Main Chamber Wall Research Articles

Performances of Rh and Mo mirrors under JET exposure

The aim of this work was to provide a comprehensive surface characterization of rhodium and molybdenum mirrors exposed on the main chamber wall of JET during campaigns with carbon plasma-facing components. As determined using X-ray photoelectron spectroscopy, Secondary Ion Mass Spectrometry, Scanning Electron Microscopy and optical measurements the co-deposits formed on the mirrors were composed of C, Be, Ni, Mo and Fe. Most of the metallic elements on the surface were in the oxidized states. After removing a part of the film by sputtering with Ar+ ions, Mo carbide state was studied by XPS. The reflectivity measurements showed the dependence of the thickness of the deposited film on the optical properties of the mirror: a thinner film on the Mo mirrors results in an identical reduction in reflectivity as that seen on the exposed Rh mirror.

Disruption heat loads and their mitigation in JET with the ITER-like wall

The new materials of the ITER-like wall (ILW) in JET have significant impact on the disruption process. The use of beryllium for the first wall and tungsten in the divertor has shown so far a strong reduction of the radiation during disruptions compared to the previously installed CFC wall. This results in slower current quench rates and accordingly a significant amount of magnetic energy is being dissipated in form of heat flux to the PFCs of the main chamber wall. This contribution gives a first analysis of the heat loads and their mitigation by massive gas injection.

Radiative type-III ELMy H-mode in all-tungsten ASDEX Upgrade

The type-III ELMy H-mode might be the solution for an integrated ITER operation scenario fulfilling the fusion power amplification factor (output fusion power to input heating power) of Q = 10 with simultaneous acceptable steady-state and transient power loads to the plasma-facing components. This highly radiative type-III ELMy H-mode is achieved by nitrogen seeding. Experiments on the tokamak ASDEX Upgrade, in which all plasma-facing components are coated with tungsten, showed favourable confinement conditions (H98(y,2) = 1) at high plasma pressure (βN = 2.4). The power load to the plasma-facing components is as low as ≈3 MW m−2 during the peak heat loads due to edge localized modes (ELMs) at high radiative power fractions of frad ≈ 0.75. In those high-density discharges the central impurity concentration is very low as a result of hollow nitrogen density profiles and minimal erosion of tungsten. The tungsten erosion is not only suppressed in between ELMs, when the divertor is detached, but also during the type-III ELM activity. Such low impurity concentrations in the plasma core might lead to even higher fusion amplification factors in ITER than 10. This is a demonstration of the compatibility of the radiating type-III ELMy H-mode with a tungsten divertor and main chamber wall, with sufficient confinement and favourable power exhaust characteristics. Together with previous results in all-carbon devices this demonstration strengthens the case for a potential application of a Q = 10 scenario on ITER.

Simulation of main chamber wall temperature rise resulting from massive neon gas injection shutdown of ITER

Simulations were performed to estimate the main chamber wall heating in ITER resulting from rapid discharge shutdown by neon massive gas injection (MGI). The TokSys current diffusion model coupled with a simplified impurity transport model was used. Impurity parallel flow was treated with a single-fluid pressure-driven flow model. Impurity cross-field diffusion was treated with an empirical diffusion coefficient estimated from present experiments, while impurity poloidal rotation was included empirically by extrapolation in minor radius from present experiments to ITER. For single-valve neon MGI, maximum wall temperatures of order 1100 K are predicted, somewhat below the melting temperature of beryllium (1560 K). Lower temperature excursions were obtained by increasing the number of gas valves, while higher wall temperatures could be obtained by turning up initial plasma thermal energy or cross-field transport coefficients. Highest wall temperatures tended to occur on the centre post during the start of the current quench phase, consistent with present experiments. These results suggest that a single port may be sufficient for safely initiating rapid shutdown in ITER, leaving other ports free for subsequent rapid shutdown tasks such as runaway electron mitigation.

