A. Loarte

Progress, since the ITER Physics Basis publication (ITER Physics Basis Editors et al 1999 Nucl. Fusion 39 2137–2664), in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Experimental areas where significant progress has taken place are energy transport in the scrape-off layer (SOL) in particular of the anomalous transport scaling, particle transport in the SOL that plays a major role in the interaction of diverted plasmas with the main-chamber material elements, edge localized mode (ELM) energy deposition on material elements and the transport mechanism for the ELM energy from the main plasma to the plasma facing components, the physics of plasma detachment and neutral dynamics including the edge density profile structure and the control of plasma particle content and He removal, the erosion of low- and high-Z materials in fusion devices, their transport to the core plasma and their migration at the plasma edge including the formation of mixed materials, the processes determining the size and location of the retention of tritium in fusion devices and methods to remove it and the processes determining the efficiency of the various fuelling methods as well as their development towards the ITER requirements. This experimental progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma–materials interaction and the further validation of these models by comparing their predictions with the new experimental results. Progress in the modelling development and validation has been mostly concentrated in the following areas: refinement in the predictions for ITER with plasma edge modelling codes by inclusion of detailed geometrical features of the divertor and the introduction of physical effects, which can play a major role in determining the divertor parameters at the divertor for ITER conditions such as hydrogen radiation transport and neutral–neutral collisions, modelling of the ion orbits at the plasma edge, which can play a role in determining power deposition at the divertor target, models for plasma–materials and plasma dynamics interaction during ELMs and disruptions, models for the transport of impurities at the plasma edge to describe the core contamination by impurities and the migration of eroded materials at the edge plasma and its associated tritium retention and models for the turbulent processes that determine the anomalous transport of energy and particles across the SOL. The implications for the expected performance of the reference regimes in ITER, the operation of the ITER device and the lifetime of the plasma facing materials are discussed.

The 'Progress in the ITER Physics Basis' (PIPB) document is an update of the 'ITER Physics Basis' (IPB), which was published in 1999 [1]. The IPB provided methodologies for projecting the performance of burning plasmas, developed largely through coordinated experimental, modelling and theoretical activities carried out on today's large tokamaks (ITER Physics R&D). In the IPB, projections for ITER (1998 Design) were also presented. The IPB also pointed out some outstanding issues. These issues have been addressed by the Participant Teams of ITER (the European Union, Japan, Russia and the USA), for which International Tokamak Physics Activities (ITPA) provided a forum of scientists, focusing on open issues pointed out in the IPB. The new methodologies of projection and control are applied to ITER, which was redesigned under revised technical objectives. These analyses suggest that the achievement of Q > 10 in the inductive operation is feasible. Further, improved confinement and beta observed with low shear (= high βp = 'hybrid') operation scenarios, if achieved in ITER, could provide attractive scenarios with high Q (> 10), long pulse (>1000 s) operation with beta <no-wall limit and benign ELMs.

Plasma wall interaction (PWI) is important for the material choice in ITER and for the plasma scenarios compatible with material constraints. In this paper, different aspects of the PWI are assessed in their importance for the initial wall materials choice: CFC for the strike point tiles, W in the divertor and baffle and Be on the first wall. Further material options are addressed for comparison, such as W divertor/Be first wall and all-W or all-C. One main parameter in this evaluation is the particle flux to the main vessel wall. One detailed plasma scenario exists for a Q = 10 ITER discharge [G. Federici et al., J. Nucl. Mater. 290–293 (2001) 260] which was taken as the basis of further erosion and tritium retention evaluations. As the assessment of steady state wall fluxes from a scaling of present fusion devices indicates that global wall fluxes may be a factor of 4 ± 3 higher, this margin has been adopted as uncertainty of the scaling. With these wall and divertor fluxes, important PWI processes such as erosion and tritium accumulation have been evaluated: It was found that the steady state erosion is no problem for the lifetime of plasma-facing divertor components. Be wall erosion may pose a problem in case of a concentration of the wall fluxes to small wall areas. ELM erosion may drastically limit the PFC lifetime if ELMs are not mitigated to energies below 0.5 MJ. Dust generation is still a process which requires more attention. Conversion from gross or net erosion to dust and the assessment of dust on hot surfaces need to be investigated. For low-Z materials the build-up of the tritium inventory is dominated by co-deposition with eroded wall atoms. For W, where erosion and tritium co-deposition are small, the implantation, diffusion and bulk trapping constitute the dominant retention processes. First extrapolations with models based on laboratory data show small contributions to the inventory. For later ITER phases and the extrapolation to DEMO additional tritium trapping sites due to neutron-irradiation damage need to be taken into account. Finally, the expected values for erosion and tritium retention are compared to the ITER administrative limits for the lifetime, dust and tritium inventory.

Budget restrictions have forced the ITER Organization to reconsider the baseline divertor strategy, in which operations would begin with carbon (C) in the high heat flux regions, changing out to a full-tungsten (W) variant before the first nuclear campaigns. Substantial cost reductions can be achieved if one of these two divertors is eliminated. The new strategy implies not only that ITER would start-up on a full-W divertor, but that this component should survive until well into the nuclear phase. This paper considers the risks engendered by such an approach with regard to known W plasma-material interaction issues and briefly presents the current status of a possible full-W divertor design.

Analysis of Type I ELMs from ongoing experiments shows that ELM energy losses are correlated with the density and temperature of the pedestal plasma before the ELM crash. The Type I ELM plasma energy loss normalized to the pedestal energy is found to correlate across experiments with the collisionality of the pedestal plasma (ν*ped), decreasing with increasing ν*ped. Other parameters affect the ELM size, such as the edge magnetic shear, etc, which influence the plasma volume affected by the ELMs. ELM particle losses are influenced by this ELM affected volume and are weakly dependent on other pedestal plasma parameters. In JET and DIII-D, under some conditions, ELMs can be observed (`minimum' Type I ELMs with energy losses acceptable for ITER), that do not affect the plasma temperature. The duration of the divertor ELM power pulse is correlated with the typical ion transport time from the pedestal to the divertor target (τ||Front = 2πRq95/cs,ped) and not with the duration of the ELM-associated MHD activity. Similarly, the timescale of ELM particle fluxes is also determined by τ||Front. The extrapolation of the present experimental results to ITER is summarized.

Plasma-facing materials (PFMs) have to withstand particle and heat loads from the plasma and neutron loads during reactor operation. For ITER knowledge has been accumulated by operation experience and dedicated tests in present tokamaks as well as by laboratory experiments and modelling. The rationale for the selection of PFMs in ITER (Be, W, carbon fibre reinforced carbon) is described with regard to the critical issues concerning PFMs, esp. erosion during transient heat loads and the T-inventory in connection with the redeposition of carbon. In the fusion reactor generation after ITER the very stringent conditions of increased surface power to be removed from the plasma, a lifetime requirement of several operational years, high neutron fluences and increased operation temperature are exerting even more severe constraints on the selection of possible materials. Comparing these boundary conditions with materials under development and their further potential, only a narrow path is left regarding heat sink and PFMs. In this context the investigations on W as first wall material carried out e.g. in ASDEX Upgrade are being discussed as well as laboratory results on W-based material systems. The implications of these results are the starting point of what should form a consistent programme towards plasma-facing and heat sink materials for a fusion reactor.

