Abstract The two best performing pulses of the so called ITER-Baseline scenario ( I p = 3.5 MA and P in ≈ 35 MW) of JET-ITER like wall, one in deuterium (D) the other in deuterium–tritium (D–T) ...plasma are examined and compared in this study. Generally, the D–T Baseline pulses exhibit an electron density level higher than the D pulses and the plasma energy is higher than in the comparable D pulses by up to 20%, reaching about 12 MJ in the pulse studied here. In contrast with the D pulses, the D–T pulses are often characterised by the increase in time of the radiated power in the mantle region (0.70 < ρ < 0.95), which may lead to the loss of the edge localised mode activity when the threshold H–L transition power is approached and to the subsequent plasma disruption due to excessive radiation. In this study we try to identify the physical mechanisms responsible for this behaviour using the available experimental data (principally the total radiated power from the bolometry) and the results of the fluid COREDIV model (1D in the core, 2D in the scrape-off-layer (SOL)), self-consistent with respect to core-SOL and also to main plasma-impurities. In fact, the loss of power caused by impurity radiation affects the temperature profile and finally the power to the divertor plate. The electron density and temperature profiles are numerically reconstructed as well as the radiated power density profiles, indicating no major difference in impurity transport in D and D–T. Indeed, the impurity transport coefficients used in COREDIV to match the experimental radiated power profiles are similar in the two pulses. The computed tungsten sources and densities are lower in the D–T pulse and the divertor impurity retention capability is a little better in the D–T pulse, indicatinga stronger collisional drag force in the SOL. The higher electron density and the broadening of its profile are the main cause of the observed increase of the radiated power in the D–T pulse.
JET is used as a test bed for ITER, to investigate beryllium migration which connects the lifetime of first-wall components under erosion with tokamak safety, in relation to long-term fuel retention. ...The (i) limiter and the (ii) divertor configurations have been studied in JET-ILW (JET with a Be first wall and W divertor), and compared with those for the former JET-C (JET with carbon-based plasma-facing components (PFCs)). (i) For the limiter configuration, the Be gross erosion at the contact point was determined in situ by spectroscopy as between 4% (Ein = 35 eV) and more than 100%, caused by Be self-sputtering (Ein = 200 eV). Chemically assisted physical sputtering via BeD release has been identified to contribute to the effective Be sputtering yield, i.e. at Ein = 75 eV, erosion was enhanced by about 1/3 with respect to the bare physical sputtering case. An effective gross yield of 10% is on average representative for limiter plasma conditions, whereas a factor of 2 difference between the gross erosion and net erosion, determined by post-mortem analysis, was found. The primary impurity source in the limiter configuration in JET-ILW is only 25% higher (in weight) than that for the JET-C case. The main fraction of eroded Be stays within the main chamber. (ii) For the divertor configuration, neutral Be and BeD from physically and chemically assisted physical sputtering by charge exchange neutrals and residual ion flux at the recessed wall enter the plasma, ionize and are transported by scrape-off layer flows towards the inner divertor where significant net deposition takes place. The amount of Be eroded at the first wall (21 g) and the Be amount deposited in the inner divertor (28 g) are in fair agreement, though the balancing is as yet incomplete due to the limited analysis of PFCs. The primary impurity source in the JET-ILW is a factor of 5.3 less in comparison with that for JET-C, resulting in lower divertor material deposition, by more than one order of magnitude. Within the divertor, Be performs far fewer re-erosion and transport steps than C due to an energetic threshold for Be sputtering, and inhibits as a result of this the transport to the divertor floor and the pump duct entrance. The target plates in the JET-ILW inner divertor represent at the strike line a permanent net erosion zone, in contrast to the net deposition zone in JET-C with thick carbon deposits on the CFC (carbon-fibre composite) plates. The Be migration identified is consistent with the observed low long-term fuel retention and dust production with the JET-ILW.
Abstract
Unusually high power loads on the beryllium limiter caused by neutral beam re-ionisation, and much cooler divertor target surfaces were observed during the recent JET tokamak tritium ...campaign. As both phenomena are driven by scrape-off layer (SOL) physics, the SOL features of 72 tritium H-mode discharges and their deuterium references have been studied. The majority (70) of tritium H-mode discharges had exponentially decaying SOL profiles. The tritium plasmas are observed to have increased separatrix density and collisionality compared to their deuterium references. This is associated with
≈
2
−
3
times broader SOL width for both density and temperature profiles. This is consistent with previous observations in highly collisional deuterium H-mode plasma on the ASDEX Upgrade tokamak (Sun
et al
2015
Plasma
Phys.
