Table of contents

A theory-based scaling for the characteristic length of a circular, limited tokamak scrape-off layer (SOL) is obtained by considering the balance between parallel losses and non-linearly saturated resistive ballooning mode turbulence driving anomalous perpendicular transport. The SOL size increases with plasma size, resistivity, and safety factor q. The scaling is verified against flux-driven non-linear turbulence simulations, which reveal good agreement within a wide range of dimensionless parameters, including parameters closely matching the TCV tokamak. An initial comparison of the theory against experimental data from several tokamaks also yields good agreement.

Measurements are presented of the boron (B⁵⁺) density in the pedestal region at the low-field side (LFS) midplane and the high-field side (HFS) midplane of Alcator C-Mod. In H-mode plasmas, a large (≳10x) in–out asymmetry in impurity density forms, with larger densities at the HFS. In contrast, there is no impurity density asymmetry in L-mode or I-mode plasmas. A comparison of pedestal parameters in H-mode plasmas shows that the HFS impurity density pedestal width and position remain fairly fixed over a range of plasma conditions, while the LFS pedestal width widens, and the pedestal position shifts in towards the core as the plasma current is lowered, indicating a change in the underlying transport.

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)).

The modelling of a controlled tungsten dust injection experiment in TEXTOR by the dust dynamics code MIGRAINe is reported. The code, in addition to the standard dust–plasma interaction processes, also encompasses major mechanical aspects of dust–surface collisions. The use of analytical expressions for the restitution coefficients as functions of the dust radius and impact velocity allows us to account for the sticking and rebound phenomena that define which parts of the dust size distribution can migrate efficiently. The experiment provided unambiguous evidence of long-distance dust migration; artificially introduced tungsten dust particles were collected 120° toroidally away from the injection point, but also a selectivity in the permissible size of transported grains was observed. The main experimental results are reproduced by modelling.

In many discharges at ASDEX Upgrade (AUG) fast particle losses can be observed due to Alfvénic gap modes, reversed shear Alfvén eigenmodes or core-localized beta Alfvén eigenmodes. For the first time, simulations of experimental conditions in the AUG fusion device are performed for different plasma equilibria (particularly for different, also non-monotonic q profiles). The numerical tool is the extended version of the HAGIS code (Pinches et al 1998 Comput. Phys. Commun.111 133–49, Brüdgam 2010 PhD Thesis), which also computes the particle motion in the vacuum region between the vessel wall in addition to the internal plasma volume. For this work, a consistent fast particle distribution function was implemented to represent the strongly anisotropic fast particle population as generated by ICRH minority heating. Furthermore, HAGIS was extended to use more realistic eigenfunctions, calculated by the gyrokinetic eigenvalue solver LIGKA (Lauber et al 2007 J. Comput. Phys.226 447–65). The main aim of these simulations is to allow fast ion loss measurements to be interpreted with a theoretical basis. Fast particle losses are modelled and directly compared with experimental measurements (García-Muñoz et al 2010 Phys. Rev. Lett.104 185002). The phase space distribution and the mode-correlation signature of the fast particle losses allows them to be characterized as prompt, resonant or diffusive (non-resonant). The experimental findings are reproduced numerically. It is found that a large number of diffuse losses occur in the lower energy range (at around 1/3 of the birth energy) particularly in multiple mode scenarios (with different mode frequencies), due to a phase space overlap of resonances leading to a so-called domino (Berk et al 1995 Nucl. Fusion35 1661) transport process. In inverted q profile equilibria, the combination of radially extended global modes and large particle orbits leads to losses with energies down to 1/10th of the birth energy.

Gamma-ray spectrometry on ITER can provide information both on confined fusion alpha particles for optimization of plasma heating and runaway electrons, which is important for safe reactor operations. For the purpose of deconvolution of gamma-ray spectra recorded in fusion plasma experiments the DeGaSum code has been developed. The code can be applied for processing of both spectra of monoenergetic gamma rays, which are born in nuclear reactions produced by alpha particles and other fast ions, and continuous bremsstrahlung spectra generated by runaway electrons in the MeV range in the plasma and reactor structure materials. Gamma-ray spectrometer response functions and bremsstrahlung spectra generated by electrons in the MeV energy range are calculated and used in the DeGaSum code. The deconvolution of the discrete spectra allows the identification of nuclear reactions, which give rise to gamma rays, and the calculation of their intensities. By applying the code for continuous hard x-ray spectra, the runaway electron energy distribution can be inferred. It can provide the maximal energy of runaway electrons with accuracy, which satisfies the ITER project requirements. The code has been used for processing of spectra recorded in JET experiments. An application of the deconvolution technique for gamma-ray emission measurements on ITER is discussed.

