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We present here the results of the spectroscopic analysis of the high-resolution visible spectra of beryllium hydride and its isotopologues (BeH, BeD, and BeT), produced during plasma–surface interactions during limiter and divertor JET-ILW (ITER-like Wall) pulses. The hydride production, being an important part of the wall erosion via chemical-assisted physical sputtering, shows visible dependence on plasma and wall conditions, and also on the isotope content of the plasma. This work shows that this dependence is also true for the molecular energy distributions, parameterized by rotational and vibrational temperatures. During the increase in the vessel wall temperature by 350 K, with constant plasma parameters, rotational temperature of the excited BeD molecules increases by 500 K, whereas the vibrational temperature decreases by 400 K. Another experiment was the scan of the integrated edge density, during which electron density increases with the increasing edge density, and electron temperature decreases. During that scan for BeD, T rot decreases by 300 K and T vib by 400 K over the change in an integrated edge density of Δ N e = + 8 × 10 18 m − 2 . Both trends were compared with the trends in other isotopes, and the qualitative results are similar, but the limited data available restrict the possibility of a full quantitative comparison. To simplify the comparison between the results presented here and other experiments or modeling data, and also to check the comparability of the pulses performed in different isotopes, the Be ion line ratios are used as benchmarks of the edge plasma condition. [ABSTRACT FROM AUTHOR]

After the second Deuterium–Tritium Campaign (DTE2) in the JET tokamak with the ITER-Like Wall (ILW) and full tritium campaigns that preceded and followed after the DTE2, a sequence of fuel recovery methods was applied to promote tritium removal from wall components. The sequence started with several days of baking of the main chamber walls at 240 °C and at 320 °C. Subsequently, baking was superimposed with Ion-Cyclotron Wall Conditioning (ICWC) and Glow Discharge Conditioning (GDC) cleaning cycles in deuterium. Diverted plasma operation in deuterium with different strike point configurations, including a Raised Inner Strike Point (RISP) configuration, and with different plasma heating—Ion Cyclotron Resonance Frequency (ICRF) and Neutral Beam Injection (NBI)—concluded the cleaning sequence. Tritium content in plasma and in the pumped gas was monitored throughout the experiment. The applied fuel recovery methods allowed reducing the residual tritium content in deuterium NBI-heated plasmas to about 0.1% as deduced from neutron rate measurements. This value is well below the requirement of 1% set by the maximum 14 MeV fusion neutron budget allocated in the ensuing deuterium plasma campaign. The quantified tritium removal over the course of the experiment was 13.4 ± 0.7 × 10 22 atoms or 0.67 ± 0.03 g with ∼58% attributed to baking, ∼12.5% to ICWC, ∼26% to GDC, and ∼3.5% to first low power RISP plasmas. The experimentally estimated amount of removed tritium is in good agreement with long-term tritium accounting by the JET tritium reprocessing plant, in which the unaccounted amount was reduced by 0.71 g after the cleaning experiment. [ABSTRACT FROM AUTHOR]

Neutral beam injection (NBI) is a flexible auxiliary heating method for tokamak plasmas, capable of being efficiently coupled to the various plasma configurations required in the Tritium and mixed deuterium-tritium experimental campaign on the Joint European Torus (JET) device. High NBI power was required for high fusion yield and alpha particle studies and to provide mixed deuterium-tritium (D-T) fuelling in the plasma core, it was necessary to operate the JET NBI systems in both deuterium and tritium. Further, the pure tritium experiments performed required T NBI for high isotopic purity and reduced 14 MeV neutron yield. Accurate power calibrations are also essential to machine safety. Previously on JET there have been a number of questions raised on the NBI power calibration, in particular following the Trace Tritium Experiments (TTEs). Operator activities on the tokamak NBI system, including calibrations, were performed in 2020. Following these activities, a series of plasma experiments were devised to further corroborate the T NBI power by comparing the plasma response to the D NBI power. A series of stationary, L-mode plasmas were performed on JET with different beam combinations used in different phases of the same pulse. By comparing the plasma response for D and T NBI it was possible to corroborate the T NBI power calibration using the D NBI power calibration. The stored energy as measured by magnetic diagnostics, corrected for fast particle stored energy, show that the uncertainty in NBI power calibration in T is comparable to that in D. [ABSTRACT FROM AUTHOR]