In ITER, as in any tokamak, the first wall and divertor plasma-facing components (PFC) must provide adequate protection of in-vessel structures, sufficient heat exhaust capability and be compatible with the requirements of plasma purity. These functions take on new significance in ITER, which will combine long pulse, high power operation with severe restrictions on permitted core impurity concentrations and which, in addition, will produce transient energy loads on a scale unattainable in today’s devices. The current ITER PFC design has now reached a rather mature stage following the 2007 ITER Design Review. This paper presents the key elements of the design, reviews the physics drivers, essentially thermal load specifications, which have defined the concept and discusses a selection of material and design issues.

Interactions between the plasma and the vessel walls constitute a major engineering problem for next step fusion devices, such as ITER, determining the choice of the plasma-facing materials. A prominent issue in this choice is the tritium inventory build-up in the vessel, which must be limited for safety reasons. The initial material selection, i.e. beryllium (Be) on the main vessel walls, tungsten (W) on the divertor upper baffle and dome, and carbon fibre composite around the strike points on the divertor plates, results both from the attempt to reduce the tritium inventory and to optimize the lifetime of the plasma-facing components.

Some of the remaining crucial plasma edge physics and plasma–material interaction issues of the ITER tokamak are discussed in this paper, using either modelling or projections of experimental results from existing tokamak operation or relevant laboratory simulations. The paper covers the following subject areas at issue in the design of the ITER device: (1) plasma thermal loads during Type I ELMs and disruptions, ensuing erosion effects and prospects for mitigating measures, (2) control of co-deposited tritium inventory when carbon is used even on small areas in the divertor near the strike points, (3) efficiency of edge and core fuelling for expected pedestal densities in ITER, and (4) erosion and impurity transport with a full tungsten divertor. Directions and priorities of future research are proposed to narrow remaining uncertainties in the above areas.

The thermal and electromagnetic loads related to disruptions in ITER are substantial and require careful design of tokamak components to ensure they reach the projected lifetime and to ensure that safety relevant components fulfil their function for the worst foreseen scenarios. The disruption load specifications are the basis for the design process of components like the full-W divertor, the blanket modules and the vacuum vessel and will set the boundary conditions for ITER operations. This paper will give a brief overview on the disruption loads and mitigation strategies for ITER and will discuss the physics basis which is continuously refined through the current disruption R&D programs.

Progress in the definition of the requirements for edge localized mode (ELM) control and the application of ELM control methods both for high fusion performance DT operation and non-active low-current operation in ITER is described. Evaluation of the power fluxes for low plasma current H-modes in ITER shows that uncontrolled ELMs will not lead to damage to the tungsten (W) divertor target, unlike for high-current H-modes in which divertor damage by uncontrolled ELMs is expected. Despite the lack of divertor damage at lower currents, ELM control is found to be required in ITER under these conditions to prevent an excessive contamination of the plasma by W, which could eventually lead to an increased disruptivity. Modelling with the non-linear MHD code JOREK of the physics processes determining the flow of energy from the confined plasma onto the plasma-facing components during ELMs at the ITER scale shows that the relative contribution of conductive and convective losses is intrinsically linked to the magnitude of the ELM energy loss. Modelling of the triggering of ELMs by pellet injection for DIII-D and ITER has identified the minimum pellet size required to trigger ELMs and, from this, the required fuel throughput for the application of this technique to ITER is evaluated and shown to be compatible with the installed fuelling and tritium re-processing capabilities in ITER. The evaluation of the capabilities of the ELM control coil system in ITER for ELM suppression is carried out (in the vacuum approximation) and found to have a factor of ∼2 margin in terms of coil current to achieve its design criterion, although such a margin could be substantially reduced when plasma shielding effects are taken into account. The consequences for the spatial distribution of the power fluxes at the divertor of ELM control by three-dimensional (3D) fields are evaluated and found to lead to substantial toroidal asymmetries in zones of the divertor target away from the separatrix. Therefore, specifications for the rotation of the 3D perturbation applied for ELM control in order to avoid excessive localized erosion of the ITER divertor target are derived. It is shown that a rotation frequency in excess of 1 Hz for the whole toroidally asymmetric divertor power flux pattern is required (corresponding to n Hz frequency in the variation of currents in the coils, where n is the toroidal symmetry of the perturbation applied) in order to avoid unacceptable thermal cycling of the divertor target for the highest power fluxes and worst toroidal power flux asymmetries expected. The possible use of the in-vessel vertical stability coils for ELM control as a back-up to the main ELM control systems in ITER is described and the feasibility of its application to control ELMs in low plasma current H-modes, foreseen for initial ITER operation, is evaluated and found to be viable for plasma currents up to 5–10 MA depending on modelling assumptions.

This paper presents the results of a preliminary assessment conducted to estimate the thermal response and erosion lifetime of the ITER divertor targets clad either with carbon-fibre composite or tungsten during type I ELMs. The one-dimensional thermal/erosion model, used for the analyses, is briefly described. It includes all the key surface heat transfer processes such as evaporation, melting, and radiation, and their interaction with the bulk thermal response, and it is based on an implicit finite-difference scheme, which allows for temperature-dependent material properties.

The experimental characteristics of divertor detachment in the JET tokamak with the Mark I pumped divertor are presented for ohmic, L mode and ELMy H mode experiments with the main emphasis on discharges with deuterium fuelling only. The range over which divertor detachment is observed for the various regimes, as well as the influence of divertor configuration, direction of the toroidal field, divertor target material and active pumping on detachment, will be described. The observed detachment characteristics, such as the existence of a considerable electron pressure drop along the field lines in the scrape-off layer (SOL), and the compatibility of the decrease in plasma flux to the divertor plate with the observed increase of neutral pressure and Dα emission from the divertor region, will be examined in the light of existing results from analytical and numerical models for plasma detachment. Finally, a method to evaluate the degree of detachment and the window of detachment is proposed, and all the observations of the JET Mark I divertor experiments are summarized in the light of this new quantitative definition of divertor detachment.

As part of the ITER Design Review and in response to the issues identified by the Science and Technology Advisory Committee, the ITER physics requirements were reviewed and as appropriate updated. The focus of this paper will be on recent work affecting the ITER design with special emphasis on topics affecting near-term procurement arrangements. This paper will describe results on: design sensitivity studies, poloidal field coil requirements, vertical stability, effect of toroidal field ripple on thermal confinement, material choice and heat load requirements for plasma-facing components, edge localized modes control, resistive wall mode control, disruptions and disruption mitigation.

Recent research in scrape-off layer (SOL) and divertor physics is reviewed; new and existing data from a variety of experiments have been used to make cross-experiment comparisons with implications for further research and ITER. Studies of the region near the separatrix have addressed the relationship of profiles to turbulence as well as the scaling of the parallel power flow. Enhanced low-field side radial transport is implicated as driving parallel flows to the inboard side. The medium-n nature of edge localized modes (ELMs) has been elucidated and new measurements have determined that they carry ∼10–20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. The predicted divertor power loads for ITER disruptions are reduced while those to main chamber plasma facing components (PFCs) increase. Disruption mitigation through massive gas puffing is successful at reducing PFC heat loads. New estimates of ITER tritium retention have shown tile sides to play a significant role; tritium cleanup may be necessary every few days to weeks. ITER's use of mixed materials gives rise to a reduction of surface melting temperatures and chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.

The ITER divertor design is the culmination of years of physics and engineering effort, building confidence that this critical component will satisfy the requirements and meet the challenge of burning plasma operation. With 54 cassette assemblies, each weighing ∼9 tonnes, nearly 3900 actively cooled high heat flux elements rated to steady-state surface power flux densities of 10 MW m−2 and a total of ∼60 000 carbon fibre composite monoblocks and ∼260 000 tungsten monoblocks/flat tiles, the ITER divertor will be the largest and most advanced of its kind ever constructed. Both the ITER Design Review and subsequent follow-up activities have led to a number of modifications to the device, including the divertor design, significantly improving ITER's operational flexibility. This paper outlines the salient features of the final divertor design, with emphasis on the physics rationale that has defined the design choices and on the performance of the resulting configuration.

