Abstract:JET underwent a transformation from a full carbon-dominated tokamak to a full metallic device with the ITER-like wall combination for the activated phase with Beryllium main chamber and Tungsten divertor. The ITER-Like Wall (ILW) experiment at JET provides an ideal test bed for ITER and shall demonstrate as primary goals the plasma compatibility with metallic walls and the reduction in fuel retention. We report on a set of experiments ( = 2.0 , = 2.0 − 2.4 , = 0.2 − 0.4) in different confinement and plasma conditions with global gas balance analysis demonstrating a strong reduction of the long term retention rate by a factor ten with respect to carbon references. All experiments have been executed in a series of identical plasma discharges in order to achieve maximum plasma duration until the analysis limit of the active gas handling system has been reached. The composition analysis shows high purity of the recovered gas, typically 99% D. For typical L-mode discharges ( = 0.5 ), type III ( = 5.0 ), and type I ELMy H-mode plasmas ( = 12.0 ) a drop of the retention rate normalised to the operational time in divertor configuration has been measured from 1.27 × 10 has been obtained with the ILW. The observed reduction by one order of magnitude confirms the expected predictions concerning the plasma-facing material change in ITER and widens the operation without active cleaning in the DT phase in comparison to a full carbon device.

Demonstrating improved confinement of energetic ions is one of the key goals of the Wendelstein 7-X (W7-X) stellarator. In the past campaigns, measuring confined fast ions has proven to be challenging. Future deuterium campaigns would open up the option of using fusion-produced neutrons to indirectly observe confined fast ions. There are two neutron populations: 2.45 MeV neutrons from thermonuclear and beam-target fusion, and 14.1 MeV neutrons from DT reactions between tritium fusion products and bulk deuterium. The 14.1 MeV neutron signal can be measured using a scintillating fiber neutron detector, whereas the overall neutron rate is monitored by common radiation safety detectors, for instance fission chambers. The fusion rates are dependent on the slowing-down distribution of the deuterium and tritium ions, which in turn depend on the magnetic configuration via fast ion orbits. In this work, we investigate the effect of magnetic configuration on neutron production rates in W7-X. The neutral beam injection, beam and triton slowing-down distributions, and the fusion reactivity are simulated with the ASCOT suite of codes. The results indicate that the magnetic configuration has only a small effect on the production of 2.45 MeV neutrons from DD fusion and, particularly, on the 14.1 MeV neutron production rates. Despite triton losses of up to 50 %, the amount of 14.1 MeV neutrons produced might be sufficient for a time-resolved detection using a scintillating fiber detector, although only in high-performance discharges.

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A comparison of the L–H power threshold (Pthr) in JET with all carbon, JET-C, and beryllium/tungsten wall (the ITER-like choice), JET-ILW, has been carried out in experiments with slow input power ramps and matched plasma shapes, divertor configuration and IP/BT pairs. The low density dependence of the L–H power threshold, namely an increase below a minimum density ne,min, which was first observed in JET with the MkII-GB divertor and C wall and subsequently not observed with the current MkII-HD geometry, is observed again with JET-ILW. At plasma densities above ne,min, Pthr is reduced by ∼30%, and by ∼40% when the radiation from the bulk plasma is subtracted (Psep), with JET-ILW compared to JET-C. At the L–H transition the electron temperature at the edge, where the pedestal later develops, is also lower with JET-ILW, for a given edge density. With JET-ILW the minimum density is found to increase roughly linearly with magnetic field, , while the power threshold at the minimum density scales as . The H-mode power threshold in JET-ILW is found to be sensitive both to variations in main plasma shape (Psep decreases with increasing lower triangularity and increases with upper triangularity) and in divertor configuration. When the data are recast in terms of Psep and Zeff or subdivertor neutral pressure a linear correlation is found, pointing to a possible role of Zeff and/or subdivertor neutral pressure in the L–H transition physics. Depending on the chosen divertor configuration, Pthr can be up to a factor of two lower than the ITPA scaling law for densities above ne,min. A shallow edge radial electric field well is observed at the L–H transition. The edge impurity ion poloidal velocity remains low, close to its L-mode values, ⩽5 km s−1 ± 2–3 km s−1, at the L–H transition and throughout the H-mode phase, with no measureable increase within the experimental uncertainties. The edge toroidal rotation profile does not contribute to the depth of the negative Er well and thus may not be correlated with the formation of the edge transport barrier in JET.

After completing the main construction phase of Wendelstein 7-X (W7-X) and successfully commissioning the device, first plasma operation started at the end of 2015. Integral commissioning of plasma start-up and operation using electron cyclotron resonance heating (ECRH) and an extensive set of plasma diagnostics have been completed, allowing initial physics studies during the first operational campaign. Both in helium and hydrogen, plasma breakdown was easily achieved. Gaining experience with plasma vessel conditioning, discharge lengths could be extended gradually. Eventually, discharges lasted up to 6 s, reaching an injected energy of 4 MJ, which is twice the limit originally agreed for the limiter configuration employed during the first operational campaign. At power levels of 4 MW central electron densities reached 3  ×  1019 m−3, central electron temperatures reached values of 7 keV and ion temperatures reached just above 2 keV. Important physics studies during this first operational phase include a first assessment of power balance and energy confinement, ECRH power deposition experiments, 2nd harmonic O-mode ECRH using multi-pass absorption, and current drive experiments using electron cyclotron current drive. As in many plasma discharges the electron temperature exceeds the ion temperature significantly, these plasmas are governed by core electron root confinement showing a strong positive electric field in the plasma centre.