Large ELM-like events triggered by core MHD in JET advanced tokamak plasmas: impact on plasmas profiles, plasma-facing components and heating systems

Large and infrequent collapse events have been observed in high βN advanced tokamak (AT) plasmas in JET. Although they have features similar to large ELMs, they were triggered by core MHD. They caused a considerable loss of the plasma thermal and fast particle energy (∼10% of the total stored energy), but the heat load in the divertor due to these collapse events was small as a fraction of the plasma energy loss compared with regular type-I ELMs. Instead, significant heating of the main chamber wall was observed. A large, toroidally asymmetric, increase in the neutral gas pressure outside the plasma was observed after such events, which caused arcs in the lower hybrid (LH) and ion cyclotron (IC) heating systems and increased reionization in the neutral beam (NB) injectors. The collapses resulted in a reduction in the electron and ion temperatures and toroidal rotation of the whole plasma, a rise in Zeff; and a sufficiently large increase in the peripheral electron density to completely black-out the ECE emission from the plasma core. These features have been modelled to gain an understanding of the plasma behaviour associated with these collapse events and the implication for the operation of AT plasma scenarios with high additional heating power will be discussed.

Overview of the second stage in the comprehensive mirrors test in JET

The first mirror test for ITER in JET with carbon walls has been completed. Thirty polycrystalline Mo mirrors including four coated with a 1 μm rhodium (Rh) film were exposed to plasma in the divertor region and in the main chamber. The mirrors were installed in eight cassettes of pan-pipe shape. The reflectivity of all mirrors exposed in the divertor has been degraded by 80–90% because of the formation of thick (>20 μm) flaking co-deposits on surfaces. Only small reflectivity losses (5–10%) occurred on mirrors located at the channel mouth of the cassettes from the main chamber wall. This is due to the in situ removal of deposited species by charge exchange neutrals. Deuterium, 12C and 9Be are the main isotopes detected on surfaces, but other isotopes (13C) are also found in some locations, thus indicating differences in the material migration. Rhodium coatings with an initial reflectivity that is 30% better than that of pure Mo survived the test without detachment, but their resultant reflectivity was the same as that of the exposed Mo surfaces.

Poloidal distribution of recycling sources and core plasma fueling in DIII-D, ASDEX-Upgrade and JET L-mode plasmas

Deuterium fueling profiles across the separatrix have been calculated with the edge fluid codes UEDGE, SOLPS and EDGE2D/EIRENE for lower single null, ohmic and low-confinement plasmas in DIII-D, ASDEX Upgrade and JET. The fueling profiles generally peak near the divertor x-point, and broader profiles are predicted for the open divertor geometry and horizontal targets in DIII-D than for the more closed geometries and vertical targets in AUG and JET. Significant fueling from the low-field side midplane may also occur when assuming strong radial ion transport in the far scrape-off layer. The dependence of the fueling profiles on upstream density is investigated for all three devices, and between the different codes for a single device. The validity of the predictions is assessed for the DIII-D configuration by comparing the measured ion current to the main chamber walls at the low-field side and divertor targets, and deuterium emission profiles across the divertor legs, and the high-field and low-field side midplane regions to those calculated by UEDGE and SOLPS.

Integrated modelling of a JET type-I ELMy H-mode pulse and predictions for ITER-like wall scenarios

A baseline type-I ELMy H-mode JET tokamak discharge in low triangularity has been analysed using the JINTRAC integrated code suite to obtain a self-consistent description of the plasma edge dynamics and core plasma confinement. The inter- and intra-ELM transport model of JINTRAC has been adapted as such to match the experimental pre- and post-ELM plasma pedestal profiles and at the same time to recover the observed ELM dynamics in terms of ELM frequency, ELM energy loss, target ELM wetted area and target heat flows. The presented and validated modelling results for the JET all-carbon device reference case are utilized to predict a type-I ELMy H-mode for the JET ITER-like wall (ILW) assuming a full-tungsten divertor and beryllium main-chamber wall using JINTRAC. By keeping all relevant transport parameters fixed for the inter- and intra-ELM phase as in the all-carbon reference case it is observed that a moderate amount of seeded neon impurity is necessary to compensate for a similar level of radiation when carbon is absent in the system. The results of the ILW model setup are finally taken to estimate the total amount of tungsten particles eroded per ELM from the target plates. A rough estimate of the core radiative fraction due to W accumulation is given, predicting that no substantial impact on typical a JET ILW type-I ELMy H-mode discharge core plasma performance is expected.