The ITER plasma-facing components directly face the thermonuclear plasma and include the divertor, the blanket and the test blanket modules with their corresponding frames. The divertor is located at the bottom of the plasma chamber and is aimed at exhausting the major part of the plasma thermal power (including alpha power) and at minimising the helium and impurity content in the plasma. The blanket system provides a physical boundary for the plasma transients and contributes to the thermal and nuclear shielding of the vacuum vessel and external machine components. It consists of modular shielding elements known as blanket modules which are attached to the vacuum vessel. Each blanket module consists of two major components: a plasma-facing first wall panel and a shield block. The test blanket modules are mock-ups of DEMO breeding blankets. There are three ITER equatorial ports devoted to test blanket modules, each of them providing for the allocation of two breeding modules inserted in a steel frame and in front of a shield block.

This paper is concerned with investigation of an erosion of the ITER-like divertor plasma facing components under plasma heat loads expected during the Type I ELMs in ITER. These experiments were carried out on plasma accelerator QSPA at the SRC RF TRINITI under EU/RF collaboration. Targets were exposed by series repeated plasma pulses with heat loads in a range of 0.5–1.5 MJ/m2 and pulse duration 0.5 ms. Erosion of CFC macrobrushes was determined mainly by sublimation of PAN-fibres that was less than 2.5 μm per pulse. The CFC erosion was negligible at the energy density less than 0.5 MJ/m2 and was increased to the average value 0.3 μm per pulse at 1.5 MJ/m2. The pure tungsten macrobrushes erosion was small in the energy range of 0.5–1.3 MJ/m2. The sharp growth of tungsten erosion and the intense droplet ejection were observed at the energy density of 1.5 MJ/m2.

Abstract The XGC1 edge gyrokinetic code is used to study the width of the heat-flux to divertor plates in attached plasma condition. The flux-driven simulation is performed until an approximate power balance is achieved between the heat-flux across the steep pedestal pressure gradient and the heat-flux on the divertor plates. The simulation results compare well against the empirical scaling λ q <?CDATA $\propto $ ?> <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" overflow="scroll"> <mml:mstyle displaystyle="false"> <mml:mo>∝</mml:mo> </mml:mstyle> </mml:math> 1/ <?CDATA $B_{{\rm P}}^{\gamma }$ ?> <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" overflow="scroll"> <mml:mstyle displaystyle="false"> <mml:msubsup> <mml:mi>B</mml:mi> <mml:mrow> <mml:mrow> <mml:mi mathvariant="normal">P</mml:mi> </mml:mrow> </mml:mrow> <mml:mrow> <mml:mi>γ</mml:mi> </mml:mrow> </mml:msubsup> </mml:mstyle> </mml:math> obtained from present tokamak devices, where λ q is the divertor heat-flux width mapped to the outboard midplane, γ = 1.19 as found by Eich et al (2013 Nucl. Fusion 53 093031), and B P is the magnitude of the poloidal magnetic field at the outboard midplane separatrix surface. This empirical scaling predicts λ q ≲ 1 mm when extrapolated to ITER, which would require operation with very high separatrix densities ( n sep / n Greenwald &gt; 0.6) (Kukushkin et al 2013 J. Nucl. Mater . 438 S203) in the Q = 10 scenario to achieve semi-detached plasma operation and high radiative fractions for acceptable divertor power fluxes. Using the same simulation code and technique, however, the projected λ q for ITER’s model plasma is 5.9 mm, which could be suggesting that operation in the ITER Q = 10 scenario with acceptable divertor power loads may be obtained over a wider range of plasma separatrix densities and radiative fractions. The physics reason behind this difference is, according to the XGC1 results, that while the ion magnetic drift contribution to the divertor heat-flux width is wider in the present tokamaks, the turbulent electron contribution is wider in ITER. Study will continue to verify further this important projection. A high current C-Mod discharge is found to be in a mixed regime: While the heat-flux width by the ion neoclassical magnetic drift is still wider than the turbulent electron heat-flux width, the heat-flux magnitude is dominated by the narrower electron heat-flux.

This paper is concerned with investigation of an erosion of the ITER-like divertor castellated targets of pure tungsten, lanthanum tungsten and CFC under plasma heat loads expected during the Type I ELMs and disruptions in ITER. These experiments were carried out on a plasma gun QSPA-T at the SRC RF TRINITI under EU/RF collaboration. The targets were exposed by series repeated plasma pulses with heat loads in the range of 0.2–2.5 MJ/m2 and a pulse duration of 0.5 ms. The erosion value as a function of pulse number and energy density were obtained. The erosion of lanthanum tungsten started at the lower energy density as compared with pure tungsten and was mainly due to a melt layer movement and a droplets ejection. Characteristics of ejected droplets were measured. The erosion of CFC macrobrushes under ELM and disruption heat loads was determined mainly by damage of PAN-fibers.

Operation of DEMO in comparison to ITER will be significantly more demanding, as various additional limitations of physical and technical nature have to be respected. In particular a set of extremely restrictive boundary conditions on divertor operation during and in between ELMs will have to be respected. It is of high importance to describe these limitations in order to consider them as early as possible in the ongoing development of the DEMO concept design. This paper extrapolates the existing physics basis on power and particle exhaust to DEMO.

DIII-D experiments on rapid shutdown runaway electron (RE) beams have improved the understanding of the processes involved in RE beam control and dissipation. Improvements in RE beam feedback control have enabled stable confinement of RE beams out to the volt-second limit of the ohmic coil, as well as enabling a ramp down to zero current. Spectroscopic studies of the RE beam have shown that neutrals tend to be excluded from the RE beam centre. Measurements of the RE energy distribution function indicate a broad distribution with mean energy of order several MeV and peak energies of order 30–40 MeV. The distribution function appears more skewed towards low energies than expected from avalanche theory. The RE pitch angle appears fairly directed (θ ∼ 0.2) at high energies and more isotropic at lower energies (ε < 100 keV). Collisional dissipation of RE beam current has been studied by massive gas injection of different impurities into RE beams; the equilibrium assimilation of these injected impurities appears to be reasonably well described by radial pressure balance between neutrals and ions. RE current dissipation following massive impurity injection is shown to be more rapid than expected from avalanche theory—this anomalous dissipation may be linked to enhanced radial diffusion caused by the significant quantity of high-Z impurities (typically argon) in the plasma. The final loss of RE beams to the wall has been studied: it was found that conversion of magnetic to kinetic energy is small for RE loss times smaller than the background plasma ohmic decay time of order 1–2 ms.

Abstract The non-linear reduced four-field RMHD model in cylindrical geometry was extended to include plasma rotation, neoclassical poloidal viscosity and two fluid diamagnetic effects. Interaction of the static resonant magnetic perturbations (RMPs) with the rotating plasmas in tokamaks was studied. The self-consistent evolution of equilibrium electric field due to RMP penetration is taken into account in the model. It is demonstrated that in the pedestal region with steep pressure gradients, mean flows perpendicular to the magnetic field, which includes <?CDATA $\vec{E}\times \vec{B}$ ?> and electron diamagnetic components plays an essential role in RMP screening by plasma. Generally, the screening effect increases for lower resistivity, stronger rotation and smaller RMP amplitude. Strong screening of central islands was observed limiting RMP penetration to the narrow region near the separatrix. However, at certain plasma parameters and due to the non-linear evolution of the radial electric field produced by RMPs, the <?CDATA $\vec{E}\times \vec{B}$ ?> rotation can be compensated by electron diamagnetic rotation locally. In this case, RMPs can penetrate and form magnetic islands. Typical plasma parameters and RMPs spectra on DIII-D, JET and ITER were used in modelling examples presented in the paper.

