Xavier Litaudon

A first-of-its-kind experiment was conducted in the DIII-D tokamak under WEST similarity constraints on plasma shape and core parameters. This work presents a detailed transport study comparing a reference regime dominated by intrinsic carbon radiation and a high-radiation regime resulting from controlled tungsten (W) injection using the laser blow-off system, with a core tungsten concentration n W / n e ∼ 3 × 10 − 4 and a radiated-power fraction f rad > 0.5 . The W-induced radiative cooling lowered the electron temperature, thereby decreasing T e / T i and stabilizing trapped-electron-mode (TEM) turbulence. This transition in turbulence regime reduced momentum and ion thermal diffusivities, yielding ion temperature peaking and a factor-of-two increase in toroidal rotation. At the outer plasma region, enhanced E × B shear and increased collisionality further suppressed ion-scale turbulence, causing a sharp drop in ion heat flux. Consequently, impurity transport, predominantly turbulent in the low-radiation regime, acquired a strong neoclassical inward W convection during radiative cooling, bootstrapping the cooling cycle. Despite f rad > 0.5 , radiative collapse was not observed, likely owing to collisional ion-to-electron energy exchange acting as an electron-energy reservoir, together with 1 / 1 MHD activity modulating the radiated power through core impurity neoclassical T i -screening. These results support preparation for a tungsten wall change in DIII-D by elucidating tungsten-induced turbulence stabilization. They also provide key insights for interpreting plasma performance in WEST and are relevant to future reactors expected to operate with radiating tungsten-walled plasmas.

Long pulse operation in magnetic fusion devices requires well controlled plasma power exhaust to the divertor & wall, and avoidance of wall hot spots that could evolve in wall damage. At WEST, 10 major plasma facing components are monitored using 10 series of temperature/power indicators, based on multiple diagnostic systems, among which the infrared viewing system is especially relevant. These indicators span from the most basic ones (temperatures, power and energy from deterministic models) to advanced processes using artificial intelligence acquired through machine learning. Some advanced processes do operate in real time, and feedback on power actuators through the plasma control system, providing active control toward remaining within the safe operational domain. Other advanced processes intervene as forensic tools post discharge to identify possible dangerous situations regarding the power loading to the wall, so that the discharge run plan is adjusted to avoid running into aggravating wall events. No critical wall power event happened during the campaigns C9 to C11 (2024–2025), totalling about 13 h of plasma, that would have affected the campaign plan. While it cannot be demonstrated that the active & intelligent wall protection enabled the new plasma duration record of 1337 s, the wall protection system as a whole plausibly contributed to obtaining these record durations by preventing wall hot spots to become critical during the campaigns.

Within the framework of EUROfusion’s technological exploitation of deuterium–tritium (DT) operation at JET—initiated under work package JET3 and continued under PrIO—the NEXP sub-project was conceived to exploit the substantial neutron yield expected during the DT campaigns (DTE2 and DTE3) to validate the numerical tools employed for neutron streaming and shutdown dose rate (SDDR) prediction in ITER. This validation, grounded in direct comparison between computational predictions and experimental measurements, constitutes a critical step in demonstrating the reliability of the SDDR methodology that underpins ITER’s radiation safety case. As part of this activity, a dedicated dosimetry system based on spherical air-vented ionization chambers was installed in the JET Torus Hall in 2015 and has since undergone successive hardware and software upgrades to address operational challenges identified across DT experimental campaigns. This work traces the evolution of the SDDR dosimetry system and presents the analysis of the measurements acquired during the DT operation of JET, in particular the third DT campaign (DTE3) and, uniquely, throughout the ongoing final shutdown, a phase offering measurement conditions unlikely to be replicated elsewhere. SDDR predictions with the Advanced D1S tool by ENEA are described as well.

