Laser-induced breakdown spectroscopy (LIBS) is one of the most promising methods for quantitative in-situ determination of fuel retention in plasma-facing components (PFCs) of magnetically confined fusion devices like ITER and JET. In this article, the current state of understanding in LIBS development for fusion applications will be presented, based on a complete review of existing results and complemented with newly obtained data. The work has been performed as part of a research programme, set up in the EUROfusion Consortium, to address the main requirements for ITER: (a) quantification of fuel from relevant surfaces with high sensitivity, (b) the technical demonstration to perform LIBS with a remote handling system and (c) accurate detection of fuel at ambient pressures relevant for ITER. For the first goal, the elemental composition of ITER-like deposits and proxies to them, including deuterium (D) or helium (He) containing W–Be, W, W–Al and Be–O–C coatings, was successfully determined with a typical depth resolution ranging from 50 up to 250 nm per laser pulse. Deuterium was used as a substitute for tritium (T) and in the LIBS experiments deuterium surface densities below 10¹⁶ D/cm² could be measured with an accuracy of ∼30%, confirming the required high sensitivity for fuel-retention investigations. The performance of different LIBS configurations was explored, comprising LIBS systems based on single pulse (pulse durations: ps–ns) and double pulse lasers with different pulse durations. For the second goal, a remote handling application was demonstrated inside the Frascati-Tokamak-Upgrade (FTU), where a compact, remotely controlled LIBS system was mounted on a multipurpose deployer providing an in-vessel retention monitor system. During a shutdown phase, LIBS was performed at atmospheric pressure, for measuring the composition and fuel content of different area of the stainless-steel FTU first wall, and the titanium zirconium molybdenum alloy tiles of the toroidal limiter. These achievements underline the capability of a LIBS-based retention monitor, which complies with the requirements for JET and ITER operating in DT with a beryllium wall and a tungsten divertor. Concerning the capabilities of LIBS at pressure conditions relevant for ITER, quantitative determination of the composition of PFC materials at ambient pressures up to 100 mbar of N₂, the D content could be determined with an accuracy of 25%, while for atmospheric pressure conditions, an accuracy of about 50% was found when using single-pulse lasers. To improve the LIBS performance in atmospheric pressure conditions, a novel approach is proposed for quantitative determination of the retained T and the D/T ratio. This scenario is based on measuring the LIBS plume emission at two different time delays after each laser pulse. On virtue of application of a double pulse LIBS system, for LIBS application at N₂ atmospheric pressure the distinguishability of the spectra from H isotopes could be significantly improved, but further systematic research is required.

Since the publication of the review Progress in the ITER Physics Basis (PIPB) in 2007, significant progress has been made in understanding the processes at the plasma-material interface. This review, part of the ITPA Nuclear Fusion Special Issue On the Path to Burning Plasma Operation, presents these developments, focusing on key areas such as the physics of plasma exhaust, plasma-material interactions, and the properties of plasma-facing materials and their evolution under plasma exposure. The coordinated efforts of the ITPA Topical Group on Scrape-Off Layer and Divertor Physics (DivSOL) have been instrumental in identifying and addressing critical research and development issues in numerous collaborative experimental and modelling projects.

ITER will operate with a tungsten divertor, a material featuring surface morphology changes when exposed to helium plasmas, in particular the formation of the so called tungsten fuzz under specific conditions. Investigating interactions between tungsten plasma facing components and helium plasmas in a tokamak environment is therefore a key point to consolidate predictions for the ITER divertor performance and lifetime. To this end, a dedicated helium campaign was performed in the full tungsten WEST tokamak, cumulating ∼2000 s of repetitive L mode discharges. It is shown that conditions for tungsten fuzz formation, as derived from linear devices experiments (incident helium energy E _inc > 20 eV, helium fluence >10²⁴ He/m², surface temperature T _surf > 700 °C), were met in the outer strike point (OSP) area of the inertially cooled tungsten divertor elements of WEST. Preliminary inspection of the components after the campaign did not show visible signs of surface modification, but points to significant net erosion in the OSP area. An extensive post mortem analysis is now ongoing to confirm these first indications. These results underline that the complex balance between erosion/redeposition (in particular linked to impurities) and tungsten fuzz formation needs to be taken into account in tokamak conditions.

JET returned to deuterium-tritium operations in 2023 (DTE3 campaign), approximately two years after DTE2. DTE3 was designed as an extension of JET’s 2022-2023 deuterium campaigns, which focused on developing scenarios for ITER and DEMO, integrating in-depth physics understanding and control schemes. These scenarios were evaluated with mixed D-T fuel, using the only remaining tritium-capable tokamak until its closure in 2023. A core-edge-SOL integrated H-mode scenario was developed and tested in D-T, showing good confinement and partial divertor detachment with Ne-seeding. Stationary pulses with good performance, no tungsten accumulation, and even without ELMs were achieved in D-T. Plasmas with pedestals limited by peeling modes were studied with D, T-rich, and D-T fuel, revealing a positive correlation between pedestal electron pressure and pedestal electron density. The Quasi-Continuous Exhaust regime was successfully achieved with D-T fuel, with access criteria similar to those in D plasmas. A scenario with full detachment, the X-point radiator regime, was established in D-T, aided by the real-time control of the radiator’s position. The crucial characterisation of tritium retention continued in DTE3, using gas balance measurements and the new LID-QMS diagnostic. Nuclear technology studies were advanced during the DTE3 campaign, addressing issues such as the activation of water in cooling loops and single event effects on electronics. Building on the previous D, T and DTE2 campaigns and the lessons learned from them, DTE3 extended our understanding of D-T plasmas, particularly in scenarios relevant to next-generation devices such as ITER and DEMO.

The net erosion and deposition patterns in the inner and outer divertor of WEST were determined after different experimental campaigns (C3 and C4) of the first operational phase using ion beam analyses and scanning electron microscopy techniques. The analyses were performed on four entire tiles from inertially cooled, W-coated divertor units with an additional Mo marker coating covered with a further W coating. Strong erosion occurred at the expected location of the inner and outer strike line area with a campaign-averaged net erosion rate of >0.1 nm s⁻¹. On the high field side of the inner strike line area, thick deposited layers were found (>10 μm; growth rate >1 nm s⁻¹), mainly composed of B, C, O, and W. Additionally, strong arcing was observed in this region. At the end of the C4 campaign, He discharges were performed to study the He-W interaction. Although the conditions for nanotendrils, i.e. fuzz formation were fulfilled around the outer strike line position, neither nanotendrils nor He bubbles (>10 nm) were observed at this area.