High repetition rate injection of deuterium pellets from the low-field side (LFS) of the DIII-D tokamak is shown to trigger high-frequency edge-localized modes (ELMs) at up to $12\ifmmode\times\else\texttimes\fi{}$ the low natural ELM frequency in $H$-mode deuterium plasmas designed to match the ITER baseline configuration in shape, normalized beta, and input power just above the $H$-mode threshold. The pellet size, velocity, and injection location were chosen to limit penetration to the outer 10% of the plasma. The resulting perturbations to the plasma density and energy confinement time are thus minimal ($&lt;10%$). The triggered ELMs occur at much lower normalized pedestal pressure than the natural ELMs, suggesting that the pellet injection excites a localized high-$n$ instability. Triggered ELMs produce up to $12\ifmmode\times\else\texttimes\fi{}$ lower energy and particle fluxes to the divertor, and result in a strong decrease in plasma core impurity density. These results show for the first time that shallow, LFS pellet injection can dramatically accelerate the ELM cycle and reduce ELM energy fluxes on plasma facing components, and is a viable technique for real-time control of ELMs in ITER.

Operating ITER in the reference inductive scenario at the design values of Ip = 15 MA and QDT = 10 requires the achievement of good H-mode confinement that relies on the presence of an edge transport barrier whose pedestal pressure height is key to plasma performance. Strong gradients occur at the edge in such conditions that can drive magnetohydrodynamic instabilities resulting in edge localized modes (ELMs), which produce a rapid energy loss from the pedestal region to the plasma facing components (PFC). Without appropriate control, the heat loads on PFCs during ELMs in ITER are expected to become significant for operation in H-mode at Ip = 6–9 MA; operation at higher plasma currents would result in a very reduced life time of the PFCs.

The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-mode's performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental observation of ion cyclotron range of frequency (ICRF) mode-conversion, ICRF flow drive, demonstration of lower-hybrid current drive at ITER-like densities and fields and, using a set of novel diagnostics, extensive validation of advanced RF codes. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. A summary of important achievements and discoveries are included.

Abstract The tritium aspects of the DT fuel cycle embody some of the most challenging feasibility and attractiveness issues in the development of fusion systems. The review and analyses in this paper provide important information to understand and quantify these challenges and to define the phase space of plasma physics and fusion technology parameters and features that must guide a serious R&amp;D in the world fusion program. We focus in particular on components, issues and R&amp;D necessary to satisfy three ‘principal requirements’: (1) achieving tritium self-sufficiency within the fusion system, (2) providing a tritium inventory for the initial start-up of a fusion facility, and (3) managing the safety and biological hazards of tritium. A primary conclusion is that the physics and technology state-of-the-art will not enable DEMO and future power plants to satisfy these principal requirements. We quantify goals and define specific areas and ideas for physics and technology R&amp;D to meet these requirements. A powerful fuel cycle dynamics model was developed to calculate time-dependent tritium inventories and flow rates in all parts and components of the fuel cycle for different ranges of parameters and physics and technology conditions. Dynamics modeling analyses show that the key parameters affecting tritium inventories, tritium start-up inventory, and tritium self-sufficiency are the tritium burn fraction in the plasma ( f b ), fueling efficiency ( η f ), processing time of plasma exhaust in the inner fuel cycle ( t p ), reactor availability factor (AF), reserve time ( t r ) which determines the reserve tritium inventory needed in the storage system in order to keep the plant operational for time t r in case of any malfunction of any part of the tritium processing system, and the doubling time ( t d ). Results show that η f f b &gt; 2% and processing time of 1–4 h are required to achieve tritium self-sufficiency with reasonable confidence. For η f f b = 2% and processing time of 4 h, the tritium start-up inventory required for a 3 GW fusion reactor is ∼11 kg, while it is &lt;5 kg if η f f b = 5% and the processing time is 1 h. To achieve these stringent requirements, a serious R&amp;D program in physics and technology is necessary. The EU-DEMO direct internal recycling concept that carries fuel directly from the plasma exhaust gas to the fueling systems without going through the isotope separation system reduces the overall processing time and tritium inventories and has positive effects on the required tritium breeding ratio (TBR R ). A significant finding is the strong dependence of tritium self-sufficiency on the reactor availability factor. Simulations show that tritium self-sufficiency is: impossible if AF &lt; 10% for any η f f b , possible if AF &gt; 30% and 1% ⩽ η f f b ⩽ 2%, and achievable with reasonable confidence if AF &gt; 50% and η f f b &gt; 2%. These results are of particular concern in light of the low availability factor predicted for the near-term plasma-based experimental facilities (e.g. FNSF, VNS, CTF), and can have repercussions on tritium economy in DEMO reactors as well, unless significant advancements in RAMI are made. There is a linear dependency between the tritium start-up inventory and the fusion power. The required tritium start-up inventory for a fusion facility of 100 MW fusion power is as small as 1 kg. Since fusion power plants will have large powers for better economics, it is important to maintain a ‘reserve’ tritium inventory in the tritium storage system to continue to fuel the plasma and avoid plant shutdown in case of malfunctions of some parts of the tritium processing lines. But our results show that a reserve time as short as 24 h leads to unacceptable reserve and start-up inventory requirements. Therefore, high reliability and fast maintainability of all components in the fuel cycle are necessary in order to avoid the need for storing reserve tritium inventory sufficient for continued fusion facility operation for more than a few hours. The physics aspects of plasma fueling, tritium burn fraction, and particle and power exhaust are highly interrelated and complex, and predictions for DEMO and power reactors are highly uncertain because of lack of experiments with burning plasma. Fueling by pellet injection on the high field side of tokamak has evolved to be the preferred method to fuel a burning plasma. Extrapolation from the DIII-D penetration scaling shows fueling efficiency expected in DEMO to be &lt;25%, but such extrapolations are highly uncertain. The fueling efficiency of gas in a reactor relevant regime is expected to be extremely poor and not very useful for getting tritium into the core plasma efficiently. Gas fueling will nonetheless be useful for feedback control of the divertor operating parameters. Extensive modeling has been carried out to predict burn fraction, fueling requirements, and fueling efficiency for ITER, DEMO, and beyond. The fueling rate required to operate Q = 10 ITER plasmas in order to provide the required core fueling, helium exhaust and radiative divertor plasma conditions for acceptable divertor power loads was calculated. If this fueling is performed with a 50–50 DT mix, the tritium burn fraction in ITER would be ∼0.36%, which is too low to satisfy the self-sufficiency conditions derived from the dynamics modeling for fusion reactors. Extrapolation to DEMO using this approach would also yield similarly low burn fraction. Extensive analysis presented shows that specific features of edge neutral dynamics in ITER and fusion reactors, which are different from present experiments, open possibilities for optimization of tritium fueling and thus to improve the burn fraction. Using only tritium in pellet fueling of the plasma core, and only deuterium for edge density, divertor power load and ELM control results in significant increase of the burn fraction to 1.8–3.6%. These estimates are performed with physics models whose results cannot be fully validated for ITER and DEMO plasma conditions since these cannot be achieved in present tokamak experiments. Thus, several uncertainties remain regarding particle transport and scenario requirements in ITER and DEMO. The safety standard requirements for protection of the public and release guidelines for tritium have been reviewed. General safety approaches including minimizing tritium inventories, reducing tritium permeation through materials, and decontaminating material for waste disposal have been suggested.