We measured single event effects (SEEs) in 40 nm and 65 nm Static Random Access Memories (SRAMs) exposed to the D–T fusion neutron environment of the JET tokamak during the DTE-3 campaign (September–October 2023), while also characterizing the neutron spectrum seen by the devices. Using consolidated JET neutron production data, the measured SEE rates show good agreement with predictions based on the measured spectra and independently determined SRAM sensitivity parameters from thermal and monoenergetic neutron tests. These results extend the validation of our spectrum-based SEE prediction methodology, previously demonstrated in a D–D fusion neutron environment in 2021, to a D–T fusion neutron field. The novelty of this work is therefore the first experimental validation of this complete prediction approach in a D–T fusion environment, using real-time neutron spectrometry and independent device sensitivity data, rather than a new SEE model. The experiments also show that a local B ₄ C shield reduces the bit-flip rate by absorbing thermal neutrons, and we discuss the conditions under which such shielding can improve electronics reliability in tokamak and accelerator neutron environments.

The WEST tokamak is equipped with a superconducting toroidal magnetic field system, a multi-megawatt radiofrequency auxiliary power system, and an actively cooled ITER-grade tungsten divertor. As such, it is well adapted to explore experimental aspects related to the long pulse operation of next-step devices. Supported by predict-first integrated modeling, bespoke scenario development has allowed zero-loop voltage pulses to be achieved. The resulting discharges, with plasma currents in the range I p ∼ 0.22 − 0.28 MA exclusively sustained by the lower hybrid current drive (LHCD) system as an auxiliary power source, have achieved durations in excess of 22 min and injected/extracted energies up to 2.61 GJ. Plasma performance is characterized by ranges of poloidal beta β p ∼ 1.6 − 2.0 , normalized toroidal beta β N ∼ 0.6 − 0.9 and confinement factor H 96 L = 1 − 1.3 . Mild MHD activity, identified as resulting from the interaction of 3/1 and 4/1 tearing modes, is occasionally present, depending on the LHCD antenna combination used. This article describes the predict-first approach thathas been employed in the context of this long-pulse scenario development endeavor. The main achievements and the physics analyses performed are reviewed, including post-experiment integrated modeling aspects. Prospects for further long-pulse developments are drawn.

Combined high-fusion performance and long-pulse operation is one of the key integration challenges for fusion energy development in magnetic devices. Addressing these challenges requires an integrated vision of physics and engineering aspects with the purpose of simultaneously increasing time duration and fusion performance. Since the previous 2023 IAEA Fusion Energy Conference (FEC), significant progresses have been made in tokamaks and stellarators including very recent achievement in duration and/or performance. These progresses are reviewed by analysing the experimental data provided by 10 tokamaks and two stellarators. The published database (Litaudon et al 2024 Nucl. Fusion 64 015001), which initially included data up to January 2022, has been significantly updated for the 2025 IAEA FEC to incorporate the latest 2023–2025 experiments (up to May 2025) including recent records performance with new entries provided by DIII-D, EAST, JET, KSTAR, WEST, and W7-X. The update dataset has been gathered and coordination have been provided by the IEA-IAEA international CICLOP group (Coordination on International Challenges on Long duration OPeration). An overview of the recent progress toward long pulse operation analysing the CICLOP database is provided in this publication.

Short-lived activation products in fusion materials are critical for assessing radiological safety and maintenance planning in future fusion reactors like ITER. These products influence shutdown dose rates, impact maintenance procedures, and provide essential validation for nuclear data and activation models in the short term. Understanding their formation and decay behaviour under deuterium–tritium (DT) plasma conditions allows for more accurate predictions of material performance and safety constraints. This study analyses the short-term activation of ITER materials exposed at the KN2 6U irradiation end in the JET (Joint European Torus) tokamak during the DTE2 and DTE3 campaigns. CuCrZr alloy, the iron–nickel–chromium alloy A286 used for the ITER in-wall shield, and SS316L(N) stainless samples were activated during DTE2, while tungsten monoblock, a key ITER divertor material, was irradiated during DTE3. Activated samples were promptly transported via a pneumatic post system for gamma spectrometry measurements using UKAEA’s 190% relative efficiency high-purity germanium detector. The efficiency calibration was performed with the MCNP 6.1 code, employing an MCNP detector model validated by a certified multi-gamma calibration source. The experimental data were compared with FISPACT-II activation calculations, enabling the determination of calculation-to-experiment ( C / E ) ratios. This study provides key validation data for nuclear activation models and improves the accuracy of computational predictions for material behaviour in fusion reactors.

