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.

The recent deuterium–tritium campaign in JET-ILW (DTE2) has provided a unique opportunity to study the isotope dependence of the L-H power threshold in an ITER-like wall environment (Be wall and W divertor). Here we present results from dedicated L-H transition experiments at JET-ILW, documenting the power threshold in tritium and deuterium–tritium plasmas, comparing them with the matching deuterium and hydrogen datasets. From earlier experiments in JET-ILW it is known that as plasma isotopic composition changes from deuterium, through varying deuterium/hydrogen concentrations, to pure hydrogen, the value of the line averaged density at which the threshold is minimum, nˉe,min , increases, leading us to expect that nˉe,min (T) < nˉe,min (DT) < nˉe,min (D) < nˉe,min (H). The new power threshold data confirms these expectations in most cases, with the corresponding ordering of the minimum power thresholds. We present a comparison of this data to power threshold scalings, used for extrapolation to future devices such as ITER and DEMO.

We present an overview of results from a series of L–H transition experiments undertaken at JET since the installation of the ITER-like-wall (JET-ILW), with beryllium wall tiles and a tungsten divertor. Tritium, helium and deuterium plasmas have been investigated. Initial results in tritium show ohmic L–H transitions at low density and the power threshold for the L–H transition (P _LH) is lower in tritium plasmas than in deuterium ones at low densities, while we still lack contrasted data to provide a scaling at high densities. In helium plasmas there is a notable shift of the density at which the power threshold is minimum ( n ¯ e , min ) to higher values relative to deuterium and hydrogen references. Above n ¯ e , min (He) the L–H power threshold at high densities is similar for D and He plasmas. Transport modelling in slab geometry shows that in helium neoclassical transport competes with interchange-driven transport, unlike in hydrogen isotopes. Measurements of the radial electric field in deuterium plasmas show that E _r shear is not a good indicator of proximity to the L–H transition. Transport analysis of ion heat flux in deuterium plasmas show a non-linearity as density is decreased below n ¯ e , min . Lastly, a regression of the JET-ILW deuterium data is compared to the 2008 ITPA scaling law.

The development of a high current baseline scenario ( I p = 3.5 M A , q 95 ≈ 3.0 , β N < 2 ) in deuterium (D), tritium (T) and deuterium–tritium (D–T) for high D–T fusion performance at JET with Be/W wall is described. We show that a suitable scenario capable of delivering up to 10 MW of fusion power, depending on the auxiliary heating power available, was successfully developed in D. However, when translated to T and D–T, the same scenario could not be sustained for the target duration of 5 s due to the impossibility to achieve the stationary compound edge localized modes regime necessary to flush the tungsten (W) from the plasma and control the density. Nevertheless, a peak fusion power in the order of 8 MW, with 60% of the power coming from thermal fusion reactions, was obtained in D–T at 3.5 MA, with ≈ 30 MW of neutral beam injection heating and 3–4 MW of ion cyclotron resonance heating, in line with the predictions obtained with the JINTRAC integrated scenario modelling suite of codes equipped with the QuaLiKiZ transport model and based on the extrapolation of the performance of similar D plasmas.

The experimental and theoretical analysis were focused on experiments conducted to assess the effect of plasma isotopes, protium (H), deuterium (D), and tritium (T) on ion cyclotron resonance heating (ICRH) related plasma wall interactions. Comparison of L-mode discharges with N = 1 ³He and N = 1 H minority ICRH heating scenarios were done for different isotopes. For the selected pulses, the behaviour of high-Z, mid-Z and low-Z intrinsic impurity and radiated power behaviour was investigated based on data from VUV, visible spectroscopy, and bolometry diagnostic at Joint European Torus. It was found that for N = 1 ³He scenario during radiofrequency antennas operation, core W, Ni content, Be source and the radiated power are higher for π/2 in comparison to dipole antenna phasing. Lowest core Ni, W content and radiated power is clearly observed for H plasmas in comparison to D and T, where for this ICRH scenario behaviour was similar. However, lower Be photon flux is observed for T in comparison to D plasmas. Be sputtering by He particles is responsible for such an effect. Additionally, several computer simulations were conducted using the COREDIV code. The difference in the electron temperature was due to the difference in the isotope masses. Increased temperature in the central plasma in the case of T plasmas leads to higher radiation in the central plasma in comparison to H plasmas. As a result, the power across separatrix is lower and the temperature on the divertor plate decreases with the increase of the isotope mass. At these temperatures on the divertor plate, W is not sputtered by the main plasma ions H, D and T and by He. For the N = 1 H ICRH scenario clear difference between D and T plasma was observed with higher metallic impurity content for T plasma in comparison to D. Impurity content in the plasmas is found to be sensitive to the power balance between the antenna straps. Its minimum is observed for the maximum of P _cen/P _tot.

Tunable diode laser absorption spectroscopy (TDLAS) is a commonly used technique to measure the temperature and density of atoms or molecules in a gas. In this work, we demonstrate that the TDLAS diagnostics could be effectively applied to measure the magnetic field in a low-density weakly magnetized plasma using the Zeeman splitting of the absorption spectrum of lines from noble gases. The laser wavelength is tailored to fit the $1s_5\to 2p_6$ transition of atomic Ar with the wavelength $\lambda =763.51\mathrm{nm}$ . Two mechanisms of line broadening and splitting are observed: Doppler broadening and Zeeman effect. The latter is especially pronounced by applying polarization-selective observation of the absorption to the TDLAS measurements. By fitting the $\sigma$ and $\pi$ components of the absorption spectrum, the line-integrated magnetic field on the order of 30–50 mT is determined. The agreement between the measured values and the vacuum field (neglecting the impact of the plasma) calculations on the axis of the PSI-2 is found to be about 15–20%.