Impurity analysis of JET DiMPle pulses

Divertor monitoring pulses (DiMPle) have been run in JET from the C35 campaign onwards. They provide an opportunity to study the impurity contamination of the plasma when it is limited by different surfaces within the machine, as well as the longer term behaviour of the impurities. In these discharges the plasma is first limited on the outer wall, then on the inner wall and, subsequently, in the X-point configuration the outer strike point is positioned on the horizontal tile 5 of the machine followed by tile 6 and then the vertical tile 7. The present study details the impurity behaviour in the DiMPle pulses from JET-ILW campaigns C35 to C38, which ran from 2015 to 2019. The impurities can…

Impurity behaviour in JET-ILW plasmas fuelled with gas and/or with pellets: a comparative study with the transport code COREDIV

Abstract This study deals with the comparison of impurity behaviour in pellet and gas fuelled JET-ITER like wall pulses with the aim of finding the mechanisms leading to the generally observed higher concentration of tungsten in pellet fuelled plasmas. In fact, tungsten is the main high-Z impurity in the JET-ILW plasmas and is responsible for most of the radiative losses in the plasma core. Analysis of the experimental data pertaining to pulses at different plasma currents, different input power and different electron densities is integrated by numerical modelling with the self-consistent fluid transport code COREDIV. Experimentally, and numerically, the ratio between the radiated power in …

Spectrometric analysis of inner divertor materials of JET carbon and ITER-like walls

Abstract One of main reasons of the Joint European torus (JET) transformation from the carbon (JET-C) to ITER-like (JET-ILW) wall was high tritium retention of carbon. In order to compare the tritium retention, samples of analogous positions of the plasma-facing side of vertical tiles No. 3 of two campaigns: JET-C (2008–2009) and JET-ILW (2011–2012) were cut out. Temperature-programmed tritium desorption spectrometry in He + 0.1% H2 gas flow showed that JET-C sample without a tungsten coating had by a factor of >20 higher surface concentration of tritium than JET-ILW tungsten-coated sample: 4.9 × 1013 and 1.7–2.2 × 1012 T atoms/cm2 respectively. Installation of metallic plasma facing wall i…

Core (XUV/VUV) and boundary (UV/vis/IR) plasma spectroscopy in fusion devices

This contribution describes the basic applications of passive optical emission spectroscopy in the visible and far-UV region of electromagnetic radiation to diagnostics of the magnetic confinement fusion plasma. To simplify and condense the broad topic it presents the most common ways of analyzing the spectra of atoms, ions and molecules in fusion plasma and disseminating results of those analysis to the non-spectroscopists. It provides the reasons for choosing some particular regions, elements and charge states to determine the impurity content and plasma-surface interactions in MCF (Magnetic Confinement Fusion) reactor. Examples used in the contribution are predominantly from measurements…

Measuring the isotope effect on the gross beryllium erosion in JET

Abstract The isotope effect, hydrogen (H) versus deuterium (D), on the gross beryllium (Be) erosion yield has been measured in ohmic limiter plasmas in JET tokamak by spectroscopic means. A simplified method to extract the effective sputtering yield from the quotient of the radiances of the D α or D γ and the Be II lines at 527 nm was applied. A clear isotope effect has been found, the erosion yield of D being about a factor of 2 larger compared to H in the whole explored plasma density range. This is in agreement with physical sputtering data obtained with H+ and D+ ion beams and also with material surface computer simulations. The already published contribution of chemically assisted phys…

Tritium in plasma-facing components of JET with the ITER-Like-Wall

The Role of Combined ICRF and NBI Heating in JET Hybrid Plasmas in Quest for High D-T Fusion Yield

Combined ICRF and NBI heating played a key role in achieving the world-record fusion yield in the first deuterium-tritium campaign at the JET tokamak in 1997. The current plans for JET include new experiments with deuterium-tritium (D-T) plasmas with more ITER-like conditions given the recently installed ITER-like wall (ILW). In the 2015-2016 campaigns, significant efforts have been devoted to the development of high-performance plasma scenarios compatible with ILW in preparation of the forthcoming D-T campaign. Good progress was made in both the inductive (baseline) and the hybrid scenario: a new record JET ILW fusion yield with a significantly extended duration of the high-performance pha…