Comparison of resonant magnetic perturbation-induced particle transport changes in H-mode (DIII-D) and L-mode (MAST)

Recent experiments show the impact of resonant magnetic perturbations (RMPs) on the density (Schmitz et al 2008 Plasma Phys. Control. Fusion 50 124029, Evans et al 2008 Nucl. Fusion 48 024002, Kirk et al 2008 Nucl. Fusion 50 024002, Liang et al 2007 Phys. Rev. Lett. 98 265004), leading to a so-called density pump-out. Previous comparisons between DIII-D and TEXTOR have focused on the similarities of the deformation of the separatrix and the creation of striations at the intersection of the main chamber wall (Schmitz et al 2008 Plasma Phys. Control. Fusion 50 124029; Schmitz et al 2009 Phys. Rev. Lett. 103 165005). In this paper, we compare the difference in magnitude of the experimentally observed density pump-out in L-mode with H-mode in two diverted tokamaks: MAST and DIII-D. In order to address the differences in magnetic field from the coils, plasma shape and q95 between the two devices, we compute a weighted magnetic diffusion coefficient with a vacuum field line tracing code. This allows us to compare the changes in density pump-out with the weighted magnetic diffusion coefficient, using a simple particle diffusion model. We find that the density pump-out is vastly different in the two confinement regimes, suggesting different particle transport mechanisms. Since one main difference in transport characteristics between L- and H-mode is turbulence, we compare turbulent particle characteristics. We find that in L-mode (MAST) the fluctuations and E × B shear increase at the plasma edge, whereas in H-mode (DIII-D) the fluctuations decrease at the plasma edge. Deeper inside the core, the E × B shear remains similar in L-mode (MAST), whereas a large decrease that quickly saturates with RMP strength is observed in H-mode (DIII-D). These results suggest that the RMP-induced particle transport at the plasma edge in L-mode (MAST) is the result from increases in turbulent particle transport, whereas the results in H-mode (DIII-D) suggest a decrease in turbulent particle transport.

Tokamak plasma edge modelling including the main chamber wall

Quantifying main chamber wall recycling, erosion and resulting material migration, at least on the basis of known or empirical far scrape-off layer (SOL) processes, is still highly uncertain, despite its relevance for ITER and fusion reactor design studies. This affects, for example, the design problem of first mirror performance of many optical diagnostics in the harsh ITER environment. Poor computational access is not least due to a fundamental technical limitation in apparently all current tokamak edge plasma fluid codes, which implicates a wide computationally unresolved gap between the outermost plasma layer treated in codes and the real vessel wall. We show how the current ITER version of the B2-EIRENE code (SOLPS-4.3) can be extended to cover also this far SOL, on the same footing as the rest of the plasma transport model. We discuss consequences of this new model for estimating plasma power and particle sink terms caused by a fairly realistic wall in ITER based on the conventional Bohm criterion along all plasma–wall interfaces.Corrections were made to this article on 14 July 2011. The authors have been assigned to the correct affiliations.

Erosion simulation of first wall beryllium armour after ITER transient heat loads and runaway electrons action

Beryllium is foreseen as plasma facing armour for the first wall (FW) in ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8–10cm. In ITER significant heat loads during transient events (TE) and runaway electrons impact are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting, evaporation, and melt motion, which determine the life-time of the plasma facing components. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the J×B forces are analyzed for bulk Be and different sizes of Be-brushes. The damage of the FW due to heat loads caused by runaway electrons is numerically simulated.

