Full Tungsten Wall Research Articles

Ripple-induced neoclassical toroidal viscous torque in Augmented-First Plasma operation phase in ITER

Abstract A systematic calculation is performed on the ripple-induced neoclassical toroidal viscous (NTV) torque for new ITER scenarios designed for the Augmented-First Plasma (A-FP) operation phase with the full tungsten wall, where the plasma-wall gap is varied in view of mitigating the impact of tungsten wall-plasma interactions. The torque calculation includes drift kinetic response of the plasma thermal and energetic particles to the n = 18 (n is the toroidal harmonic number) ripple field. For the plasma scenario with ~45 cm plasma-wall gap at the outboard mid-plane and considering the corrected ripple level of 0.17% by the ferritic steel inserts, the computed net NTV torque acting on the plasma column is in the sub-Nm level. However, with decreasing the plasma-wall gap, the computed net NTV torque can reach a level comparable to that produced by the neutral-beam momentum injection in ITER. Ripple correction by ferritic inserts reduces the net torque by a factor of 3.3 for all the three A-FP scenarios considered. The n ω d = l ω b (with ω d and ω b being the toroidal precession and bounce frequencies of trapped particles, respectively, and l an integer number) type of resonance-enhancement of the NTV torque, due to thermal particles, is found to be weak in ITER despite high-n of 18. The same also holds for the ITER 10 MA steady state scenario from the D-T operation phase, where the aforementioned resonance associated with fusion-born alphas is also included. The ripple-induced NTV torque is well below that produced by the resonant magnetic perturbation applied for controlling the type-I edge-localized mode in ITER.

Impact of increasing plasma-wall gap on plasma response to RMP fields in ITER

The impact of increasing plasma-wall gap on controlling edge-localized modes (ELMs) is numerically evaluated for the ITER 5 MA/2.65 T H-mode scenarios with full tungsten wall, based on the MARS-F computed plasma response to the applied n= 3–5 (n is the toroidal mode number) resonant magnetic perturbation (RMP) fields. Three new scenarios, referred to as standard, clearance and outergap, are considered assuming different plasma-wall gap sizes over a range on which vertically stability can be maintained by in-vessel coils in ITER. The latter are shown to have both direct and indirect effects on the plasma response and hence ELM control in ITER. The indirect and also primary influence occurs via change of the equilibrium edge safety factor q95 , which decreases with increasing the plasma-wall gap (at fixed plasma current and toroidal field), leading to a multi-peaking structure in the plasma response as measured by the plasma displacement near the X-point or the edge-localized resonant radial magnetic field perturbation. The direct, albeit secondary effect, is the reduction of local peak amplitudes with increasing the plasma-wall gap thus weakening the RMP field efficiency for ELM control with a given current in the control coils. A slight reduction of the plasma current, from 5 MA to 4.77, 4.92 and 4.65 MA for the standard, clearance and outergap scenarios, respectively, is found to be sufficient to access the q 95 window for the best ELM control with the n= 3 RMP. The n= 4 coil current configuration with the n= 5 sideband is also found favorable for ELM control in ITER, by producing RMP fields with mixed toroidal spectra compared to n = 3.

Impact of divertor neutral pressure on confinement degradation of advanced tokamak scenarios at ASDEX Upgrade

Over previous campaigns, an intense experimental program on advanced tokamak (AT) scenarios, has been carried out at the ASDEX Upgrade tokamak with full-tungsten wall. These discharges have been executed shortly after the boronization of the first wall to reduce the density and the impurity influx. The confinement level of such AT discharges was found to vary considerably, even when discharges with similar, if not identical, engineering parameters were carried out. This work investigates the causes of such confinement variations. Among all plasma quantities analyzed, confinement quality of AT scenarios correlates best with divertor neutral pressure, highlighting the key role of edge and scrape-off layer physics in determining global plasma confinement. In particular, it is found that the main cause of confinement degradation is the reduction of pedestal stability, which is in turn caused by the outward shift of the maximum density gradient position typically observed when the divertor neutral pressure increases. Owing to the low density of AT discharges under analysis, the movement of the maximum density gradient position can be caused entirely by changes in deuterium outgassing from the wall, which is strongly influenced by the boron layer deposited on the plasma-facing components and by the deuterium wall inventory. Finally, the predictive capability of confinement quality with the integrated model IMEP [Luda et al., Nucl. Fusion 60, 036023 (2020)] is tested on these discharges and shows promising results.

