetic field (beta) achievable in a spherical tokamak to target steady state power with a relatively low confining magnetic field and relatively compact size. These unique features set it apart from other approaches and broaden the exciting – and increasingly competitive – global development landscape.
With most areas from the plasma itself to materials pushing the boundaries of existing understanding, the evolution of the plant design requires careful balancing and management of the technical risks to be run.
To pursue the programme goals within an acceptable risk appetite the current phase of the programme features significant effort in technology demonstration, including:
- model coils to prove out quench-safe HTS toroidal field coils, jointed to permit vertical-access maintenance;
- extensive materials testing including high heat flux and irradiation campaigns
- development of multi-stage depressed collector gyrotrons to reduce recirculating power
- Electron Bernstein Wave current drive on MAST-Upgrade
- fast ramp power cycle rig
- tritium breeder to carrier gas exchange characterisation
At the same time preparations for the construction of this major infrastructure at the West Burton site – i.e. planning consent, regulatory engagement and the integration of major industry partners – continue at pace.
Tuesday 22 September 9:10 am
Thursday 24 September 9:50 am
Thursday 24 September 8:30 am
The Wendelstein 7-X (W7-X) stellarator is the world’s largest superconducting optimized stellarator designed to demonstrate the viability of the stellarator concept for steady-state fusion power. Unlike tokamaks, W7-X employs a complex, three-dimensional magnetic configuration engineered to minimize neoclassical transport and enable very good plasma confinement without relying on large internal plasma currents. The device is specifically designed to achieve long-pulse, high-performance plasmas: up to 1800 seconds at 10 MW electron cyclotron resonance heating (ECRH). This mission is enabled by a suite of technically demanding and fully qualified systems, including a superconducting magnet system with robust quench protection, an actively water-cooled first wall, divertor, ports, and plasma vessel, and a low-loss quasi-optical ECRH beam line.
Recent operational campaigns have successfully validated the design principles of W7-X. The machine has achieved a record energy turnaround of 1.8 GJ in a 350-second pulse (5 MW ECRH), and demonstrated high-performance plasmas with tokamak-like H-mode characteristics (Ti ≈ Te ≈ 2.8 keV) for up to 10 seconds under combined neutral beam and ECRH heating. Furthermore, W7-X has reached record stellarator values of the plasma triple product (nτT), approaching modern tokamak performance, and has realized high-beta scenarios with a volume-averaged β of 3% and a peak core β of 10%, underlining the stability and efficiency of the optimized magnetic configuration under significant thermal pressure. The recent successful deployment of a steady-state-capable extruder pellet injector extends the operational envelope significantly.
However, the transition to the final step of the long-pulse, high-performance plasma program reveals further engineering challenges. Increasing heating power leads to escalating heat loads on plasma-facing components (PFCs), demanding enhanced thermal management, power transfer throughput and structural integrity. Other challenges include integrated operation experience, investigation and selection of suitable magnetic field configurations and the development of feedback schemes for detached divertor operation.
To overcome the hurdles a systematic development strategy has been implemented, based on stepwise system hardening, failure mode and effects analysis (FMEA), and non-conformity management as well as advanced modeling via finite-element calculations. This paper presents the comprehensive technical and operational roadmap for the transition to the full design envelope of W7-X, detailing the current performance status, and the development of its core building blocks.
Friday 25 September 10:30 am
In recent years, research and technology development activities devoted to the achievement of fusion energy have expanded worldwide at a very rapid pace. The research environment has also changed substantially, with private actors now actively leading development programs with aggressive schedules. Within this frame, magnetic confinement fusion approaches have identified High-Temperature Superconducting (HTS) technologies as one of the key enablers for the achievement of commercial fusion plants, either integrating Low-Temperature Superconductors (LTS) to improve
performance, or completely substituting them, thus allowing the access to different operating temperature / field ranges and proposing alternative approaches to magnet technology.
HTS, Rare-Earth-Barium-Copper Oxide (REBCO) coated conductors have been demonstrated the capability to meet fusion magnets requirements, not only in terms of superconducting performance, but also in terms of industrial production capabilities and maturity. In parallel, a lot of effort is being spent in the R&D on multi-tape, high-current / high-field conductors, used to wind the magnets, with a number of prototypes proposed and tested. After recalling the main challenges and development steps undertaken on large size, high-current LTS Cable-in-Conduit Conductors for fusion, the state-of-theart of HTS-based ones will be illustrated, with respect to the typical requirements for fusion coils, and in particular for the specific features and operating regimes of HTS technologies. The main challenges toward the achievement of a truly mature and robust technology for high-current HTS conductors will be commented, along with the on-going or planned experimental activities to address them. As a relevant example, the approach being pursued for the hybrid (LTS + HTS) DTT Central Solenoid system will be discussed, based on the outcome of the development and testing activities carried out on the SECAS-BRAST HTS conductor concept.
Friday 25 September 11:10 am
As fusion research advances toward the realization of burning plasma and reactor-scale devices, the increasing complexity of plasma behavior calls for a shift in control and data analysis. Artificial intelligence (AI) is emerging as a key tool, offering capabilities in real-time prediction, adaptive control, and large-scale data interpretation.
KSTAR is a fully superconducting tokamak capable of steady-state, long-pulse operation, providing a uniquely advantageous environment for repeated trial-and-error control experiments at a scale where disruption events can be managed without prohibitive risk. This capability makes KSTAR an optimal platform for systematically developing and validating advanced plasma control strategies under reactor-relevant conditions.