Recent experiments on the Type I ELMy H-mode regime performed at JET with improved diagnostics have expanded the range of parameters for the study of Type I ELM energy and particle losses. Deviations from the standard behaviour of such losses in some areas of the Type I ELMy H-mode operating space have revealed that the ELM losses are correlated with the parameters (density and temperature) of the pedestal plasma before the ELM crash, while other global ELM characteristics (such as ELM frequency) are a consequence of the ELM-driven energy and particle flux and of the in-between ELM energy and particle confinement. The relative Type I ELM plasma energy loss (to the pedestal energy) is found to correlate well with the collisionality of the pedestal plasma, showing a weak dependence on the method used to achieve those pedestal plasma parameters: plasma shaping, heating, pellet injection and impurity seeding. Effects of edge plasma collisionality and transport along the magnetic field on the Type I ELM particle and energy fluxes onto the divertor target have also been observed. Two possible physical mechanisms that may give rise to the observed collisionality dependence of ELM energy losses are proposed and their consistency with the experimental measurements investigated: collisionality dependence of the edge bootstrap current with its associated influence on the ELM MHD origin and the limitation of the ELM energy loss by the impedance of the divertor target sheath to energy flow during the ELM event.

The design, construction and operation of advanced divertors have been the main topics of tokamak research during the last decade. The design of these divertors (carried out with two-dimensional plasma modelling codes) has been optimized to provide: a large operating density range for partially detached plasmas with large radiative losses; satisfactory particle control by efficient deuterium pumping; and He exhaust capability sufficient to remove fusion ashes when extrapolated to next step devices. This article explains the criteria that have guided this optimization process and then critically reviews the experimental evidence that confirms or refutes the physics-based design criteria on which the various existing advanced divertors have been based.

We present the results of experiments in JET to study the effect of plasma shape on high density ELMy H-modes, with geometry of the magnetic boundary similar to that envisaged for the standard Q = 10 operation in ITER. The experiments described are single lower null plasmas, with standard q profile, neutral beam heating and gas fuelling, with average plasma triangularity δ calculated at the separatrix ~0.45-0.5 and elongation κ~1.75. In agreement with the previous results obtained in JET and other divertor Tokamaks, the thermal energy confinement time and the maximum density achievable in steady state for a given confinement enhancement factor increase with δ. The new experiments have confirmed and extended the earlier results, achieving a maximum line average density ne~1.1nGR for H98~0.96. In this plasma configuration, at 2.5 MA/2.7 T (q95~2.8), a line average density ~95% nGR with H98 = 1 and βN~2 are obtained, with plasma thermal stored energy content Wth being approximately constant with increasing density, as long as the discharge maintains Type I ELMs, up to nped~nGR (and ne~1.1nGR).

New experimental results on transient loads during ELMs and disruptions in present divertor tokamaks are described and used to carry out a extrapolation to ITER reference conditions and to draw consequences for its operation. In particular, the achievement of low energy/convective type I edge localized modes (ELMs) in ITER-like plasma conditions seems the only way to obtain transient loads which may be compatible with an acceptable erosion lifetime of plasma facing components (PFCs) in ITER. Power loads during disruptions, on the contrary, seem to lead in most cases to an acceptable divertor lifetime because of the relatively small plasma thermal energy remaining at the thermal quench and the large broadening of the power flux footprint during this phase. These conclusions are reinforced by calculations of the expected erosion lifetime, under these load conditions, which take into account a realistic temporal dependence of the power fluxes on PFCs during ELMs and disruptions.

Analysis of Type I ELMs from present experiments shows that ELM energy losses decrease with increasing pedestal plasma collisionality (ν∗ped) and/or increasing τFront∥, where (τ∥Front=2πRq95/cs,ped) is the typical ion transport time from the pedestal to the divertor target. ν∗ped and τFront∥ are not the only parameters that affect the ELMs, also the edge magnetic shear influences the plasma volume affected by the ELMs. ELM particle losses are influenced by this ELM affected volume and are weakly dependent on other pedestal plasma parameters. ‘Minimum’ Type I ELMs, with energy losses acceptable for ITER, where there is no change in the plasma temperature profile during the ELM, are observed for some conditions in JET and DIII-D. The duration of the divertor ELM power pulse is well correlated with τFront∥ and not with the duration of the ELM-associated MHD activity. Similarly, the time scale of ELM particle fluxes is also determined by τFront∥. The extrapolation of present experimental results to ITER is summarised.

Edge-Localized-Modes (ELMs) are expected to present a significant transient flux of energy and particles to the ITER divertor. The threshold for ablation of the graphite target will be reached if the ELM transient exceeds Q/t1/2 ∼ 45 MJ m−2 s−1/2 where Q is the ELM deposition energy density and t the ELM deposition time. The ablation parameter in ITER can be determined by scaling four factors from present experiments: the ELM energy loss from the core plasma, the fraction of ELM energy deposited on the divertor target, the area of the ELM profile onto the target, and finally the time for the ELM deposition. Review of the ELM energy loss of Type I ELM data suggests an ITER ELM energy loss of 2–6% of the stored energy or 25–80 MJ. The fraction of heating power crossing the separatrix due to ELMs is nearly constant (20–40%) resulting in an inverse relationship between ELM amplitude and frequency. Measurements on DIII-D and ASDEX-Upgrade indicate that 50–80% of the ELM energy is deposited on the target. There is currently no evidence for a large fraction of the ELM energy being dissipated through radiation. Profiles of the ELM heat flux are typically 1–2 times the width of steady heat flux between ELMs, with the ELM amplitude usually larger on the inboard target. The ELM deposition time varies from about 0.1 ms in JET to as high as 1.0 ms in ASDEX-Upgrade and DIII-D. The ELM deposition time for ITER will depend upon the level of conductive versus convective transport determined by the ratio of energy to particles released by the ELM. Preliminary analysis suggests that large Type I ELMs for low recycling H-mode may exceed the ablation parameter by a factor of 5. Promising regimes with smaller ELMS have been found at other edge operational regimes, including high density with gas puffing, use of RF heating and operation with Type III ELMs.

Results of a detailed study of the parameter space of the ITER divertor with the B2-Eirene code are presented. Relations between plasma parameters at the separatrix, the interface between the core and edge plasma, are parametrized to provide a set of boundary conditions for the core models. The reference ITER divertor geometry is compared with the straight target option, and the possibility of controlling the edge density by shifting the plasma equilibrium in ITER is explored.

The paper describes the results of a physics analysis of a modified divertor cassette for ITER. The issues addressed are the impact on the operational window, the effect of gas leaks through the broader gaps between the divertor cassettes and radiation power loading of different components of the cassettes. The analysis shows that the new design ensuring more flexibility for ITER operation remains acceptable within the framework of the usual trade-off between the target power loading and helium removal efficiency. The radiation load on the side walls of the cassette structures in the inter-cassette gaps is identified as a design constraint not previously considered.

This paper describes the response of plasma facing components manufactured with tungsten (macro-brush) and CFC to energy loads characteristic of Type I ELMs and disruptions in ITER, in experiments conducted (under an EU/RF collaboration) in two plasma guns (QSPA and MK-200UG) at the TRINITI institute. Targets were exposed to a series of repetitive pulses in QSPA with heat loads in a range of 1–2 MJ/m 2 lasting 0.5 ms. Moderate tungsten erosion, of less than 0.2 μm per pulse, was found for loads of ∼1.5 MJ/m 2 , consistent with ELM erosion being determined by tungsten evaporation and not by melt layer displacement. At energy densities of ∼1.8 MJ/m 2 a sharp growth of tungsten erosion was measured together with intense droplet ejection. MK-200UG experiments were focused on studying mainly vapor plasma production and impurity transport during ELMs. The conditions for removal of thin metal deposits from a carbon substrate were characterized.