Investigation of helium exhaust dynamics at the ASDEX Upgrade tokamak with full-tungsten wall

An efficient removal of He ash by active pumping in future fusion devices is necessary to avoid fuel dilution and not degrade the core confinement properties. Therefore, a deep understanding of the underlying physics mechanisms is mandatory. Helium exhaust has been experimentally investigated at the ASDEX Upgrade tokamak. This is an ideal test environment, thanks to the ITER-like divertor geometry, an extensive diagnostics coverage and the presence of plasma-facing components made of tungsten. The exhaust efficiency, characterized by the He compression in the divertor, was found to improve with increasing divertor neutral pressure but to degrade with detachment. A multi-reservoir particle balance model was developed to interpret the observed exhaust dynamics, accounting for plasma transport and wall retention. The limited performance of the pumping system and the efficient helium retention capability of the tungsten wall were identified to have the strongest impact on the exhaust dynamics.

The versatile fueling systems of WEST

New fueling systems have been designed for WEST, a full tungsten wall tokamak with divertor configuration. They consist of standard gas injection, supersonic molecular beam injection and pellets injection systems, each being able to fuel continuously the plasma discharge up to a duration of 1000s. A massive gas injection system is also implemented to mitigate the impact of disruptions and runaway electrons. The standard gas system consists of 4 independent lines allowing the injection of up to 4 different gas species during a plasma discharge through 11 injections areas, 4 of them being located in the divertors and in the private flux region. The pellet injection system is able to inject up to 10 pellets per second through 4 tracks: one on the low field side, and the other 3 on the high field side. The supersonic molecular beam injection system consists of 3 injectors located close to the midplane, two on the high field side and one on the low field side. On trigger, the massive gas injection system can deliver in a single puff up 800 Pa.m3 of noble gas in order to stop the plasma discharge. All the fueling systems have been characterized and can be used routinely on plasma.

The updated ITPA global H-mode confinement database: description and analysis

The multi-machine International Tokamak Physics Activity (ITPA) Global H-mode Confinement Database has been upgraded with new data from JET with the ITER-like wall and ASDEX Upgrade with the full tungsten wall. This paper describes the new database and presents results of regression analysis to estimate the global energy confinement scaling in H-mode plasmas using a standard power law. Various subsets of the database are considered, focusing on type of wall and divertor materials, confinement regime (all H-modes, ELMy H or ELM-free) and ITER-like constraints. Apart from ordinary least squares (OLS), two other, robust regression techniques are applied, which take into account uncertainty on all variables. Regression on data from individual devices shows that, generally, the confinement dependence on density and the power degradation are weakest in the fully metallic devices. Using the multi-machine scalings, predictions are made of the confinement time in a standard ELMy H-mode scenario in ITER. The uncertainty on the scaling parameters is discussed with a view to practically useful error bars on the parameters and predictions. One of the derived scalings for ELMy H-modes on an ITER-like subset is studied in particular and compared to the IPB98(y, 2) confinement scaling in engineering and dimensionless form. Transformation of this new scaling from engineering variables to dimensionless quantities is shown to result in large error bars on the dimensionless scaling. Regression analysis in the space of dimensionless variables is therefore proposed as an alternative, yielding acceptable estimates for the dimensionless scaling. The new scaling, which is dimensionally correct within the uncertainties, suggests that some dependencies of confinement in the multi-machine database can be reconciled with parameter scans in individual devices. This includes vanishingly small dependence of confinement on line-averaged density and normalized plasma pressure (β), as well as a noticeable, positive dependence on effective atomic mass and plasma triangularity. Extrapolation of this scaling to ITER yields a somewhat lower confinement time compared to the IPB98(y, 2) prediction, possibly related to the considerably weaker dependence on major radius in the new scaling (slightly above linear). Further studies are needed to compare more flexible regression models with the power law used here. In addition, data from more devices concerning possible ‘hidden variables’ could help to determine their influence on confinement, while adding data in sparsely populated areas of the parameter space may contribute to further disentangling some of the global confinement dependencies in tokamak plasmas.