Nitrogen (N2) will be used in ITER to enhance the radiative fraction to ∼90%, thereby cooling the edge plasma and preventing damage to the plasma-facing components. However, the reactivity of N2 with hydrogen isotopes can lead to the formation of tritiated ammonia (NT3). This should be considered in terms of the in-vessel tritium inventory, the regeneration of the cryo pumps, and the processes in the ITER de-tritiation plant. In the 'W' Environment in Steady-state Tokamak (WEST), a series of long L-mode discharges (∼50 s), with a constant N2 seeding from the outer strike point region has been performed on the upper actively cooled divertor. In the absence of active pumping, the N2 balance shows steady-state retention during plasma discharge, and is partially (∼35%) released in between discharges. Although a significant amount of N2(18.65 Pa m3) has been injected, the wall still exhibited N2 pumping capabilities. Under these conditions, as long as this N2 reservoir is not saturated, there is not enough N available for the detectable threshold of ND3 formation to be reached. In these WEST experiments, no ammonia is detected during the pulse or after the pulse in the outgassing phase. These results are consistent with and complementary to the N2 seeded experiments performed in the Joint European Torus (JET) with its ITER-like wall and in the Axially Symmetric Divertor Experiment (ASDEX) upgrade. [ABSTRACT FROM AUTHOR]

The 1-dimensional reaction-diffusion-convection code TOMATOR-1D describes plasma production by RF waves inside a tokamak using the Braginskii continuity and heat balance equations. The model simulates self-consistent radial density and temperature profiles for magnetised plasma mixtures of hydrogen and helium. The model reproduces the density profiles of X2 electron cyclotron resonance heating (ECRH) plasmas on TCV and proposes a Bohm-like poloidal magnetic field dependent scaling for anomalous diffusion and a convection scaling that results from drifts in the toroidal magnetic field configuration. A relation is proposed between the anomalous diffusion and the outward convection in toroidal plasmas. It is found that the EC absorption efficiency decreases at higher power, which is understood from the acceleration of electrons beyond the optimal energy for the electron impact ionisation of helium. A dramatic increase of the absorption efficiency is seen at intermediate vertical magnetic field values of which results in the highest density plasmas. Losses along the field lines in the vertical direction become dominant at higher fields which effectively reduces the plasma density in these discharges. To arrive at predictive capabilities towards ECRH plasmas on JT-60SA and ITER, the proposed scalings, subsuming dependencies on the torus major radius and the toroidal field strength, need to be validated in a multi-machine study. [ABSTRACT FROM AUTHOR]

After vertical displacement events (VDEs) in the JET-ILW tokamak, debris is produced in liquid form and ejected from the vessels beryllium upper dump plates (UDPs). Most of the droplets splash near the UDPs outboard tile 8 as documented by an in-vessel photographic survey. Moreover, the systematic collection of 'dust' from the main chamber and the divertor region provides evidence of small solid spherical Be particles and flat splats (with fairly circular, 'pancake-like', or elongated shapes) related to melting events (with a diameter of few µm prior the deposition). The migration of metallic Be is studied with the numerical code DUSTTRACK, capable of describing the full droplet dynamics in a non-stationary tokamak plasma configuration. The description of the JET plasma during the VDE proposed here allows for a sufficiently consistent reconstruction of the transient conditions of the background plasma, relying on actual physical measurements. Due to the lack of experimental data related to the initial properties of the ejecta, the statistically relevant sample of 4×105 particles was considered in the DUSTTRACK simulations. A reasonable fit of the initial parameters for the ejected droplets, as well as the time-dependent plasma parameters for the VDE resulted in good qualitative agreement with the post-mortem experimental findings from dust collectors and with the Be deposition pattern near UDPs on the low-field side. [ABSTRACT FROM AUTHOR]

The integrated commissioning of the tungsten (W) environment in steady-state tokamak (WEST) was started with the superconducting magnet cool-down and completed one year after by the injection of 2.5 MW of additional power. It consisted of several steps: local system commissioning, integrated commissioning without plasma, and first plasma operation. This article reviews these WEST commissioning phases and highlights the experiences and the lessons learned. The magnet cooling down operation was initiated in late September 2016. The vacuum vessel was pumped out three weeks before the first plasma. The impregnation and curing of the two in-vessel divertor coils was performed at 180 °C during two days contributing to the baking of the vacuum vessel. In the meanwhile, rehearsal sessions for testing the CODAC systems were organized every month. The toroidal, poloidal, and divertor field coils were energized and their performance was tested successfully allowing moving to the first plasma operation phase. After a few attempts, plasma breakdown was achieved confirming that all systems were properly running and synchronized. Due to several micro air leaks on fiber optic feedthroughs and issues with the design of the in-vessel stabilizing plates, plasma current ramp-up could not be achieved in a first step. Once these issues have been fixed, confined plasmas could be obtained. Diverted plasmas were then achieved and the plasma current was increased up to 800 kA for more than 10 s. Up to 2.5 MW of additional power was eventually successfully coupled to the plasma. [ABSTRACT FROM AUTHOR]