This paper mainly focuses on a sector transport maintenance scheme from the aspects of high plant availability. In this study, three different maintenance schemes are considered based on (1) the number of maintenance ports and (2) the insertion direction. The design study clarifies critical design factors and key engineering issues on the maintenance scheme: (1) how to support an enormous overturning force of the toroidal field coils in the large open port for sector transport and (2) define the transferring mechanism of sectors in the vacuum vessel. On reviewing these assessment factors, the sector transport using a limited number of horizontal maintenance ports is found to be a more realistic maintenance scheme. In addition, evaluating maintenance scenarios under high decay heat is proposed for the first time. The key design factors are the cool-down time in the reactor and the cooling method in the maintenance scheme to keep components under operational temperature. Based on one-dimensional heat conduction analysis, after one month cool-down time, each sector of SlimCS could be transported to the hot cell facility by gas cooling.

Geodesic acoustic mode (GAM) and low-frequency zonal flow (LFZF) are both observed through Langmuir probe arrays during electron cyclotron resonance heating (ECRH) on the HL-2A tokamak edge. The radial distributions of the amplitude and peak frequency of GAM in floating potential fluctuations are investigated through rake probe arrays under different ECRH powers. It is observed that the GAM frequency would decrease and the intensity of carbon line emission would increase as the ECRH power exceeds a certain threshold. The analyses suggest that the impurity ions may play an important role in the GAM frequency at the edge region. It is also found that during the ECRH phase besides the mean flow, both GAM and LFZF are strengthened. The total fluctuation power and the fraction of that power associated with zonal flows both increase with the ECRH power, consistent with a predator–prey model. The auto- and cross-bicoherence analyses show the coupling between GAM and its second harmonic during the ECRH phase. Moreover, the results also suggest that the couplings between GAM and the components with multiple GAM frequency are strengthened. These couplings may be important for GAM saturation during the ECRH phase.

The 'European Transport Simulator' (ETS) (Coster et al 2010 IEEE Trans. Plasma Sci.38 2085–92, Kalupin et al 2011 Proc. 38th EPS Conf. on Plasma Physics (Strasbourg, France, 2011) vol 35G (ECA) P. 4.111) is the new modular package for 1D discharge evolution developed within the EFDA Integrated Tokamak Modelling (ITM) Task Force. It consists of precompiled physics modules combined into a workflow through standardized input/output data structures. Ultimately, the ETS will allow for an entire discharge simulation from the start up until the current termination phase, including controllers and sub-systems. The paper presents the current status of the ETS towards this ultimate goal. It discusses the design of the workflow, the validation and verification of its components on the example of impurity solver and demonstrates a proof-of-principles coupling of a local gyrofluid model for turbulent transport to the ETS. It also presents the first results on the application of the ETS to JET tokamak discharges with the ITER like wall. It studies the correlations of the radiation from impurity to the choice of the sources and transport coefficients.

The impact of edge localized modes (ELMs) and externally applied resonant and non-resonant magnetic perturbations (MPs) on fast-ion confinement/transport have been investigated in the ASDEX Upgrade (AUG), DIII-D and KSTAR tokamaks. Two phases with respect to the ELM cycle can be clearly distinguished in ELM-induced fast-ion losses. Inter-ELM losses are characterized by a coherent modulation of the plasma density around the separatrix while intra-ELM losses appear as well-defined bursts. In high collisionality plasmas with mitigated ELMs, externally applied MPs have little effect on kinetic profiles, including fast-ions, while a strong impact on kinetic profiles is observed in low-collisionality, low q₉₅ plasmas with resonant and non-resonant MPs. In low-collisionality H-mode plasmas, the large fast-ion filaments observed during ELMs are replaced by a loss of fast-ions with a broad-band frequency and an amplitude of up to an order of magnitude higher than the neutral beam injection prompt loss signal without MPs. A clear synergy in the overall fast-ion transport is observed between MPs and neoclassical tearing modes. Measured fast-ion losses are typically on banana orbits that explore the entire pedestal/scrape-off layer. The fast-ion response to externally applied MPs presented here may be of general interest for the community to better understand the MP field penetration and overall plasma response.