Overview of physics studies on ASDEX Upgrade

The ASDEX Upgrade (AUG) programme, jointly run with the EUROfusion MST1 task force, continues to significantly enhance the physics base of ITER and DEMO. Here, the full tungsten wall is a key asset for extrapolating to future devices. The high overall heating power, flexible heating mix and comprehensive diagnostic set allows studies ranging from mimicking the scrape-off-layer and divertor conditions of ITER and DEMO at high density to fully non-inductive operation (q95 = 5.5, ) at low density. Higher installed electron cyclotron resonance heating power 6 MW, new diagnostics and improved analysis techniques have further enhanced the capabilities of AUG.Stable high-density H-modes with MW m−1 with fully detached strike-points have been demonstrated. The ballooning instability close to the separatrix has been identified as a potential cause leading to the H-mode density limit and is also found to play an important role for the access to small edge-localized modes (ELMs). Density limit disruptions have been successfully avoided using a path-oriented approach to disruption handling and progress has been made in understanding the dissipation and avoidance of runaway electron beams. ELM suppression with resonant magnetic perturbations is now routinely achieved reaching transiently . This gives new insight into the field penetration physics, in particular with respect to plasma flows. Modelling agrees well with plasma response measurements and a helically localised ballooning structure observed prior to the ELM is evidence for the changed edge stability due to the magnetic perturbations. The impact of 3D perturbations on heat load patterns and fast-ion losses have been further elaborated.Progress has also been made in understanding the ELM cycle itself. Here, new fast measurements of and Er allow for inter ELM transport analysis confirming that Er is dominated by the diamagnetic term even for fast timescales. New analysis techniques allow detailed comparison of the ELM crash and are in good agreement with nonlinear MHD modelling. The observation of accelerated ions during the ELM crash can be seen as evidence for the reconnection during the ELM. As type-I ELMs (even mitigated) are likely not a viable operational regime in DEMO studies of ‘natural’ no ELM regimes have been extended. Stable I-modes up to have been characterised using -feedback.Core physics has been advanced by more detailed characterisation of the turbulence with new measurements such as the eddy tilt angle—measured for the first time—or the cross-phase angle of and fluctuations. These new data put strong constraints on gyro-kinetic turbulence modelling. In addition, carefully executed studies in different main species (H, D and He) and with different heating mixes highlight the importance of the collisional energy exchange for interpreting energy confinement. A new regime with a hollow profile now gives access to regimes mimicking aspects of burning plasma conditions and lead to nonlinear interactions of energetic particle modes despite the sub-Alfvénic beam energy. This will help to validate the fast-ion codes for predicting ITER and DEMO.

Real-time wall conditioning by controlled injection of boron and boron nitride powder in full tungsten wall ASDEX Upgrade

Abstract In ASDEX Upgrade with full-tungsten wall, boronization is an important tool to control tungsten sources and allow reliable operation at low collisionality. The duration of the beneficial effects is limited by the erosion of the boron layer, in particular on the tungsten ion-cyclotron antenna limiters. We report results from experiments exploring injection boron-rich powders in tokamak plasmas as a way to replenish boron coating and extend the lifetime of boronization effects. Pure boron and boron nitride powders were introduced gravitationally in plasma discharges at rates up to 60 mg/s for durations up to 3 s. Boron injection with cumulative amounts >100 mg appeared to improve wall conditions similarly to boronization, with indications of reduced influx of oxygen and tungsten from the antenna limiter. Moreover, cumulative injection of 340 mg of B appeared sufficient to enable successful execution of low collisionality scenarios critically relying on freshly boronized walls.

Turbulence in high-beta ASDEX upgrade advanced scenarios

Recent experiments at ASDEX Upgrade achieve non-inductive operation in full tungsten wall conditions by applying electron cyclotron and neutral beam current drive. These discharges are characterised by a well-measured safety factor profile, which does not drop below one, and a good energy confinement. By reproducing the experimental heat fluxes, nonlinear gyrokinetic simulations suggest that the observed strong peaking of the ion temperature in the core is caused by the stabilising impact of a significant beam ion content, as well as strong electromagnetic effects on turbulent transport. Quasilinear transport models are not yet applicable in this interesting and reactor relevant parameter regime, but available simulation data may serve as a testbed for improvements. As the present plasma is close to the kinetic ballooning (KBM) threshold, elevating the safety factor profile under otherwise identical conditions is proposed to clarify, whether profiles are ultimately limited by KBM turbulence, or by global stability constraints.