Modelling rock fracturing and blast-induced rock mass failure via advanced discretisation within the discontinuous deformation analysis framework

Rock mass failure is a particularly complex process that involves the opening and sliding of existing discontinuities and the fracturing of the intact rock. This paper adopts an advanced discretisation approach to simulate rock failure problems within the discontinuous deformation analysis (DDA) framework. The accuracy of this approach in continuum analysis is verified first. Then, the advanced discretisation approach for fracturing modelling is presented, and the discretisation strategy is discussed. Sample rock static failures are simulated and the results are compared with experimental results. Thereafter, with a generalised definition of the artificial joints, this approach is further extended and applied in the simulation of blast-induced rock mass failures in which the instant explosion gas pressure obtained by the detonation pressure equation of state is loaded on the main blast chamber walls and the induced surrounding connected fracture surfaces. In the simulation instance of rock mass cast blasting, the whole process, including the blast chamber expansion, explosion gas penetration, rock mass failure and cast, and the formation of the final blasting pile, is wholly reproduced.

Density and temperature measurements of pulsed plasma produced inside a curved vacuum chamber

Recent experimental evidence suggests the importance of fast radial plasma transport in the scrape-off-layer (SOL) of tokamaks. The outward transport appears to be convective rather than diffusive, extends into the far SOL, and can produce significant recycling from the main-chamber walls, partially by passing the divertor. A plausible theoretical mechanism to explain this phenomenon is the radial transport of locally dense plasma created by turbulent processes. In our experiment a blob of plasma is produced using a plasma-gun and injected radially to the curved vacuum chamber. In this paper we report the measurement of electron density and temperature of the plasma blob produced by the plasma-gun inside the curved vacuum chamber using cylindrical Langmuir probes. The probes are moved both radially and transversely to flow direction. The electron density and temperature are measured ~ 1016 m-3 and 10-1 eV respectively for radial flow at constant discharging potential, base pressure as well as plasma pulse time.

Transverse drift velocity of a pulsed-plasma in a curved magnetic field

Recent experimental evidence suggests the importance of fast radial plasma transport in the scrape-off-layer (SOL) of tokamaks. The outward transport appears to be convective rather than diffusive, extends into the far SOL, and can produce significant recycling from the main-chamber walls, partially bypassing the divertor. A plausible theoretical mechanism to explain this phenomenon is the radial transport of blobs of locally dense plasma created by turbulent processes. In our experiment, a blob of plasma is produced using a plasma-gun and injected transverse to a non-uniform curved magnetic field inside a curved vacuum chamber. The plasma-blob from the washer stacked plasma gun interacts with the curved magnetic field, which has been created by a pair of coils. The coil is being energized by a capacitor bank. In this paper we report the radial velocity of the plasma-blob in presence of magnetic field. The radial velocity of plasma-blob decreases in the presence of magnetic field in comparison with the velocity without magnetic field.

Geometry and expected performance of the solid tungsten outer divertor row in JET

At JET new plasma-facing components for the main chamber wall and the divertor are being designed and built to mimic the expected ITER plasma wall conditions in the deuterium–tritium operation phase. The main wall elements at JET will be made of beryllium and the divertor plasma-facing surface will be made of tungsten. Most of the divertor tiles will consist of tungsten-coated Carbon Fibre Composite (CFC) material. However one toroidal row in the outer divertor will be made of solid, inertially cooled tungsten. The geometry of these solid tungsten divertor components is optimized within the boundary conditions of the interfaces and the constraints given by the electrodynamical forces. Shadowing calculations as well as rough field line penetration analysis is used to define the geometry of the tungsten lamella stacks. These calculations are based on a set of magnetic equilibria reflecting the operation domain of current JET plasma scenarios. All edges in poloidal and toroidal direction are shadowed to exclude near perpendicular field line impact. In addition, the geometry of the divertor structure is being optimized so that the fraction of the plasma wetted surface is maximised. On the basis of the optimized divertor geometry, performance calculations are done with the help of ANSYS to assess the maximum power exhaust possible with this inertially cooled divertor row.