The energy loss caused by the edge-localized mode (ELM) needs to be reduced for ITER operations with ELMy H-mode plasmas. The reduction in ELM energy loss by pellet injection for ELM pacing is studied by an integrated core/scrape-off layer/divertor transport code TOPICS-IB with a magnetohydrodynamic stability code and a pellet model taking account of the E × B drift of the pellet plasma cloud. It is found that the energy loss can be significantly reduced by a pellet injected to the pedestal plasma equivalent to that at the middle timing in the natural ELM cycle, whose pressure height is only about 5% lower than that of the natural ELM onset. In this case, pellet injection from the low-field side enables a small pellet, with about 1–2% of pedestal particle content and a speed high enough to approach the pedestal top, to reduce the energy loss significantly. With the above suitable conditions for ELM pacing, a pellet penetrates deep into the pedestal and triggers high-n ballooning modes with localized eigenfunctions near the pedestal top, where n is the toroidal mode number. Under suitable conditions, ELM pacing with reduced energy loss is successfully demonstrated in simulations, in which the gas puff reduction and the enhancement of divertor pumping can compensate for the core density increase due to additional particle fuelling by the pacing pellet.

The beam-based plasma diagnostics on ITER require high accuracy and reliability which accordingly put challenging requirements on the collisional–radiative (CR) models used for description of the excited states of beam atoms. These states are known to play an essential role in implementation and interpretation of neutral beam-based diagnostics, such as the motional Stark effect (MSE), charge-exchange recombination spectroscopy (CXRS), beam-emission spectroscopy (BES) and in the dynamics of beam penetration. The latest analyses demonstrate that the widely used assumption of statistical populations among the beam excited states is questionable for high beam energies and high magnetic fields. Here we report an empirical verification of the recently developed non-statistical nkm-resolved CR model (Marchuk et al 2010 J. Phys. B: At. Mol. Opt. Phys.43 011002) and the short extrapolation to the relevant parameter range for ITER. The experiment was performed on the Alcator C-Mod tokamak, which operates in a unique range of parameters well suited for testing CR models for ITER beams. Beam emission spectra are collected for a selected range of plasma parameters. The line ratios σ₁/σ₀, π₄/π₃ and Σσ/Σπ are measured and compared to the n-resolved statistical population model and the new non-statistical nkm-resolved CR model. The measured ratios show clear deviations from the statistical model and a good agreement with the non-statistical results. The difference between the experimental values and nkm-resolved simulations for the most part is within 10% for all three line ratios. The largest deviation here is for the σ_1/σ₀ ratio that reaches up to 14% for the lowest electron density of 0.6 × 10²⁰ m⁻³. In contrast, the difference between experiment and the n-resolved (statistical) model is within 13–27% for the 50 keV/u Alcator C-Mod beam and expected to be up to 32% for 100 keV/u and up to 43% for 500 keV/u ITER beams. The discussed effect must be taken into account for the proposed CXRS/BES and MSE diagnostics for ITER.

A rigorous validation of the gyrokinetic model was performed in both ion temperature gradient (ITG) and trapped electron mode (TEM) dominated Alcator C-Mod plasmas at (normalized midplane minor radius) r/a = 0.5 and 0.8. Analysis focuses on two L-mode discharges operated with 1.2 and 3.5 MW of ion cyclotron resonance heating. In depth investigation into the experimental uncertainties and simulation sensitivities in these discharges allows for a stringent test of the gyrokinetic model implemented by the GYRO code (Candy and Waltz 2003 J. Comput. Phys.186 545) in both the centre of the stiff gradient region (r/a = 0.5) and the middle of the region often associated with the transport 'shortfall'(r/a = 0.8). To identify the nature of the plasma turbulence and to ensure a robust evaluation of the model's ability to reproduce experiment, the sensitivity of the simulation results to experimental uncertainty in turbulence drive and suppression terms were determined at both radial locations. When significant TEM activity is present, nonlinear gyrokinetic simulations are found to reproduce both electron and ion experimental heat fluxes within their diagnosed uncertainties. In contrast, in the absence of TEM, electron heat fluxes are robustly under predicted by low-k, gyrokinetic simulation.