The detection limit for helium (He) partial pressure monitoring via the Penning discharge optical emission diagnostic, mainly used for tokamak divertor effluent gas analysis, is shown here to be possible for He concentrations down to 0.1% in predominantly deuterium effluents. This result from a dedicated laboratory study means that the technique can now be extended to intrinsically (non-injected) He produced as fusion reaction ash in deuterium-tritium experiments. The paper also examines threshold ionization mass spectroscopy as a potential backup to the optical technique, but finds that further development is needed to attain with plasma pulse-relevant response times. Both these studies are presented in the context of continuing development of plasma pulse-resolving, residual gas analysis for the upcoming JET deuterium-tritium campaign (DTE2) and for ITER. [ABSTRACT FROM AUTHOR]

We present electron transport parameters (the first Townsend coefficient, the dissociative attachment coefficient, the fraction of electron energy lost by collisions with NF3 molecules, the average and characteristic electron energy, the electron mobility and the drift velocity) in NF3 gas calculated from published elastic and inelastic electron–NF3 collision cross-sections using the BOLSIG+ code. Calculations were performed for the combined RB (Rescigno 1995 Phys. Rev. E 52 329, Boesten et al 1996 J. Phys. B: At. Mol. Opt. Phys.29 5475) momentum-transfer cross-section, as well as for the JB (Joucoski and Bettega 2002 J. Phys. B: At. Mol. Opt. Phys.35 783) momentum-transfer cross-section. In addition, we have measured the radio frequency (rf) breakdown curves for various inter-electrode gaps and rfs, and from these we have determined the electron drift velocity in NF3 from the location of the turning point in these curves. These drift velocity values are in satisfactory agreement with those calculated by the BOLSIG+ code employing the JB momentum-transfer cross-section. [ABSTRACT FROM AUTHOR]

This paper presents a demonstration experiment of ion cyclotron wall conditioning (ICWC) on TEXTOR covering all ITER wall conditioning aims and discusses the implications for ITER. O2/He-ICWC applied to erode carbon co-deposits removed 6.6 × 1021 C-atoms (39 pulses, 158 s cumulated discharge time). Large oxygen retention (71% of injected oxygen) prevented subsequent ohmic discharge initiation. Plasma operation was recovered by a 1h47 multi-pulse D2-ICWC procedure including pumping time between pulses with duty cycle of 2 s/20 s, cleaning the vessel from oxygen impurities, followed by a 23 min He-ICWC procedure (2 s/20 s), applied to desaturate the deuterium-loaded walls. A stable ohmic discharge was established on the first attempt right after the recovery procedure. The discharges showed improved density control and only slightly increased oxygen characteristic radiation levels (1–1.5 times). After the recovery procedure 36% of the injected O-atoms remained retained in the vessel, derived from mass spectrometry measurements. This amount is in the estimated range for storage in remote areas obtained from surface analysis of locally exposed samples. The removed amount of oxygen by D2 and He-ICWC obtained from mass spectrometry corresponds to the retention in plasma-wetted areas estimated by surface analysis. It is concluded that most of the removed oxygen stems from plasma-wetted areas while shadowed areas, e.g. behind poloidal limiters, may feature net retention of the discharge gas. On ITER, designed with a shaped first wall, the ICWC plasma-wetted area will approach the total surface area, reducing consequently the retention in remote areas. A tentative extrapolation of the carbon removal on TEXTOR to tritium removal from co-deposits on ITER in the 39 × 4 s O2/He-ICWC discharges, including pumping time between the RF pulses, corresponds on ITER to a tritium removal in the order of the estimated retention per 400 s DT-burn (140–500 mgT (Shimada and Pitts 2011 J. Nucl. Mater.415 S1013–6)). [ABSTRACT FROM AUTHOR]