Experiments in Alcator C-Mod in (Enhanced D-alpha) EDA H-modes with extrinsic impurity seeding (N2, Ne, and Ar) have demonstrated a direct correlation between plasma energy confinement and edge power flow, achieving values of H98 ≥ 1 for edge power flows only marginally exceeding the scaled power for access to H-mode confinement in these conditions. For lower Z impurity seeding (N2 and Ne), plasmas with high energy confinement are obtained with a radiative power fraction of 85% or larger and a reduction of the peak heat flux at the divertor by more than a factor of 5 compared to similar attached conditions. The H-mode plasmas thus achieved in Alcator C-Mod meet or exceed the requirements both in terms of divertor heat flux handling and energy confinement for ITER QDT = 10 operation and with an edge power flow only marginally above the H-mode threshold power (by 1.0–1.4) as expected in ITER.

Views Icon Views Article contents Figures & tables Video Audio Supplementary Data Peer Review Share Icon Share Twitter Facebook Reddit LinkedIn Tools Icon Tools Reprints and Permissions Cite Icon Cite Search Site Citation M. N. A. Beurskens, T. H. Osborne, P. A. Schneider, E. Wolfrum, L. Frassinetti, R. Groebner, P. Lomas, I. Nunes, S. Saarelma, R. Scannell, P. B. Snyder, D. Zarzoso, I. Balboa, B. Bray, M. Brix, J. Flanagan, C. Giroud, E. Giovannozzi, M. Kempenaars, A. Loarte, E. de la Luna, G. Maddison, C. F. Maggi, D. McDonald, R. Pasqualotto, G. Saibene, R. Sartori, Emilia R. Solano, M. Walsh, L. Zabeo, The DIII-D Team, The ASDEX Upgrade Team, JET-EFDA Contributors; H-mode pedestal scaling in DIII-D, ASDEX Upgrade, and JET . Physics of Plasmas 1 May 2011; 18 (5): 056120. https://doi.org/10.1063/1.3593008 Download citation file: Ris (Zotero) Reference Manager EasyBib Bookends Mendeley Papers EndNote RefWorks BibTex toolbar search Search Dropdown Menu toolbar search search input Search input auto suggest filter your search All ContentAIP Publishing PortfolioPhysics of Plasmas Search Advanced Search |Citation Search

The mitigation of divertor heat fluxes is an active topic of investigation on existing tokamaks. One approach uses radiation, both inside and outside the last closed flux surface (LCFS), to convert plasma thermal energy, usually directed towards dedicated plasma facing components, to soft X-ray and ultraviolet radiation, spread over a much larger surface area. Recent enhanced D-α H-mode experiments on Alcator C-Mod varied the ICRF input power and radiative power losses via impurity seeding to demonstrate that normalized energy confinement depends strongly on the difference between input power and the radiated power inside the LCFS. These investigations also show that when seeded with either Ne or N2, a factor of two and higher reduction in outer divertor heat flux is achieved while maintaining H98,y2 ∼ 1.0. Conversely, when seeding with Ar, confinement is limited to H98,y2 ∼ 0.8 for a similar level of exhaust power.

In-vessel, non-axisymmetric, control coils have proven to be an important option for mitigating and suppressing edge-localized modes (ELMs) in high performance operating regimes on a growing number of tokamaks. Additionally, an in-vessel non-axisymmetric ELM control coil is being considered in the ITER baseline design. In preparing for the initial operation of this coil set, a comprehensive study was carried out to characterize the linear superposition of the 3D vacuum magnetic field, produced by the ELM coil, on a series of equilibria representing nine standard ITER operating scenarios. Here, the spatial phase angle of toroidally distributed currents, specified with a cosine waveform, in the upper and lower rows of the ITER ELM coil (IEC) set is varied in 2° steps while holding the current in the equatorial row of coils constant. The peak current in each of the three toroidal rows of window-frame coils making up the IEC is scanned between 5 kAt and 90 kAt in 5 kAt steps and the width of the edge region covered by overlapping vacuum field magnetic islands is calculated. This width is compared to a vacuum field ELM suppression correlation criterion found in DIII-D. A minimum coil current satisfying the DIII-D criterion, along with an associated set of phase angles, is identified for each ITER operating scenario. These currents range from 20 kAt to 75 kAt depending on the operating scenario being used and the toroidal mode number (n) of the cosine waveform. Comparisons between the scaling of the divertor footprint area in cases with n = 3 perturbation fields versus those with n = 4 show significant advantages when using n = 3. In addition, it is found that the DIII-D correlation criterion can be satisfied in the event that various combinations of individual IEC window-frame coils need to be turned off due to malfunctioning components located inside the vacuum vessel. Details of these results for both the full set of 27 window-frame coils and various reduced sets, using either n = 3 and n = 4 perturbation fields, are discussed.

A 24 MW Electron Cyclotron (EC) system operating at 170 GHz and 3600 s pulse length is to be installed on ITER. The EC plant shall deliver 20 MW of this power to the plasma for Heating and Current Drive (H&amp;CD) applications. The EC system is designed for plasma initiation, central heating, current drive, current profile tailoring, and Magneto-hydrodynamic control (in particular, sawteeth and Neo-classical Tearing Mode) in the flat-top phase of the plasma. A preliminary design review was performed in 2012, which identified a need for extended application of the EC system to the plasma ramp-up, flattop, and ramp down phases of ITER plasma pulse. The various functionalities are prioritized based on those applications, which can be uniquely addressed with the EC system in contrast to other H&amp;CD systems. An initial attempt has been developed at prioritizing the allocated H&amp;CD applications for the three scenarios envisioned: ELMy H-mode (15 MA), Hybrid (∼12 MA), and Advanced (∼9 MA) scenarios. This leads to the finalization of the design requirements for the EC sub-systems.

A self-consistent analysis of the relevant physics regarding the formation and termination of runaway beams during mitigated disruptions by Ar and Ne injection is presented for selected ITER scenarios with the aim of improving our understanding of the physics underlying the runaway heat loads onto the plasma facing components (PFCs) and identifying open issues for developing and accessing disruption mitigation schemes for ITER. This is carried out by means of simplified models, but still retaining sufficient details of the key physical processes, including: (a) the expected dominant runaway generation mechanisms (avalanche and primary runaway seeds: Dreicer and hot tail runaway generation, tritium decay and Compton scattering of γ rays emitted by the activated wall), (b) effects associated with the plasma and runaway current density profile shape, and (c) corrections to the runaway dynamics to account for the collisions of the runaways with the partially stripped impurity ions, which are found to have strong effects leading to low runaway current generation and low energy conversion during current termination for mitigated disruptions by noble gas injection (particularly for Ne injection) for the shortest current quench times compatible with acceptable forces on the ITER vessel and in-vessel components (). For the case of long current quench times (), runaway beams up to ∼10 MA can be generated during the disruption current quench and, if the termination of the runaway current is slow enough, the generation of runaways by the avalanche mechanism can play an important role, increasing substantially the energy deposited by the runaways onto the PFCs up to a few hundreds of MJs. Mixed impurity (Ar or Ne) plus deuterium injection proves to be effective in controlling the formation of the runaway current during the current quench, even for the longest current quench times, as well as in decreasing the energy deposited on the runaway electrons during current termination.