KSTAR is planned to further evolve into a comprehensive test bed for AI-driven plasma control and data analysis. A next-generation plasma control system (PCS), designed to meet ITER level requirements, will support the integration of AI-based control modules. In parallel, KSTAR experimental data will be systematically transformed into AI-ready formats, supporting the deployment of AI agents for real-time experimental assistance, decision support, and automated optimization of plasma scenarios.
To support these developments, a wide range of diagnostics and actuators, designed with applicability to future demonstration reactors, will be installed and integrated. These upgrades aim to establish a closely connected framework across diagnostics, modeling, and control, enabling predictive and increasingly autonomous plasma operation. Machine learning approaches will play a key role in plasma state estimation, instability prediction, and adaptive feedback control, improving operational reliability and performance.
This presentation will highlight KSTAR’s ongoing achievements and future plans in advancing AI-enabled plasma control. Particular emphasis will be placed on its role as a key national facility contributing to the development of reactor-relevant technologies and bridging the gap toward next-step devices such as DEMO. By combining its unique operational capabilities with planned AI integration and system upgrades, KSTAR is expected to make a significant contribution to the realization of burning plasma and the broader advancement of fusion energy.
Thursday 24 September 9:10 am
The Divertor Tokamak Test (DTT) [1] facility represents a crucial milestone in the international roadmap toward commercial fusion energy. Management of the power exhaust remains one of the most significant engineering challenges. High heat fluxes on plasma–facing components can compromise the integrity of the machine, making the validation of reliable exhaust solutions mission–critical. DTT is specifically designed to fill this knowledge gap, serving as a high–field, flexible testbed to investigate various divertor configurations and liquid metal technologies under reactor–relevant conditions. By bridging the gap between current experiments and the next generation of fusion power plants, DTT ensures that the heat exhaust strategies are robust and sustainable.
The project foresees reaching more than 80% of total construction commitments in 2026, securing the timeline for operations to commence in the early 2030s. A significant development is the start of new building construction in the second half of 2026, and in particular the 35m cubic Torus Hall, following the demolition of existing site structures. This progress enables tokamak assembly to begin in the second half of 2029.
The start of Vacuum Vessel (VV) manufacturing marks also a major turning point. By resolving numerous interfaces with in–vessel and out–vessel components, the VV fabrication schedule anchors the three–phase assembly strategy: i) Final design of assembly procedures and tooling; ii) Qualification of special processes, such as superconducting joint insulation and on–site in–vessel coil fabrication; iii) Assembly of the tokamak within the Torus Hall.
To ensure timely execution, all primary components are already in production. All 18 superconducting Toroidal Field (TF) winding packs are either completed or in the final stages of fabrication. Integration into their casing structures will begin in early 2027, with the first completed TF coil scheduled for delivery in early 2028. Testing will take place at the new cryogenic facility currently being established at ENEA. Simultaneously, the Poloidal Field (PF) coil system is moving into production, with the manufacturing contract slated for signature in the first half of 2026. Power supply for the in–vessel and TF coils have been delivered on site. Procurement of the materials for the 54 divertor modules has started. ECH and ICH main components are in advanced manufacturing phase. A satellite facility for the testing of the remote handling system will be completed in the second
half of 2026.
These milestones collectively demonstrate DTT’s readiness to provide from the early 2030s the critical experimental data required to engineer the future of power exhaust.
[1] Francesco Romanelli et al 2024 Nucl. Fusion 64 112015
Wednesday 23 September 9:50 am
Commonwealth Fusion Systems (CFS) and its partners are transitioning the SPARC tokamak (B_T = 12.2 T, R_0 = 1.85 m, a = 0.57 m, designed to achieve Q ~ 11) from a collection of high-performance designs into an integrated, operational fusion device.
SPARC is currently under construction and assembly in Devens, Massachusetts. This presentation provides a status update on the delivery and installation of major fusion core components, including high-temperature superconducting (HTS) toroidal field coils, poloidal field coils, copper magnets, vacuum vessel, in-vessel components and diagnostic systems.
As the project shifts from component delivery to large-scale assembly, the team has encountered and resolved the real-world engineering complexities inherent in the first-of-a-kind integration of a compact, high-field device. This presentation discusses the technical solutions developed to overcome some of these challenges, ensuring structural, electrical and thermal requirements are met across the entire integrated machine and plant architecture.
Beyond physical assembly, CFS is executing a strategic, phased commissioning roadmap to bridge the gap toward first plasma. This begins with a partially integrated “dry” commissioning phase, or dry dress rehearsal (DDR), on track to complete in 2026. Unlike purely virtual testing, the DDR involves the active utilization of the actual plant and control systems, augmented by Hardware-in-the-Loop (HITL) simulations, to orchestrate physical plant subsystems (including cryogenics, power supplies, and vacuum systems) in a non-nuclear environment. This strategy is designed to retire infrastructure risks early, ensuring that the critical path to Q > 1 is dictated by the performance of the tokamak itself rather than its supporting systems. These operations leverage the in-house Neutrino framework, transitioning it from a developmental environment into its role as the machine’s unified control and data architecture.
By iteratively building upon the operational tools and frameworks established during the dry commissioning, CFS is establishing a repeatable pathway for fully integrated commissioning. This rigorous approach ensures the machine is prepared for its initial operational campaign, focusing on the milestone achievement of First Plasma and the ultimate scientific demonstration of Q > 1 net energy. Progress on SPARC remains one of the primary vehicles for informing the design and accelerated deployment of ARC, the first fusion power plant, in parallel to other R&D efforts.
Wednesday 23 September 8:30 am