The pedestal profile measurements in high triangularity JET plasmas show that with low fuelling the pedestal width decreases during the ELM cycle and with high fuelling it stays constant. In the low fuelling case the pedestal pressure gradient keeps increasing until the ELM crash and in the high fuelling case it initially increases then saturates during the ELM cycle.

Stability analysis reveals that both JET plasmas become unstable to finite-n ideal MHD peeling–ballooning modes at the end of the ELM cycle. During the ELM cycle, n = ∞ ideal MHD ballooning modes and kinetic ballooning modes are found to be locally stable in most of the steep pressure gradient region of the pedestal owing to the large bootstrap current, but to be locally unstable in a narrow region of plasma at the extreme edge.

Unstable micro-tearing modes are found at the JET pedestal top, but they are sub-dominant to ion temperature gradient modes. They are insensitive to collisionality and stabilized by increasing density gradient.

Power exhaust for a 3 GW class fusion reactor with an ITER-sized plasma was investigated by enhancing the radiation loss from seeding impurity. The impurity transport and plasma detachment were simulated under the Demo divertor condition using an integrated divertor code SONIC, in which the impurity Monte-Carlo code, IMPMC, can handle most kinetic effects on the impurity ions in the original formula. The simulation results of impurity species from low Z (neon) to high Z (krypton) and divertor length with a plasma exhausted power of 500 MW and radiation loss of 460 MW, and a fixed core–edge boundary of 7 × 10¹⁹ m⁻³ were investigated at the first stage for the Demo divertor operation scenario and the geometry design. Results for the different seeding impurities showed that the total heat load, including the plasma transport $(q_{{\rm target}}^{{\rm plasma}} )$ and radiation $(q_{{\rm target}}^{{\rm rad}} )$, was reduced from 15–16 MW m⁻² (Ne and Ar) to 11 MW m⁻² for the higher Z (Kr), and $q_{{\rm target}}^{{\rm rad}}$ extended over a wide area accompanied by increasing impurity recycling. The geometry effect of the long-leg divertor showed that full detachment was obtained, and the peak q_target value was decreased to 12 MW m⁻², where neutral heat load became comparable to $q_{{\rm target}}^{{\rm plasma}}$ and $q_{{\rm target}}^{{\rm rad}}$ due to smaller flux expansion. Fuel dilution was reduced but was still at a high level. These results showed that a divertor design with a long leg with higher Z seeding such as Ar and Kr is not fulfilled, but will be appropriate to obtain the divertor scenario for the Demo divertor. Finally, influences of χ and D_⊥ enhancement were seen significantly in the divertor, i.e. the radiation and density profiles became wider, leading to full detachment. Both q_target near the separatrix and T_e at the outer flux surfaces were decreased to a level for the conventional technology design. On the other hand, the problem of fuel dilution became worse. Extrapolation of the plasma transport coefficients to ITER and Demo, where density and temperature will be higher than ITER and edge-localized modes are mitigated, is a key issue for the divertor design.

Radiation-induced damage in tungsten (W) and W alloys has been considered as one of the most important issues in fusion research, because radiation-produced defects not only degrade the mechanical property but also change the behaviours of H and He in W significantly, such as the retention of H. Nano-structured W has been developed to reduce accumulation of defects within grains and further mitigate radiation-induced damage. However, the fundamental role of a grain boundary (GB) in healing radiation damage in W is not yet well understood. Using molecular dynamics and statics, we evaluate energetically and kinetically the role of a GB in defect evolution (vacancy and interstitial segregation and their annihilation) near the GB in W, by calculating the vacancy (interstitial) formation energy, segregation energy, diffusion barrier, vacancy–interstitial annihilation barrier near the GB and the corresponding influence range of the GB. We find that, as reported and expected, interstitials are preferentially trapped into GBs over vacancies during irradiation, with vacancies dominant near the GB and interstitials highly localized at the GB. On the one hand, the GB serves as a sink both for vacancies and interstitials near itself by reducing their formation energy and diffusion barrier. The formation energy of the vacancy decreases only by ∼0.86 eV, but 7.5 eV is reduced for the formation energy of the interstitial in the GB core, indicating that the sink is unexpectedly stronger for interstitials than vacancies. The average barrier of vacancy diffusion is 0.98 eV much less than 1.8 eV in the bulk; the interstitial migrates into the GB via a barrier-free process. On the other hand, the GB acts as a catalyst for the vacancy–interstitial recombination at the GB by lowering the annihilation barrier. The annihilation with the average barrier as low as 0.31 eV works even at a low temperature of 121 K; besides, the annihilation of a close vacancy–interstitial pair is spontaneous. Meanwhile, the annihilation mechanism near the GB is modified due to the exceptionally large reduction in the interstitial formation energy. The influence range of the GB is small (1–1.5 nm), leading to a small volume fraction of the GB region working as the sink and the catalyst. This indicates that GBs in fine W grains may play a limited role in improving radiation performance.