An overview of a comprehensive First Mirror Test for ITER at JET

The test was performed with 32 stainless steel and molybdenum mirrors placed in pan-pipe shaped cassettes and exposed in JET in the divertor and on the main chamber wall for 127 000 s including 97 000 s of X-point operation. Surface composition and total reflectivity were determined afterwards. All mirrors from the divertor were coated with deuterated carbon deposits causing the reflectivity loss by a factor of 6–10 in the visible range. Flaking and exfoliation of deposits were observed in some cases. On the main chamber wall the deposition occurred mainly on mirrors located deep in cassette channels, whereas mirrors close to the channels entrances were free from deposits and retained fair reflectivity (∼90% of initial value) especially in the infra-red range. No significant differences in behaviour of steel and molybdenum were noted. The need for development of methods for mirror cleaning and/or protection in a reactor-class device is addressed.

Dynamics of erosion and deposition in tokamaks

In recent years, a general qualitative understanding has been reached about the major pathways of material migration in divertor tokamaks. Main chamber wall components have been identified as the major source of material erosion. The eroded material is transported by scrape-off layer flows, in the case of the ion B × ∇ B drift pointing towards the X-point, predominately towards the inner divertor leg, where it is deposited in the form of amorphous layers. On JET, where carbon is the main plasma-facing material, it has been found that the presence of deposited carbon rich layers determines the dynamic characteristics of further re-distribution of carbon, in particular towards remote areas. The transport from the strike point to the deposition location is mainly line-of-sight. The amount of eroded carbon depends on the surface type, with lower rates for the bare CFC and higher rates for deposited layers. The erosion rates in the inner divertor increase non-linearly with increasing ELM energies.

Be wall sources and migration in L-mode discharges after Be evaporation in the JET tokamak

First wall material erosion and migration after fresh Be evaporation in the JET tokamak were studied in a series of consecutive identical L-mode discharges. The evolution of Be and C wall and divertor sources towards steady state conditions after deposition of a finite amount of Be at the carbon first wall of JET by the beryllium evaporation procedure provides a data set for benchmarking impurity transport simulations. Furthermore the experiment serves as a reference case for comparison of Be erosion to that of the planned ITER-like wall experiment with mainly Be plasma facing components (PFCs) in the main chamber. The experimental results confirm the migration pattern obtained by campaign integrated accounting of impurity sources and sinks, which is characterised by the main chamber wall and outer divertor as erosion dominated zones and migration of eroded material predominantly to the inner divertor where the material is finally deposited at the target plates.

Modelling of radiation impact on ITER Beryllium wall

In the ITER H-Mode confinement regime, edge localized instabilities (ELMs) will perturb the discharge. Plasma lost after each ELM moves along magnetic field lines and impacts on divertor armour, causing plasma contamination by back propagating eroded carbon or tungsten. These impurities produce enhanced radiation flux distributed mainly over the beryllium main chamber wall. The simulation of the complicated processes involved are subject of the integrated tokamak code TOKES that is currently under development. This work describes the new TOKES model for radiation transport through confined plasma. Equations for level populations of the multi-fluid plasma species and the propagation of different kinds of radiation (resonance, recombination and bremsstrahlung photons) are implemented. First simulation results without account of resonance lines are presented.

Erosion simulation of first wall beryllium armour under ITER transient heat loads

The beryllium is foreseen as plasma facing armour for the first wall in the ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8–10 cm. In ITER significant heat loads during transient events (TE) are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. Melting thresholds and melt layer depth of the Be armour under transient loads are estimated for different temperatures of the bulk Be and different shapes of transient loads. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the Lorentz force are analyzed for bulk Be and different sizes of Be-brushes. The damage of FW under radiative loads arising during mitigated disruptions is numerically simulated.