JET underwent a transformation from a full carbon-dominated tokamak to a fully metallic device with beryllium in the main chamber and a tungsten divertor. This material combination is foreseen for the activated phase of ITER. The ITER-Like Wall (ILW) experiment at JET shall demonstrate the plasma compatibility with metallic walls and the reduction in fuel retention. We report on a set of experiments (Ip = 2.0 MA, Bt = 2.0-2.4 T, δ = 0.2-0.4) in different confinement and plasma conditions with global gas balance analysis demonstrating a strong reduction in the long-term retention rate by more than a factor of 10 with respect to carbon-wall reference discharges. All experiments are 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 is reached. The composition analysis shows high purity of the recovered gas, typically 99% D. For typical L-mode discharges (Paux = 0.5 MW), type III (Paux = 5.0 MW) and type-I ELMy H-mode plasmas (Paux = 12.0 MW) a drop of the deuterium retention rate normalized to the operational time in divertor configuration is measured from 1.27 x 1021, 1.37 x 1021 and 1.97 x 1021 Ds-1 down to 4. 8 1019, 7.2 x 1019 and 16 x 1019 Ds-1, respectively. The dynamic retention increases in the limiter phase in comparison with carbon-fibre composite, but also the outgassing after the discharge has risen in the same manner and overcompensates this transient retention. Overall an upper limit of the long-term retention rate of 1.5 x 1020 Ds-1 is 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 is in line with identification of fuel co-deposition with Be as the main mechanism for the residual long-term retention. The reduction widens the operational space without active cleaning in the DT phase in comparison with a full carbon device. [ABSTRACT FROM AUTHOR]

Tritium (T) retention constitutes an outstanding constraint for ITER. It has been proposed that the end of the discharge could be used for reducing the amount of tritium trapped in the device by switching to He or H2 injection during the ~200 s of plasmas following the burning phase (power and plasma current ramp down). Thanks to the long discharge capabilities of Tore Supra, long pulse experiments (>mn) have been carried out to evaluate the effectiveness of such a scenario in reducing the tritium inventory during plasma operations. Starting with the device operated only in D2, series of changeover experiments from D2 to He and from D2 to H2 have been carried out in Tore Supra. The results demonstrate that with He the amount of D recovered after 130 s is limited to 0.8 x 1022 D whilst no further gain is foreseen. From these experiments, it is demonstrated that He injection will not contribute to the drop of the tritium inventory in the vessel. In contrast, with H2 injection the amount of D recovered after 250 s is ~4.2 x 1022 D with no limitation observed in the amount that could be removed from the vessel. The higher efficiency in removing D from the vessel by H2 injection compared to He is attributed to the H charge-exchange (CX) flux (four to six times larger than the He CX flux) allowing for a significantly stronger plasma wall interaction with carbon deposition and layer areas. In Tore Supra, since most of the D retention through co-deposition with eroded material (C) takes place in these areas, H plasmas result in a better removal efficiency of D(T) from these regions. These experimental observations are supported by the results obtained using the EIRENE code for evaluating both the ion and CX fluxes for He and H plasmas. Finally, the consequences of removing D(T) from the vessel for the next discharges are unfavourable for both the He and H2 removal methods. Indeed, in both cases, twice the amount of D(T) removed through the isotope exchange has to be re-injected since co-deposition of the re-injected D(T) will also take place in addition to the plasma wall isotope exchange. In these conditions, the low efficiency of the H2 gas injection for controlling the plasma isotopic ratio inhibits a recovery of the initial plasma isotopic ratio over a time scale in the range of 200 s. [ABSTRACT FROM AUTHOR]

This paper studies the normal and abnormal regimes of a weak-current rf discharge in ammonia, nitrogen, hydrogen and N2O for the rf electric field frequencies of 13.56 and 27.12 MHz. We reveal that only the abnormal regime of burning is observed at low pressures when the current growth is accompanied by an rf voltage increase while the surface of the electrodes is completely covered with the discharge. The normal regime occurs at higher gas pressures when the current growth is due to the increase in the surface area occupied by the discharge on the electrodes. The discharge burns in the abnormal mode after the surface of the electrodes is completely covered with the discharge. We demonstrate that the normal current density is directly proportional to the gas pressure and it depends approximately on the square of the rf electric field frequency. We present an analytical model for two limiting cases: constant free path length and constant mobility of positive ions furnishing a satisfactory description of the experimental data. [ABSTRACT FROM AUTHOR]