Resonant magnetic perturbations (RMP) have extensively been demonstrated as a plausible technique for mitigating or suppressing large edge localized modes (ELMs). Associated with this is a substantial amount of theory and modelling efforts during recent years. Various models describing the plasma response to the RMP fields have been proposed in the literature, and are briefly reviewed in this work. Despite their simplicity, linear response models can provide alternative criteria, than the vacuum field based criteria, for guiding the choice of the coil configurations to achieve the best control of ELMs. The role of the edge peeling response to the RMP fields is illustrated as a key indicator for the ELM mitigation in low collisionality plasmas, in various tokamak devices.

Abstract This paper describes joint ITPA studies of the I-mode regime, which features an edge thermal barrier together with L-mode-like particle and impurity transport and no edge localized modes (ELMs). The regime has been demonstrated on the Alcator C-Mod, ASDEX Upgrade and DIII-D tokamaks, over a wide range of device parameters and pedestal conditions. Dimensionless parameters at the pedestal show overlap across devices and extend to low collisionality. When they are matched, pedestal temperature profiles are also similar. Pedestals are stable to peeling–ballooning modes, consistent with lack of ELMs. Access to I-mode is independent of heating method (neutral beam injection, ion cyclotron and/or electron cyclotron resonance heating). Normalized energy confinement H 98, y 2 ⩾ 1 has been achieved for a range of 3 ⩽ q 95 ⩽ 4.9 and scales favourably with power. Changes in turbulence in the pedestal region accompany the transition from L-mode to I-mode. The L–I threshold increases with plasma density and current, and with device size, but has a weak dependence on toroidal magnetic field B T . The upper limit of power for I-modes, which is set by I–H transitions, increases with B T and the power range is largest on Alcator C-Mod at B &gt; 5 T. Issues for extrapolation to ITER and other future fusion devices are discussed.

Erosion of the first-wall and divertor, and distribution of eroded material in combination with tritium codeposition (primarily with eroded carbon) over many pulses, remain critical issues for the design, operation, and safety of a long-pulse next-step fusion device, such as ITER. These issues are currently being investigated by experiments in tokamaks and in laboratories, as well as by modelling. In this study, we analyse erosion (e.g., by sputtering, ELMs, and off-normal transients) and codeposition effects in the reduced-size ITER device, called `ITER-FEAT', with a strike-point carbon divertor target and metallic walls, for a `semi-detached' edge plasma regime using two-dimensional profiles of plasma edge parameters, modelled by the code B2-EIRENE. This paper accompanies the overview paper given by G. Janeschitz et al. [Plasma wall interactions in ITER-FEAT, these Proceedings]. Tritium codeposition with chemically eroded carbon still presents removal/control challenges, albeit to a somewhat lesser extent than in the 1998 ITER design, and demands efficient tritium inventory removal/control techniques. Due to numerous model uncertainties, not the least of which are the plasma solutions themselves, our intent is to provide a scoping analysis, defining trends and suggesting further research needs.

This paper presents the experimental characterization of pedestal parameters, edge localized mode (ELM) energy, and particle losses from the main plasma and the corresponding ELM energy fluxes on plasma facing components for a series of dedicated experiments in the Joint European Torus (JET). From these experiments, it is demonstrated that the simple hypothesis relating the peeling-ballooning linear instability to ELM energy losses is not valid. Contrary to previous observations at lower triangularities, small energy losses at low collisionality have been obtained in regimes at high plasma triangularity and q95∼4.5, indicating that the edge plasma magnetohydrodynamic stability is linked with the transport mechanisms that lead to the loss of energy by conduction during type I ELMs. Measurements of the ELM energy fluxes on the divertor target show that their time scale is linked to the ion transport along the field and the formation of a high energy sheath, in agreement with kinetic modeling of ELMs. Higher density ELMs, of a convective nature, lead to overall much longer time scales for the ELM energy flux, with more than 80% of the ELM energy flux arriving after the surface divertor temperature has reached its maximum value. On the contrary, for low density ELMs, of a conductive nature, up to 40% of the energy flux arrives at the divertor target before the surface divertor temperature has reached its maximum value. These large and more conductive ELMs may lead to up to ∼50% of the ELM energy reaching the main wall plasma facing components instead of the divertor target. The extrapolation to the International Thermonuclear Experimental Reactor of the obtained results is described and the main uncertainties discussed.

The ITER divertor power deposition database is described and analysed in this paper. The database contains experimental measurements from the major divertor experiments in L-mode and H-mode regimes. These measurements are used to derive multi-machine scaling laws for the peak divertor heat flux and width, particularly of their machine size dependence. The physical basis for these scalings is discussed and the laws obtained are used to extrapolate from existing experiments to the parameters expected in the ITER-EDA device.

Asymmetries are a ubiquitous feature of the scrape-off layer (SOL) and divertor plasmas in any tokamak and are thought to be driven primarily by a variety of drift flows, the directions of which reverse with reversal of the main toroidal field. The understanding of precisely how these field dependent drifts combine to yield any given experimental observation is still very much incomplete. A recent campaign of reversed field operation at JET designed to match a variety of discharges to their more frequently executed forward field counterparts has been executed in an attempt to contribute to this understanding. This paper summarises the most important findings from these experiments and includes some new EDGE2D simulation results describing the SOL flow.

Recent progress in experimental and theoretical studies of edge localized mode (ELM) physics is reviewed for the reactor relevant plasma regimes, namely the high confinement regimes, that is, H-modes and advanced scenarios.

Several small/no ELM regimes such as EDA, grassy ELM, HRS, QH-mode, type II and V ELMs with good confinement properties have been obtained in Alcator C-Mod, ASDEX-Upgrade, DIII-D, JET, JFT-2M, JT-60U and NSTX. All these regimes show considerable reduction of instantaneous ELM heat load onto divertor target plates in contrast to conventional type I ELM, and ELM energy losses are evaluated as less than 5% of the pedestal stored energy. These small/no ELM regimes are summarized and widely categorized by their pedestal conditions in terms of the operational space in non-dimensional pedestal parameters and requirement of plasma shape/configuration. The characteristics of edge fluctuations and activities of ideal MHD stability leading to small/no ELMs are also summarized.

The main objective of this paper is investigation of methods for reduction of divertor heat loads in order to increase the lifetime of divertor tiles in future fusion reactors. Special emphasis is given to studies of reduction of transient heat loads due to edge localized modes (ELMs). Two methods are compared: argon seeded type-I ELMy H-modes and nitrogen seeded type-III ELMy H-modes. In both scenarios, the impurity seeding leads to a reduction in the pedestal energy and hence a reduction in the energy released by the ELM. This consequentially reduces the power load to the divertor targets. At high radiative power fractions in type-III ELMy H-modes, part of that released ELM energy (small ELMs, below 20 kJ) is dissipated by radiation in the scrape off layer (SOL). Modelling of the ELM mitigation supports the experimental findings. This ELM mitigation by radiative dissipation is not effective for larger ELMs. In between ELMs, the plasma is detached and radiates strongly from the X-point region. During an ELM, the nitrogen in the X-point and divertor region becomes ionized into more weakly radiating higher charge states and the plasma re-attaches for large ELMs. At JET, argon radiates predominantly in the main plasma and not so much in the cold divertor region. Hence, the effect of radiative dissipation of ELM heat fluxes by argon is very low due to the limited argon density in the divertor region. Nevertheless, both scenarios might be compatible with an integrated ITER scenario, with respect to acceptable divertor lifetime and acceptable confinement.