The optimized, superconducting stellarator Wendelstein 7-X went into operation and delivered first measurement data after 15 years of construction and one year commissioning. Errors in the magnet assembly were confirmend to be small. Plasma operation was started with 5 MW electron cyclotron resonance heating (ECRH) power and five inboard limiters. Core plasma values of T 8 e > keV, T 2 i > keV at line-integrated densities n 3 10 m 19 2 ≈ ⋅ − were achieved, exceeding the original expectations by about a factor of two. Indications for a coreelectron-root were found. The energy confinement times are in line with the international stellarator scaling, despite unfavourable wall conditions, i.e. large areas of metal surfaces and particle sources from the limiter close to the plasma volume. Well controlled shorter hydrogen discharges at higher power (4 MW ECRH power for 1 s) and longer discharges at lower power (0.7 MW ECRH power for 6 s) could be routinely established after proper wall conditioning. The fairly large set of diagnostic systems running in the end of the 10 weeks operation campaign provided first insights into expected and unexpected physics of optimized stellarators.

The original goals of the JET ITER-like wall included the study of the impact of an all W divertor on plasma operation (Coenen et al 2013 Nucl. Fusion 53 073043) and fuel retention (Brezinsek et al 2013 Nucl. Fusion 53 083023). ITER has recently decided to install a full-tungsten (W) divertor from the start of operations. One of the key inputs required in support of this decision was the study of the possibility of W melting and melt splashing during transients. Damage of this type can lead to modifications of surface topology which could lead to higher disruption frequency or compromise subsequent plasma operation. Although every effort will be made to avoid leading edges, ITER plasma stored energies are sufficient that transients can drive shallow melting on the top surfaces of components. JET is able to produce ELMs large enough to allow access to transient melting in a regime of relevance to ITER. Transient W melt experiments were performed in JET using a dedicated divertor module and a sequence of IP = 3.0 MA/BT = 2.9 T H-mode pulses with an input power of PIN = 23 MW, a stored energy of ∼6 MJ and regular type I ELMs at ΔWELM = 0.3 MJ and fELM ∼ 30 Hz. By moving the outer strike point onto a dedicated leading edge in the W divertor the base temperature was raised within ∼1 s to a level allowing transient, ELM-driven melting during the subsequent 0.5 s. Such ELMs (δW ∼ 300 kJ per ELM) are comparable to mitigated ELMs expected in ITER (Pitts et al 2011 J. Nucl. Mater. 415 (Suppl.) S957–64). Although significant material losses in terms of ejections into the plasma were not observed, there is indirect evidence that some small droplets (∼80 µm) were released. Almost 1 mm (∼6 mm3) of W was moved by ∼150 ELMs within 7 subsequent discharges. The impact on the main plasma parameters was minor and no disruptions occurred. The W-melt gradually moved along the leading edge towards the high-field side, driven by j × B forces. The evaporation rate determined from spectroscopy is 100 times less than expected from steady state melting and is thus consistent only with transient melting during the individual ELMs. Analysis of IR data and spectroscopy together with modelling using the MEMOS code Bazylev et al 2009 J. Nucl. Mater. 390–391 810–13 point to transient melting as the main process. 3D MEMOS simulations on the consequences of multiple ELMs on damage of tungsten castellated armour have been performed. These experiments provide the first experimental evidence for the absence of significant melt splashing at transient events resembling mitigated ELMs on ITER and establish a key experimental benchmark for the MEMOS code.

The W-transport in the core plasma of JET is investigated experimentally by deriving the W-concentration profiles from the modelling of the signals of the soft x-ray cameras. For the case of pure neutral beam heating W accumulates in the core (r/a < 0.3) approaching W-concentrations of 10 −3 in between the sawtooth crashes, which flatten the W-profile to a concentration of about 3 × 10 −5 . When central Ion cyclotron resonant heating is additionally applied the core W-concentration decays in phases that exhibit a changed mode activity, while also the electron temperature increases and the density profile becomes less peaked. The immediate correlation between the change of magnetohydrodymanic (MHD) and the removal of W from the plasma core supports the hypothesis that the change of the MHD activity is the underlying cause for the change of transport. Furthermore, a discharge from the ASDEX Upgrade is investigated. In this case the plasma profiles exhibit small changes only, while the most prominent change occurs in the W-content of the confined plasma caused by the reduction of the fuelling deuterium gas puff. Concomintantly, the W-concentration profiles in the core plasma r/a < 0.2 steepen up reminescent to the well-known connection between central radiation and transport during cases with strong, established W-accumulation, while in the present analysis such a causality between the two during the onset of W-accumulation could not be pinned down. Both case studies exemplify that small changes of the core parameters of a plasma my influence the W-transport in the plasma core drastically.