S. Brezinsek

Beryllium Migration in JET ITER-like Wall Plasmas

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.

Moderation of divertor heat loads by fuelling and impurity seeding in well-confined ELMy H-mode plasmas on JET

In anticipation of revised operational limits accompanying upgrade of the JET tokamak to an all-metal, ITER-like wall, systematic studies have been conducted of moderating divertor heat load in type I ELMy H-mode by enhanced deuterium fuelling and seeding with the extrinsic impurities neon or nitrogen. These were found to differ mainly in the roughly ten times higher level of nitrogen than neon presently required to approach outboard detachment, and in their effects upon ELMs. With either species up to input power of ?17?MW, strongly reduced target loads between ELMs, compatible with the new wall, were sustained for longer than 10 energy confinement times while losing <10% of normalized energy confinement and keeping density above 90% of the Greenwald value. At the highest nitrogen levels, ELMs were also markedly decreased in size, yielding a regime still with reasonable confinement plus attenuated exhaust both between and during ELMs.

Development of steady-state scenarios compatible with ITER-like wall conditions

A key issue for steady-state tokamak operation is to determine the edge conditions that are compatible both with good core confinement and with the power handling and plasma exhaust capabilities of the plasma facing components (PFCs) and divertor systems. A quantitative response to this open question will provide a robust scientific basis for reliable extrapolation

Measurements of plasma composition in the TEXTOR tokamak by collective Thomson scattering

We demonstrate the use of collective Thomson scattering (CTS) for spatially localized measurements of the isotopic composition of magnetically confined fusion plasmas. The experiments were conducted in the TEXTOR tokamak by scattering millimeter-wave probe radiation off plasma fluctuations with wave vector components nearly perpendicular to the magnetic field. Under such conditions the sensitivity of the CTS spectrum to plasma composition is enhanced by the spectral signatures of the ion cyclotron motion and of weakly damped ion Bernstein waves. Recent experiments on TEXTOR demonstrated the ability to resolve these signatures in the CTS spectrum as well as their sensitivity to the ion species mix in the plasma. This paper shows that the plasma composition can be inferred from the measurements through forward modeling of the CTS spectrum. We demonstrate that spectra measured in plasmas consisting of hydrogen, deuterium and 3He can be accurately reproduced by theory and yield inferred plasma compositions consistent with expectations. The potential to use CTS for measurements of plasma composition is of significant interest since CTS is well suited for reactor environments and since there is at present no established method to measure the fuel ion density ratio in the core of a burning fusion plasma.

Comparison of H-mode Plasmas in JET-ILW and JET-C with and without Nitrogen Seeding

In high confinement mode, highly shaped plasmas with edge localized modes in JET, and for heating power of 15–17 MW, the edge fluid code EDGE2D-EIRENE predicts transition to detachment assisted by nitrogen at the low field side (LFS) target when more than 50% of the power crossing the separatrix between ELMs is radiated in the divertor chamber, i.e. ~4 MW. This is observed both in the ITER-like wall (JET-ILW) and in the carbon wall (JET-C) configurations and is consistent with experimental observations within their uncertainty. In these conditions, peak heat fluxes below 1 MW m−2 are measured at the LFS target and predicted for both wall configurations. When the JET-C configuration is replaced with the JET-ILW, a factor of two reduction in the divertor radiated power and 25–50% increase in the peak and total power deposited to the LFS divertor plate is predicted by EDGE2D-EIRENE for unseeded plasmas similar to experimental observations. At the detachment threshold, EDGE2D-EIRENE shows that nitrogen radiates more than 80% of the total divertor radiation in JET-ILW with beryllium contributing less than a few %. With JET-C, nitrogen radiates more than 70% with carbon providing less than 20% of the total radiation. Therefore, the lower intrinsic divertor radiation with JET-ILW is compensated by stronger nitrogen radiation contribution in simulations leading to detachment at similar total divertor radiation fractions. 20–100% higher deuterium molecular fraction in the divertor recycling fluxes is predicted with light JET-C materials when compared to heavy tungsten. EDGE2D-EIRENE simulations indicate that the stronger molecular contribution can reduce the divertor peak power deposition in high recycling conditions by 10–20% due to enhanced power dissipation by molecular interaction.

Gas analyses of the first complete JET cryopump regeneration with ITER-like wall

Analytical results of a complete JET cryopump regeneration, including the nitrogen panel, following the first ITER-like wall campaign are presented along with the in situ analyses of residual gas. H/D mixtures and impurities such as nitrogen and neon were injected during plasma operation in the vessel to study radiation cooling in the scrape-off-layer and divertor region. The global gas inventory over the campaign is incomplete, suggesting residual volatile impurities are remaining on the cryogenic panel. This paper presents results on (i) residual deuterium on the panel which is very loosely related to the campaign, (ii) impurities like nitrogen which stick on the panel, and (iii) the ammonia production which can be observed by mass spectrometry.

JET divertor diagnostic upgrade for neutral gas analysis.

With installation of the ITER-like wall in JET a major diagnostic upgrade to measure the neutral gas pressure and composition in the sub-divertor region has been completed, to characterise retention and outgassing of the new metallic first wall. The upgrade includes two new magnetically shielded systems consisting of sensitive capacitance manometers and residual gas analysers, both capable of providing data during plasma operation. These enable absolute pressure and gas composition measurements (pressure range: 10(-5)-10(-1) mbar, mass range: 1-200 amu, respectively) and have been used to characterise the neutral gas behaviour under various plasma conditions.

Impact of divertor geometry on radiative divertor performance in JET H-mode plasmas

Radiative divertor operation in JET high confinement mode plasmas with the ITER-like wall has been experimentally investigated and simulated with EDGE2D-EIRENE in horizontal and vertical low field side (LFS) divertor configurations. The simulations show that the LFS divertor heat fluxes are reduced with N2-injection in similar fashion in both configurations, qualitatively consistent with experimental observations. The simulations show no substantial difference between the two configurations in the reduction of the peak LFS heat flux as a function of divertor radiation, nitrogen concentration, or pedestal Zeff. Consistently, experiments show similar divertor radiation and nitrogen injection levels for similar LFS peak heat flux reduction in both configurations. Nevertheless, the LFS strike point is predicted to detach at 20% lower separatrix density in the vertical than in the horizontal configuration. However, since the peak LFS heat flux in partial detachment in the vertical configurations is shifted towards the far scrape-off layer (SOL), the simulations predict no benefit in the reduction of LFS peak heat flux for a given upstream density in the vertical configuration relative to a horizontal one. A factor of 2 reduction of deuterium ionization source inside the separatrix is observed in the simulations when changing to the vertical configuration. The simulations capture the experimentally observed particle and heat flux reduction at the LFS divertor plate in both configurations, when adjusting the impurity injection rate to reproduce the measured divertor radiation. However, the divertor D α -emissions are underestimated by a factor of 2–5, indicating a short-fall in radiation by the fuel species. In the vertical configuration, detachment is experimentally measured and predicted to start next to the strike point, extending towards the far SOL with increasing degree of detachment. In contrast, in the horizontal configuration, the entire divertor particle flux profile is reduced uniformly with increasing degree of detachment.

Modelling of plasma-edge and plasma-wall interaction physics at JET with the metallic first-wall

An overview is given on the recent progress on edge modelling activities for the JET ITER-like wall using the computational tools like the SOLPS or EDGE2D-EIRENE code. The validation process of the ...