ITER as a superconducting fusion machine needs efficient wall conditioning techniques for application in the presence of the permanent high toroidal magnetic field for (i) reducing the in-vessel impurity content, (ii) controlling the surface hydrogen isotopic ratio and (iii) mitigating the in-vessel long-term tritium inventory build-up. Encouraging results recently obtained with ion-cyclotron wall conditioning (ICWC) in the present-day tokamaks and stellarators have raised ICWC to the status of one of the most promising techniques available to ITER for routine inter-pulse and overnight conditioning with the ITER main ICRF heating system in the presence of the permanent high toroidal magnetic field. This paper is dedicated to a milestone experiment in ICWC research: the first simulation of ICWC operation in an equivalent ITER full-field scenario and the assessment of the wall conditioning effect on the carbon wall in the largest present-day tokamak JET. In addition, we address in this paper the following topics: (i) an analysis of the radio frequency (RF) physics of ICWC discharges, (ii) the optimization of the operation of ICRF antennas for plasma startup and (iii) an outlook for the performance of ICWC in ITER using the ICRF heating system. Important operational aspects of the conventional ICRF heating system in JET (the so-called A2 antenna system) for use in the ICWC mode are highlighted: (i) the ability of the antenna to ignite the cleaning discharge safely and reliably in different gases, (ii) the capacity of the antennas to couple a large fraction of the RF generator power (>50%) to low-density (1016-1018 m-3) plasmas and (iii) the ICRF absorption schemes aimed at improved RF plasma homogeneity and enhanced conditioning effect. Successful optimization of the JET-ICWC discharge parameters (BT = 3.3T, f = 25 MHz) resulted in a reliable operation of the JET A2 antennas and a high conditioning efficiency in a scenario imitating closely ITER full-field operation (BT = 5.3T, f = 40 MHz) with the fundamental ion-cyclotron resonance for deuterium (ω = ΩD+ ) located on-axis. Numerical modelling with the 3D electromagnetic code Micro Wave Studio, a 1D RF full wave code and a 0D plasma code allows extrapolating the results obtained on JET and other present-day tokamaks to ITER and provides good prospects for the use of the ITER ICRF antennas for ICWC purposes. [ABSTRACT FROM AUTHOR]

A shielded residual gas analyzer (RGA) system on Tore Supra can function during plasma operation and is set up to monitor the composition of the neutral gas in one of the pumping ducts of the toroidal pumped limited. This 'diagnostic RGA' has been used in long-pulse (up to 6 min) discharges for continuous monitoring of up to 15 masses simultaneously. Comparison of the RGA-measured evolution of the H2/D2 isotopic ratio in the exhaust gas to that measured by an energetic neutral particle analyzer in the plasma core provides a way to monitor the evolution of particle balance. RGA monitoring of corrective H2 injection to maintain proper minority heating is providing a database for improved ion cyclotron resonance heating, potentially with RGA-base feedback control. In very long pulses (>4 min) absence of significant changes in the RGA-monitored, hydrocarbon particle pressures is an indication of proper operation of the actively cooled, carbon-based plasma facing components. Also H2 could increase due to thermodesorption of overheated plasma facing components. [ABSTRACT FROM AUTHOR]

This paper reports current-voltage characteristics and pressure-voltage transition curves from the weak-current α-mode to the strong-current γ-mode for rf capacitive discharges in N2O at frequencies of 2 MHz, 13.56 MHz, and 27.12 MHz. At 2 MHz the rf discharge is mostly resistive whereas at 13.56 MHz and 27.12 MHz it is mostly capacitive. The weak-current α-mode was found to exist only above a certain minimum gas pressure for all frequencies studied. N. Yatsenko [Sov. Phys. Tech. Phys. 26, 678 (1981)] previously proposed that the α-γ transition corresponds to breakdown of the sheaths. However, we show that this is the case only for sufficiently high gas pressures. At lower pressure there is a smooth transition from the weak-current α-mode to a strong-current γ-mode, in which the sheaths produce fast electrons but the sheath has not undergone breakdown. [ABSTRACT FROM AUTHOR]

This paper reports measured and calculated breakdown curves in several gases of rf capacitive discharges excited at 13.56 MHz in chambers of three different geometries: parallel plates surrounded by a dielectric cylinder (“symmetric parallel plate”), parallel plates surrounded by a grounded metallic cylinder (“asymmetric parallel plate”), and parallel plates inside a much larger grounded metallic chamber (“large chamber”). The breakdown curves for the symmetric chamber have a multivalued section at low pressure. For the asymmetric chamber the breakdown curves are shifted to lower pressures and rf voltages, but the multivalued feature is still present. At higher pressures the breakdown voltages are much lower than for the symmetric geometry. For the large chamber geometry the multivalued behavior is not observed. The breakdown curves were also calculated using a numerical model based on fluid equations, giving results that are in satisfactory agreement with the measurements. [ABSTRACT FROM AUTHOR]

In this paper a spectroscopic investigation of an expanding surface wave produced plasma is presented. The electron temperature is evaluated by means of emission spectroscopy. The results show a strong and sharp increase of the electron temperature toward the end of the discharge, where the plasma is expanding into a region with lowered gas pressure. [ABSTRACT FROM AUTHOR]