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.

For the first time, similarity experiments between DIII-D and WEST were performed in the ITER "hybrid-like" regime during dedicated campaigns in April and May 2025. The matched parameters include elongation, triangularity, ion ∇B drift direction toward the X-point, qprofile, and core normalized physics quantities in terms of normalized pressure, normalized gyroradius, electron collisionality, ratio of ion to electron temperature, Ti/Te. Core transport physics is explored with different aspect ratio (R/a) values (typically 3 at DIII-D and 5 on WEST). DIII-D explored high-beta conditions (electromagnetic effect) with low torque injection (~0 ± 0.5 N•m) using high heating power (up to 6 MW NBI and 2 MW ECRH powers), while scanning the heating mix (ion vs electron), beta, Ti/Te, core radiation via controlled tungsten injection using the Laser Blow-Off system. WEST extended operation toward long-duration pulses using its actively cooled tungsten divertor, achieving dominated electron heating regimes with reduced tungsten contamination. Boron impurity injection were scanned on WEST to control edge conditions and core performance. It is found that core confinement improves-manifested by higher electron temperature, total energy content, neutron rate, and ion temperatureunder conditions of low separatrix density, consistent with previous observations [Bourdelle et al., Nucl. Fusion 63 (2023) 056021]. Conditions for Hmode access and for ion heating in electron-dominated regimes in both WEST and DIII-D will be discussed and compared. The ratio of the thermal energy confinement time (τE) to the volume-averaged electron-ion collisional heat exchange time (τe-i) is a key parameter to enhance ion heating and potentially facilitate H-mode access in electron-heated regimes. These first-of-a-kind coordinated DIII-D and WEST experiments provide a unique multi-machine dataset to validate predictive models and to optimize ITER hybrid-scenario performance under diverse core and edge conditions. * βN = βt x (Ip/aBT) -1 [%] where βt = 2μ0 /BT 2 , is the volume averaged total plasma pressure, Bt the toroidal field on axis in Tesla [T], a the minor radius [m] , Ip the plasma current [MA]

As the primary coolant, water is used in most current fission reactors and is also a promising coolant for future fusion devices. During normal operation, the fusion device produces plasma discharges. The primary cooling water circulates through in-vessel cooling channels and thus passes through regions of high neutron flux ( E ∼ 14 MeV ) produced by D–T fusion. During this process, cooling water can be activated and produce radioactive nuclides that emit gamma rays and neutrons (neutron energy range 0.4–1.7 MeV). Predicting the dose-rate field—the spatial and temporal distribution of radiation dose rates ( Sv ⋅ h − 1 ) produced by activated nuclides in the flowing coolant—requires coupling computational fluid dynamics (CFD) with radiation-transport and activation calculations. To validate these coupled CFD + radiation-transport + activation models and to investigate how production, transport and decay of activation products together create the observed dose-rate field, the KATANA water-activation loop at the Jožef Stefan Institute Training, Research, Isotope production, General Atomics Mark II reactor was commissioned. This paper examines the newly established KATANA water-activation loop and evaluates the feasibility of using the neutron activation method for neutron measurements at the facility. The studies show that this method can be successfully applied to detect and quantify neutrons originating from 17 N decay. Furthermore, the measurements demonstrate that this technique is capable of characterising the neutron emission rate at the KATANA water-activation loop. Finally, the paper proposes a method for estimating the neutron fluence at selected sample positions. The results encourage further use of the neutron activation method to investigate neutrons emitted by activated water. Future experiments could determine the neutron fluence-rate profile, thereby validating fluid-activation codes for ITER, DEMO and other fusion devices. This would also contribute to reducing the uncertainty of the 17 O ( n , p ) 17 N reaction cross-section.