Impact of lithium pellets on plasma performance in the ASDEX Upgrade all-metal-wall tokamak

The impact of lithium (Li) on plasma performance was investigated at the ASDEX Upgrade tokamak, which features a full tungsten wall. Li pellets containing 1.6 × 1020 Li atoms were launched with a speed of 600 m s−1 to achieve deep penetration into the plasma and minimize the impact on the first wall. Homogeneous transient Li concentrations in the plasma of up to 15% were established. The Li sustainment time in the plasma decreased with an increasing heating power from 150 to 40 ms. Due to the pellet rate being restricted to 2 Hz, no Li pile-up could take place. No significant positive impact on plasma properties, as reported from other tokamak devices, could be found; the Li pellets rather caused a small reduction in plasma energy, mainly due to enhanced radiation. Due to pellet injection, a short-lived Li layer was formed on the plasma-facing components, which lasted a few discharges and led to moderately beneficial effects during plasma start-up. Most pellets were found to trigger type-I ELMs, either by their direct local perturbation or indirectly by the altered edge conditions; however, reliability was less than 100%.

On the role of the edge density profile for the L–H transition power threshold in ASDEX Upgrade

The L–H transition power threshold () in full tungsten (W) wall discharges is lower by 25% compared to those with graphite (C) mix tungsten walls in ASDEX Upgrade (Ryter et al 2013 Nucl. Fusion 53 113003). The lower power threshold in the full tungsten wall discharges has been found to correlate with higher edge density as well as steeper edge density gradient. An estimate of the minimum in the neoclassical radial electric field well inside the separatrix yields a constant value for all analyzed L–H transitions at fixed toroidal magnetic field (). The decrease of the threshold power is explained by the steeper edge density gradient in the discharges with full tungsten wall.

Pedestal and edge localized mode characteristics with different first wall materials and nitrogen seeding in ASDEX Upgrade

A comparison of ASDEX Upgrade (AUG) discharges performed with carbon and the full tungsten wall shows that the pedestal performance at low triangularity is not altered without gas puffing. The pedestal electron pressure is the same for both wall materials as is the confinement. With the tungsten wall the natural density is higher even without an additional gas puff. In typical operation with gas puffing the density is again higher in tungsten. This results in a higher collisionality with the tungsten wall. Pedestal pressure and plasma confinement, however, are not degraded until very large amounts of deuterium are puffed.The edge localized mode (ELM) crash in typical AUG discharges is observed to be composed of two independent phases. This is observed for both the carbon and the tungsten wall. The 1st phase of the crash is unaffected by scans of the plasma parameters as long as the pedestal pressure remains constant. The duration of the 2nd phase is strongly anti-correlated with the separatrix density and can be suppressed by the application of nitrogen seeding for divertor cooling. A consistent explanation for the two phases of the ELM crash does not seem possible when considering only the pre-ELM pedestal profiles. The scrape off layer (SOL) plasma provides the necessary free parameter for a consistent explanation, indicating the importance of the SOL in understanding the ELM crash evolution.