Transitions between low (L-mode) and intermediate (I-phase) confinement regimes triggered by sawteeth are investigated using multiple Langmuir probe arrays in the edge plasmas of the HL-2A tokamak. The I-phase is characterized by limit-cycle oscillations (LCOs). Repeated L–I–L transitions induced by sawtooth heat pulses are also observed. Statistical analyses show that the delay time of the L–I transition relative to sawtooth crashes is less than ∼1 ms. The intensity of turbulence bursts in the I-phase is stronger than those observed in L-mode plasmas. Measurements are consistent with zonal flow reducing turbulence intensity in the frequency band ∼20–100 kHz, while the LCO turbulence of frequency higher than 100 kHz has more power than that in the L-mode. The analyses of time-resolved power transfers between turbulence and zonal flows suggest that power transfers exist between turbulence and zonal flows. The direction of the power transfers is from turbulence into geodesic acoustic modes in ohmic plasmas.

In this work the finite β-effects of an electron branch of the geodesic acoustic mode (el-GAM) driven by electron temperature gradient (ETG) modes is presented. The work is based on a fluid description of the ETG mode retaining non-adiabatic ions and the dispersion relation for el-GAMs driven non-linearly by ETG modes is derived. The ETG growth rate from the fluid model is compared with the results found from gyrokinetic simulations with good agreement. A new saturation mechanism for ETG turbulence through the interaction with el-GAMs is found, resulting in a significantly enhanced ETG turbulence saturation level compared with the mixing length estimate. It is shown that the el-GAM may be stabilized by an increase in finite β as well as by increasing non-adiabaticity. The decreased GAM growth rates is due to the inclusion of the Maxwell stress.

This paper investigates the effect of the ITER-like wall (ILW) on runaway electron (RE) generation through a comparative study of similar slow argon injection JET disruptions, performed with different wall materials. In the carbon wall case, a RE plateau is observed, while in the ITER-like wall case, the current quench is slower and the runaway current is negligibly small. The aim of the paper is to shed light on the reason for these differences by detailed numerical modelling to study which factors affected the RE formation. The post-disruption current profile is calculated by a one-dimensional model of electric field, temperature and runaway current taking into account the impurity injection. Scans of various impurity contents are performed and agreement with the experimental scenarios is obtained for reasonable argon and wall impurity contents. Our modelling shows that the reason for the changed RE dynamics is a complex, combined effect of the differences in plasma parameter profiles, the radiation characteristics of beryllium and carbon, and the difference of the injected argon amount. These together lead to a significantly higher Dreicer generation rate in the carbon wall case, which is less prone to being suppressed by RE loss mechanisms. The results indicate that the differences are greatly reduced above ∼50% argon content, suggesting that significant RE current is expected in future massive gas injection experiments on both JET and ITER.

Infrared imaging of hot spots induced by localized magnetic perturbations using the test blanket module (TBM) mock-up on DIII-D is in good agreement with beam-ion loss simulations. The hot spots were seen on the carbon protective tiles surrounding the TBM as they reached temperatures over 1000 °C. The localization of the hot spots on the protective tiles is in fair agreement with fast-ion loss simulations using a range of codes: ASCOT, SPIRAL and OFMCs while the codes predicted peak heat loads that are within 30% of the measured ones. The orbit calculations take into account the birth profile of the beam ions as well as the scattering and slowing down of the ions as they interact with the localized TBM field. The close agreement between orbit calculations and measurements validate the analysis of beam-ion loss calculations for ITER where ferritic material inside the tritium breeding TBMs is expected to produce localized hot spots on the first wall.