Overview of the JET results in support to ITER

The 2014–2016 JET results are reviewed in the light of their significance for optimising the ITER research plan for the active and non-active operation. More than 60 h of plasma operation with ITER first wall materials successfully took place since its installation in 2011. New multi-machine scaling of the type I-ELM divertor energy flux density to ITER is supported by first principle modelling. ITER relevant disruption experiments and first principle modelling are reported with a set of three disruption mitigation valves mimicking the ITER setup. Insights of the L–H power threshold in Deuterium and Hydrogen are given, stressing the importance of the magnetic configurations and the recent m…

Tritium retention measurements by accelerator mass spectrometry and full combustion of W-coated and uncoated CFC tiles from the JET divertor

Abstract Accelerator mass spectrometry (AMS) and the full combustion method (FCM) followed by liquid scintillation counting were applied to quantitatively determine the tritium retention in the tungsten-coated carbon fibre composites (CFC), in comparison to uncoated CFC tiles from the JET divertor. The tiles were adjacent and exposed to plasma operations between 2007 and 2009. The tritium depth profiles are showing that the tritium retention on the W-coated tile was reduced by a factor of 13.5 in comparison to the uncoated tile whereas the bulk tritium concentration is approximately the same for both tiles.

Letter

We present a study of the power threshold for L–H transitions (PLH) in almost pure helium plasmas, obtained in recent experiments at JET with an ITER-like wall (Be wall and W divertor). The most notable new result is that the density at which PLH is minimum, ${\bar{n}}_{\text{e},\mathrm{min}}$, is considerably higher for helium than for deuterium and hydrogen plasmas. We discuss the possible implications for ITER in its pre-fusion operating power phase.

Fuel inventory and material migration of JET main chamber plasma facing components compared over three operational periods

Fuel retention and material migration results from JET ITER-like wall beryllium limiter tiles are presented for three operating periods. Ion beam analysis results support the general picture of erosion during limiter configurations with local deposition on tile ends far into the scrape off layer. Similar trends of fuel concentrations are observed in all JET operating periods; (i) low on surfaces exposed to high heat flux and erosion and (ii) higher in deposits. The pattern of fuel retention and deposition correlates with heat flux and distribution of limiter plasmas touching inner and outer limiters. The D/Be ratio in the thickest deposit is similar to 0.01. Global fuel retention attributed…

Optimising the use of ICRF waves in JET hybrid plasmas for high fusion yield

In the recent JET experimental campaign, good progress was made in the development of high-performance plasma scenarios compatible with the ITER-like wall (ILW). This paper reports on the optimisation of the use of ion cyclotron resonance frequency (ICRF) waves for the hybrid scenario in combination with neutral beam injection. The hybrid scenario is a candidate for ITER long-pulse operation. The combined NBI+ICRF power was increased to 33 MW and the record JET ILW fusion yield, averaged over 100 ms, from 2.3x1016 neutrons/s to 2.9x1016 neutrons/s with respect to the previous 2014 JET ILW fusion record. Impurity control with ICRF waves was one of the key means for extending the duration of …

Recent progress in L-H transition studies at JET: tritium, helium, hydrogen and deuterium

Abstract 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 …

Erosion and screening of tungsten during inter/intra-ELM periods in the JET-ILW divertor

Abstract Intra-ELM tungsten sources, which dominate the total W source, are quantified in the inner and outer divertor of JET-ILW. The amount of the sputtered W atoms for individual ELMs demonstrates a clear dependence on the ELM frequency. It decreases when the pedestal temperature is lower and, correspondingly, the ELM frequency is higher. Nevertheless, the entire gross erosion W source (the number of eroded W atoms per second due to ELMs) increases initially with ELM frequency and reaches its maximum at fELM ≈ 50–55 Hz followed by its reduction in the high frequency range. The in/out asymmetry of the intra-ELM W sources during ELMs is a critical issue and is investigated in this contribu…

Analysis of metallic impurity content by means of VUV and SXR diagnostics in hybrid discharges with hot-spots on the JET-ITER-like wall poloidal limiter

In preparation for the upcoming JET D-T campaign, great effort has been devoted during the 2015-2016 JET campaigns with the ITER-like wall (ILW) to the extension of the high performance H-mode phase in baseline and hybrid scenarios. Hybrid discharges were the only ones that have been stopped by the real-time vessel protection system due hot-spot formation on the outboard poloidal limiter. Generation of hot-spots was linked to the application of high neutral beams injection and ion cyclotron resonance heating (ICRH) power. In tokamaks with high-Z plasma components, the use of ICRH heating is also accompanied by an increased metallic impurity content. Simultaneous control of hot-spot temperat…