The effect of a metal wall on confinement in JET and ASDEX Upgrade

In both JET and ASDEX Upgrade (AUG) the plasma energy confinement has been affected by the presence of a metal wall by the requirement of increased gas fuelling to avoid tungsten pollution of the plasma. In JET with a beryllium/tungsten wall the high triangularity baseline H-mode scenario (i.e. similar to the ITER reference scenario) has been the strongest affected and the benefit of high shaping to give good normalized confinement of H98 ∼ 1 at high Greenwald density fraction of fGW ∼ 0.8 has not been recovered to date. In AUG with a full tungsten wall, a good normalized confinement H98 ∼ 1 could be achieved in the high triangularity baseline plasmas, albeit at elevated normalized pressure βN > 2. The confinement lost with respect to the carbon devices can be largely recovered by the seeding of nitrogen in both JET and AUG. This suggests that the absence of carbon in JET and AUG with a metal wall may have affected the achievable confinement. Three mechanisms have been tested that could explain the effect of carbon or nitrogen (and the absence thereof) on the plasma confinement. First it has been seen in experiments and by means of nonlinear gyrokinetic simulations (with the GENE code), that nitrogen seeding does not significantly change the core temperature profile peaking and does not affect the critical ion temperature gradient. Secondly, the dilution of the edge ion density by the injection of nitrogen is not sufficient to explain the plasma temperature and pressure rise. For this latter mechanism to explain the confinement improvement with nitrogen seeding, strongly hollow Zeff profiles would be required which is not supported by experimental observations. The confinement improvement with nitrogen seeding cannot be explained with these two mechanisms. Thirdly, detailed pedestal structure analysis in JET high triangularity baseline plasmas have shown that the fuelling of either deuterium or nitrogen widens the pressure pedestal. However, in JET-ILW this only leads to a confinement benefit in the case of nitrogen seeding where, as the pedestal widens, the obtained pedestal pressure gradient is conserved. In the case of deuterium fuelling in JET-ILW the pressure gradient is strongly degraded in the fuelling scan leading to no net confinement gain due to the pedestal widening. The pedestal code EPED correctly predicts the pedestal pressure of the unseeded plasmas in JET-ILW within ±5%, however it does not capture the complex variation of pedestal width and gradient with fuelling and impurity seeding. Also it does not predict the observed increase of pedestal pressure by nitrogen seeding in JET-ILW. Ideal peeling ballooning MHD stability analysis shows that the widening of the pedestal leads to a down shift of the marginal stability boundary by only 10–20%. However, the variations in the pressure gradient observed in the JET-ILW fuelling experiment is much larger and spans a factor of more than two. As a result the experimental points move from deeply unstable to deeply stable on the stability diagram in a deuterium fuelling scan. In AUG-W nitrogen seeded plasmas, a widening of the pedestal has also been observed, consistent with the JET observations. The absence of carbon can thus affect the pedestal structure, and mainly the achieved pedestal gradient, which can be recovered by seeding nitrogen. The underlying physics mechanism is still under investigation and requires further understanding of the role of impurities on the pedestal stability and pedestal structure formation.

Overview on plasma operation with a full tungsten wall in ASDEX Upgrade

Operation with all tungsten plasma facing components has become routine in ASDEX Upgrade. The conditioning of the device is strongly simplified and short glow discharges are used only on a daily basis. The long term fuel retention was reduced by more than a factor of 5 as demonstrated in gas balance as well as in post mortem analyses. Injecting nitrogen for radiative cooling, discharges with additional heating power up to 23MW have been achieved, providing good confinement (H98y2=1), divertor power loads around 5MWm−2 and divertor temperatures below 10eV. ELM mitigation by pellet ELM pacemaking or magnetic perturbation coils reduces the deposited energy during ELMs, but also keeps the W density at the pedestal low. As a recipe to keep the central W concentration sufficiently low, central (wave) heating is well established and low density H-Modes could be re-established with the newly available ECRH power of up to 4MW. The ICRH induced W sources could be strongly reduced by applying boron coatings to the poloidal guard limiters.

Integrated Modelling of ASDEX Upgrade Nitrogen Seeded Discharges

AbstractIn this paper we present integrated numerical modelling for the first time applied to ASDEX Upgrade nitrogen seeded discharges with full tungsten wall using the COREDIV code. It has been found that the code is able to reproduce experimental data for majority of the considered shots. In particular, satisfactory agreement with global core parameters like Pdiv, Prad, Zeff, Wmhd, τE and ΓpuffN has been found. However, COREDIV shows significantly larger W concentrations (factor 3‐10) than in experiment and the reason for this discrepancy is not clear. Some plasma profiles are also reasonably simulated by the code. This includes profiles of plasma density, temperatures, Zeff and plasma radiations (© 2012 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