Prompt neutral beam-ion loss due to non-resonant scattering caused by toroidicity-induced and reversed shear Alfvén eigenmodes (TAE/RSAEs) have been observed in DIII-D. The coherent losses are of full-energy beam ions born on unperturbed trapped orbits that would carry them close to a fast-ion loss detector (FILD) within one poloidal transit. However, in the presence of AEs, the particles are expelled from the plasma before completing their first poloidal orbits. The loss signals on FILD emerge within 100 µs after the beam switch-on (which is the time scale of a single poloidal transit) and oscillate at mode frequencies. Time-resolved loss measurements show a linear dependence on the AE fluctuation amplitude and a radial 'kick' of ∼10 cm by an n = 2 RSAE at δB/B ⩽ 1 × 10⁻³ can be directly inferred from the measurements. Full-orbit modelling of the fast-ion displacement caused by the AEs is in good quantitative agreement with the measurements. Direct interactions of the mode and the beam-ion orbit can account for a large fraction of fast-ion losses observed in such DIII-D discharges. The first orbit non-resonant loss mechanism may also contribute to enhanced localized losses in ITER and future reactors. A new diagnostic method of the radial displacement is inspired by these findings and can be used to study the interaction between fast ions and various MHD modes as well as three-dimensional fields.

The coupling of the internal kink to an external m/n = 1/1 perturbation is studied for profiles that are known to result in a saturated internal kink in the limit of a cylindrical tokamak. It is found from three-dimensional equilibrium calculations that, for A ≈ 30 circular plasmas and A ≈ 3 elliptical shapes, this coupling of the boundary perturbation to the internal kink is strong; i.e., the amplitude of the m/n = 1/1 structure at q = 1 is large compared with the amplitude applied at the plasma boundary. Evidence suggests that this saturated internal kink, resulting from small field errors, is an explanation for the TEXTOR and JET measurements of q₀ remaining well below unity throughout the sawtooth cycle, as well as the distinction between sawtooth effects on the q-profile observed in TEXTOR and DIII-D. It is proposed that this excitation, which could readily be applied with error field correction coils, be explored as a mechanism for controlling sawtooth amplitudes in high-performance tokamak discharges. This result is then combined with other recent tokamak results to propose an L-mode approach to fusion in tokamaks.

Modelling tritium retention in radiation-damaged pure-tungsten samples, the concentration of deuterium retained in the tungsten-ion-induced damage zone decreases with increasing exposure temperature; whereas for a similarly treated tungsten–rhenium alloy, it can drop significantly faster than in pure tungsten, given sufficiently high temperatures. In contrast, similar changes in retention behaviour were not observed in undamaged samples. Based on these findings, as well as on corresponding previous TEM results, it is concluded that the concentration of high-energy radiation-induced defects responsible for trapping of deuterium is lower in the alloy than in pure tungsten. Therefore, a small amount of transmutation rhenium in damaged tungsten should be able to keep tritium retention to a low level in ITER.

In the wall-stabilized high-β plasmas in JT-60U and DIII-D, interactions between energetic particle (EP) driven modes (EPdMs) and edge localized modes (ELMs) have been observed. The interaction between the EPdM and ELM are reproducibly observed. Many EP diagnostics indicate a strong correlation between the distorted waveform of the EPdM and the EP transport to the edge. The waveform distortion is composed of higher harmonics (n ⩾ 2) and looks like a density snake near the plasma edge. According to statistical analyses, ELM triggering by the EPdMs requires a finite level of waveform distortion and pedestal recovery. ELM pacing by the EPdMs occurs when the repetition frequency of the EPdMs is higher than the natural ELM frequency. EPs transported by EPdMs are thought to contribute to change the edge stability.

The mechanisms of edge-localized mode (ELM) energy deposition are studied by means of non-linear magnetohydrodynamic (MHD) simulation of ELMs. The footprint of the ELM heat flux at the divertor is found to increase approximately linearly with the total ELM energy loss for JET-scale plasmas, which is similar to the experimentally observed broadening of the ELM energy deposition with ELM energy loss. For these relatively large ELMs, in which conductive losses dominate, the divertor footprint broadening is due to an increase in the magnetic perturbation of the ballooning mode with increasing ELM energy loss, which results in a widening of the homoclinic tangles intersecting the target. The first results from ELM simulations in the ITER Q = 10 scenario indicate that on the ITER scale the broadening is similar for conductive and convective ELMs at least up to an ELM energy loss of 4 MJ. For the larger conductive-type ELMs the magnetic perturbation and its homoclinic tangles determine the pattern of the ELM heat flux at the divertor target similar to the JET-scale results. For the smaller convective ELMs, the ELM footprint is determined by the radial distance travelled by plasma filaments expelled by the ELM and the loss of the plasma energy in the filaments along the magnetic field lines.