Control.
Fusion
57
125011) and interpreted as high collisionality enhancing cross-field transport across the separatrix and resulting in the broadening of near SOL above a critical value. The other two tritium H-mode discharges had near flat SOL density profiles, similar to the so-called ‘
density shoulder formation
’ observed in L-mode plasma. The SOL collisionality of these two pulses lies within the range of T pulses without density shoulder formation. This supports the conclusion of previous studies (Vianello
et al
2017
Nucl. Fusion
57
116014; Wynn
et al
2018
Nucl. Fusion
58
056001) that increased collisionality is not sufficient for the formation of a ‘
density shoulder
’ and additional factors, likely divertor condition or interaction with neutrals, are required. JET tritium plasma provides evidence of favourable and unfavourable effects of enhanced cross-field SOL transport on machine operation. The larger limiter power loads due to re-ionisation of neutral beam injection observed in the T pulses relative to their D references has been shown to be consistent with the combined effects of the broadening of the SOL profile and larger beam ion Larmor radius. The enhanced cross-field particle transport and the resulting broader SOL width provides more particles to ionize the fast Beam neutrals, causing the unfavourable power load issue on the beryllium limiter. The broader near SOL profiles of the T plasma spreads the heat load over a larger area and, together with the increased separatrix density, results in a favourably cooler divertor target surface.
Abstract
The reference ion cyclotron resonance frequency (ICRF) heating schemes for ITER deuterium–tritium (D-T) plasmas at the full magnetic field of 5.3 T are second harmonic heating of T and
3
He ...minority heating. The wave-particle resonance location for these schemes coincide and are central at a wave frequency of 53 MHz at 5.3 T. Experiments have been carried out in the second major D-T campaign (DTE2) at JET, and in its prior D campaigns, to integrate these ICRF scenarios in JET high-performance plasmas and to compare their performance with the commonly used hydrogen (H) minority heating. In 50:50 D:T plasmas, up to 35% and 5% larger fusion power and diamagnetic energy content, respectively, were obtained with second harmonic heating of T as compared to H minority heating at comparable total input powers and gas injection rates. The core ion temperature was up to 30% and 20% higher with second harmonic T and
3
He minority heating, respectively, with respect to H minority heating. These are favourable results for the use of these scenarios in ITER and future fusion reactors. According to modelling, adding ICRF heating to neutral beam injection using D and T beams resulted in a 10%–20% increase of on-axis bulk ion heating in the D-T plasmas due to its localisation in the plasma core. Central power deposition was confirmed with the break-in-slope and fast Fourier transform analysis of ion and electron temperature in response to ICRF modulation. The tail temperature of fast ICRF-accelerated tritons, their enhancement of the fusion yield and time behaviour as measured by an upgraded magnetic proton recoil spectrometer and neutral particle analyser were found in agreement with theoretical predictions. No losses of ICRF-accelerated ions were observed by fast ion detectors, which was as expected given the high plasma density of
n
e
≈ 7–8 × 10
19
m
−3
in the main heating phase that limited the formation of ICRF-accelerated fast ion tails.
3
He was introduced in the machine by
3
He gas injection, and the
3
He concentration was measured by a high-resolution optical penning gauge in the sub-divertor region. The DTE2 experiments with
3
He minority heating were carried with a low
3
He concentration in the range of 2%–4% given the fact that the highest neutron rates with
3
He minority heating in D plasmas were obtained at low
3
He concentrations of ∼2%, which also coincided with the highest plasma diamagnetic energy content. In addition to
3
He introduced by
3
He gas injection, an intrinsic concentration of
3
He of the order of 0.2%–0.4% was measured in D-T plasmas before
3
He was introduced in the device, which is attributed to the radioactive decay of tritium to
3
He. According to modelling, even such low intrinsic concentrations of
3
He lead to significant changes in ICRF power partitioning during second harmonic heating of T due to absorption of up to 30% of the wave power by
3
He.
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