FAST: Feasibility Analysis for a Completely Superconducting Magnet System

FAST (Fusion Advanced Studies Torus), the Italian proposal for a European satellite facility to ITER, is a compact tokamak ( R <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">0</sub> = 1.82 m, a= 0.64 m, triangularity δ = 0.4) able to investigate non linear dynamics effects of alpha-particle behavior in burning plasmas and to test technical solutions for the first wall/divertor directly relevant for ITER and DEMO (e.g.: full-tungsten wall and divertor and advanced liquid metal divertor). The machine is designed to operate with Deuterium plasmas in a dimensionless parameter range close to that of ITER and to access advanced tokamak regimes with long pulse duration with respect to the current diffusion time. It foresees a maximum magnetic field on plasma axis of 8.5 T and a maximum plasma current of 8 MA. In the present design phase, the feasibility of a superconducting solution for the magnet system is being investigated by ENEA. It consists of 18 Toroidal Field, 6 Poloidal Field and 6 Central Solenoid module coils, all of which wound by Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn and NbTi Cable-In-Conduit Conductors. All the main aspects driving the magnets' design, from mechanical to neutronic and thermal analyses, are here presented and discussed.

Compatibility of ITER scenarios with full tungsten wall in ASDEX Upgrade

The transition of ASDEX Upgrade (AUG) from a graphite device to a full tungsten device is demonstrated with a reduction by an order of magnitude in both the carbon deposition and deuterium retention. The tungsten source is dominated by sputtering from intrinsic light impurities, and the tungsten influxes from the outboard limiters are the main source for the plasma. In H-mode discharges, central heating (neutral beams, ECRH) is used to increase turbulent outward transport avoiding tungsten accumulation. ICRH can only be used after boronization as its application otherwise results in large W influxes due to light impurities accelerated by electrical fields at the ICRH antennas. ELMs are important in reducing the inward transport of tungsten in the H-mode edge barrier and are controlled by gas puffing. Even without boronization, stationary, ITER baseline H-modes (confinement enhancement factor from ITER 98(y, 2) scaling H98 ∼ 1, normalized beta βN ∼ 2), with W concentrations below 3 × 10−5 were routinely achieved up to 1.2 MA plasma current.The compatibility of high performance improved H-modes with unboronized W wall was demonstrated, achieving H98 = 1.1 and βN up to 2.6 at modest triangularities δ ⩽ 0.3 as required for advanced scenarios in ITER. With boronization the light impurities and the radiated power fraction especially in the divertor were reduced and the divertor plasma was actively cooled by N2 seeding. N2 seeding does not only protect the divertor tiles but also considerably improves the performance of improved H-mode discharges. The energy confinement increased to H98-factors of 1.25 (βN ∼ 2.7) and thereby exceeded the best values in a carbon-dominated AUG machine under similar conditions. Recent investigations show that this improvement is due to higher temperatures rather than to peaking of the electron density profile.Further ITER discharge scenario tests include the demonstration of ECRF assisted low voltage plasma start-up and current rise to q95 = 3 at toroidal electric fields below 0.3 V m−1, to achieve a ITER compatible range of plasma internal inductance of 0.71–0.97. The results reported here strongly support tungsten as a first wall material solution.

Compatibility of ITER scenarios with an all-W wall

In 2008, ASDEX Upgrade has started its second experimental campaign with full tungsten coverage of the plasma facing components. In the transition from a partially W-coated device (69% tungsten coverage in 2004/2005) to a full tungsten device (since 2007), post campaign analyses show a reduction by an order of magnitude in both the carbon deposition and deuterium retention for the experimental campaigns. Spectroscopic measurements show that the outer divertor is by far the strongest tungsten source region. However, influxes from the outboard limiters are the most important source for the tungsten content in the plasma. The application of ICRH results in large W influxes due to sputtering from light impurities accelerated by electrical fields at the ICRH antennas. In H-mode discharges, ELMs reduce the inward transport of tungsten in the H-mode edge transport barrier. Central heating provided by neutral beams and the upgraded ECRH systems is used to avoid tungsten accumulation in the core of the plasma. Stationary, ITER relevant H-modes (H98 ∼ 1, βN ∼ 1.6–2), with W concentrations below 3 × 10−5, were routinely achieved. High performance H-modes have been obtained before the first boronization, achieving H98 = 1.1–1.2 and βN up to 2.6 as required for advanced scenarios in ITER. Specific ITER studies performed in 2008 include the demonstration of low voltage plasma start-up with ECRH and heating during the current rise to q95 = 3, to achieve a range of plasma inductance of 0.71–0.97. The new results reported here form the basis of further enhancing the operational space of ASDEX Upgrade with the full tungsten wall, preparing for ITER and the ITER-like wall project in JET.