The paper presents results of a first analysis of the divertor performance during the L–H transition in ITER. The integrated model consists of the SOLPS4.3 code suite for the SOL and divertor, and the ASTRA code for the core and pedestal regions. The results of SOLPS4.3 are parametrized and used as the boundary conditions for ASTRA, ensuring a consistent description of the plasma core and the edge. Boundary conditions switch from those for wide (L-mode) to narrow (H-mode) SOL once the transition criterion is met. The results show that, for conditions for which a full-power operational space with acceptable power loading of the targets exists, a transition from the initial L-mode operation to H-mode can be found for the same assumptions, i.e. the full-power H-mode regime is accessible.

The paper presents results of SOLPS modelling of the edge plasma performance during the low-activation phase of ITER operation. The calculations show that the peak power loading of the divertor targets can reach the reactor-relevant level of 3 to 5 MW m⁻², even without the fusion reactions, rendering commissioning of the high heat flux components possible in this phase. Parametrization of the output of the SOLPS runs for the predominantly helium plasma concerned by the studies reported here is performed, thus providing the boundary conditions for modelling of the core and allowing efficient integration of the core and edge models. This approach, using the ASTRA code for core simulations, is applied to the analysis of hydrogen accumulation in helium plasmas due to H pellet injection. The latter is the only available option for early testing of ELM pace-making as an ELM control tool assuming H-mode in hydrogen will not be possible. Critical dilution with H down to 70% He in the core plasma can be reached in only 0.5 to 1 s or even shorter, depending on the assumptions made.

An assessment of ITER plasma parameters is carried out for the low activation phase that is required for commissioning the basic ITER systems including plasma control, heating and current drive. Such an operation is analysed for hydrogen, helium and deuterium plasmas for full field and current, as well as with magnetic field and plasma current reduced to half of their design values, B₀ = 2.65 T, I_p = 7.5 MA. Both hydrogen and deuterium neutral beam injection (NBI) are considered. We assess the possible domain for safe operation, and the possible target plasmas for commissioning the NBI, electron cyclotron heating (ECH) and ion cyclotron heating (ICH) systems, taking into account the constraints imposed by NB shine-through loss, Greenwald limit and access to H-mode operation. Simulations with the Automated System for Transport Analysis (ASTRA) show that for 33 MW of NBI with 20 MW of ECH, H-mode access is marginal for hydrogen plasmas. Good H-mode confinement, expected at P_NB + P_EC + P_IC > 1.5 P_L–H, is more likely for the helium and deuterium cases. It is found that plasma parameters, such as normalized beta, plasma density and current flat-top duration, for full power/half field/half current operation can be similar to those required for the DT long pulse operation. Preliminary assessment is also made of the maximum of tritium and neutron yield achievable in a single shot at the deuterium phase of ITER operation.

Nonlinear excitation of geodesic acoustic mode by drift waves is investigated in toroidally rotating tokamak plasmas using the ideal magnetohydrodynamic equations. It is found that both the radial wave number and the nonlinear growth rate of the driven geodesic acoustic mode increase with the speed of the toroidal rotation, where the decisive factor is the frequency shift of the geodesic acoustic mode induced by the toroidal rotation. The additional nonlinear terms induced by the toroidal rotation velocity make little effect, even though they promote the nonlinear excitation at all times.

The next step in the Wendelstein stellarator line is the large superconducting device Wendelstein 7-X, currently under construction in Greifswald, Germany. Steady-state operation is an intrinsic feature of stellarators, and one key element of the Wendelstein 7-X mission is to demonstrate steady-state operation under plasma conditions relevant for a fusion power plant. Steady-state operation of a fusion device, on the one hand, requires the implementation of special technologies, giving rise to technical challenges during the design, fabrication and assembly of such a device. On the other hand, also the physics development of steady-state operation at high plasma performance poses a challenge and careful preparation. The electron cyclotron resonance heating system, diagnostics, experiment control and data acquisition are prepared for plasma operation lasting 30 min. This requires many new technological approaches for plasma heating and diagnostics as well as new concepts for experiment control and data acquisition.

The PDF file provided contains web links to all articles in this volume.