This paper describes experiments with highly shaped JET H-mode plasmas, which were directed to developing regimes where Type I ELMs are replaced by other edge relaxations, while maintaining the pedestal pressure of Type I ELMy H-modes. It was found that Type II ELMs coexisted with Type I, up to densities of the order of the Greenwald limit, where Type III ELMs appear, and the good confinement was lost. Only at the highest edge collisionality was it observed that Type II ELMs completely replace Type I. At high βp and q95, 'grassy' ELMs replace Type I completely. The MHD spectra characteristics for grassy ELMs are significantly different from those of Type II ELMs. This paper details the experiments, briefly compares the results to those obtained elsewhere and suggests open lines of investigations for the assessment of the potential of grassy ELM regimes as an ELM mitigation technique.

Edge localized mode (ELM) measurements in many tokamaks, including ASDEX-Upgrade, DIII-D, JET, JT-60U and MAST, are reviewed, which includes progress in experimental observations at the plasma edge region by means of fast-time resolved diagnostics with high precision, such as scanning probe, radial interferometer chord, BES and tangentially viewing fast-gated camera at the midplane. ELM dynamics data show that the majority of the ELM particle and energy transport should be dominated by ion convection physics and associated timescales. Furthermore, recent diagnostic upgrades on many tokamaks reveal the ELM filament structure and their complex motion towards radial, poloidal and toroidal directions. Approaches to control the Type-I ELMs, in addition to the alternative scenarios to Type-I ELMy H-mode operation (so-called, small/no ELM regimes) are also a key area of research for current tokamaks, which demonstrated a high confinement (being comparable to that of Type-I ELMy H-mode plasmas at similar parameters) in the absence of large, ELM induced, transient heat/particle fluxes to the divertor targets. Although tolerable ELM regimes are obtained in existing devices, their application to ITER is uncertain. Issues of these regimes towards further experiments and power deposition on divertor targets and main chamber wall are discussed.

Using infra-red (IR) thermography, power loads onto the MKII Gas-Box divertor targets have been investigated in Type-I ELMy H-Mode plasmas at JET in medium current discharges (I-p = 2.6MA and B-T = 2.7 T). Heat fluxes are calculated from the measured divertor target tile surface temperatures taking into account the influence of co-deposited surface layers on tile surfaces. This is particularly important when estimating the energy deposition during transient events such as ELMs. Detailed energy balance analysis is used, both from IR and tile embedded thermocouples, to demonstrate an approximately constant ELM-averaged in/out divertor target asymmetry of approximate to 0.55 and to show that the ELM in/out energy deposition ratio ranges from 1 : 1 to 2 : 1. The inter-ELM in/out ratio is close to the ELM-averaged value at low pedestal collisionalities and decreases down to values close to zero when the inner target plasma detaches at the highest pedestal collisionalities. The fraction of ELM transported energy is observed to behave differently for the inner and the outer divertor. At higher pedestal collisionalities nearly the full inner target load is due to the ELMs whereas for the outer target the ELM transported energy never exceeds values of approximate to 0.3 of the total energy deposited there. The fraction of ELM energy arriving at the divertor compared with the pedestal loss energy in JET is found to be in the range of 0.75 for small ELMs down to 0.4 for large ELMs systematically decreasing with normalized ELM size. Since ITER is bound to use small ELMs the corresponding ELM wall load is expected to be small. The latter experimental result is in fair agreement with the observation that larger ELMs tend to travel faster across the SOL than smaller ELMs. However, a comparison of the presented data with models of ELM perpendicular transport is not conclusive due to the large experimental errorbars and uncertainties in the model assumptions.

A number of possible designs of external and in-vessel coils generating resonant magnetic perturbations (RMPs) for Type I edge localized modes (ELMs) control in ITER are analysed for the reference scenarios (H-mode, Hybrid and Steady-State) taking into account physical, technical and spatial constraints. The level of stochasticity (Chirikov parameter ∼1 at ψ 1/2 ∼ 0.95) generated by the I-coils in the DIII-D experiments on ELMs suppression was taken as a reference. Designs with a toroidal symmetry n = 3 were considered to avoid lower n numbers producing larger central islands, a potential trigger of MHD instabilities. The evaluation of RMP coils designs is done with respect to the RMPs spectrum that should produce enough edge ergodisation and minimum central perturbations at minimum current. The proposed designs include in-vessel, mid-ports and external coils. Changes in the equilibrium due to changes in the internal inductance l i , the poloidal beta β p and the edge magnetic shear in a reasonable range for ITER scenarios were demonstrated to have a small effect on the edge ergodisation. Present estimations were done without margins and for vacuum fields neglecting plasma response on RMPs. The validity of the vacuum approach was estimated analytically in the visco-resistive linear response regime using [1]. The typical radial magnetic field amplitudes produced by RMP coils in DIII-D and ITER are an order of magnitude or slightly above the critical values for the ‘downward’ bifurcation to the reconnected stage indicating the possibility of the islands formation in the pedestal region. Central islands (from the top of the pedestal) are expected to be screened.

Plasma disruptions affect plasma-facing and structural components of tokamaks due to electromechanical forces, thermal loads and generation of high energy runaway electrons (REs). Asymmetries in poloidal halo and toroidal plasma current can now be routinely measured in four positions 90° apart. Their assessment is used to validate the design of the ITER vessel support system and its in-vessel components. The challenge of disruption thermal loads comes from both the short duration over which a large energy has to be lost and the potential for asymmetries. The focus of this paper will be on localized heat loads. Resonant magnetic perturbations failed to reduce the generation of REs in JET. An explanation of the limitations applying to these attempts is offered together with a minimum guideline. The REs generated by a moderate, but fast, Ar injection in limiter plasmas show evidence of milder and more efficient losses due to the high Ar background density.

The poloidal field (PF) coil system on ITER, which provides both feedforward and feedback control of plasma position, shape, and current, is a critical element for achieving mission performance. Analysis of PF capabilities has focused on the 15 MA Q = 10 scenario with a 300–500 s flattop burn phase. The operating space available for the 15 MA ELMy H-mode plasma discharges in ITER and upgrades to the PF coils or associated systems to establish confidence that ITER mission objectives can be reached have been identified. Time dependent self-consistent free-boundary calculations were performed to examine the impact of plasma variability, discharge programming and plasma disturbances. Based on these calculations a new reference scenario was developed based upon a large bore initial plasma, early divertor transition, low level heating in L-mode and a late H-mode onset. Static equilibrium analyses for this scenario, which determine PF coil currents to produce a given plasma configuration, indicate that the original PF coil limitations do not allow low li(<0.8) operation or plasmas with lower flux consumption, and the flattop burn durations were predicted to be less than the desired 400 s. This finding motivates the expansion of the operating space, considering several upgrade options to the PF coils. Analysis was also carried out to examine the feedback current reserve required in the central solenoid and PF coils during a series of disturbances, heating and current drive sources for saving volt-seconds in rampup, a feasibility assessment of the 17 MA scenario was undertaken, and the rampdown phase of the discharge is discussed. Results of the studies show that the new scenario and modified PF system will allow a wide range of 15 MA 300–500 s operation and more limited but finite 17 MA operation.

The paper concerns experimental investigations of plasma facing components erosion under the plasma heat loads expected in ITER divertor during transient events such as the Type I Edge-Localized Modes and the disruptions. The experiments were carried out at the TRINITI plasma gun QSPA-T. The carbon fiber composite and tungsten macrobrush targets designed for ITER were exposed to multiple plasma pulses of duration 0.5 ms and deposited energy in the range of 0.2–2.5 MJ/m2. Between some of the pulses the eroded surface was analyzed with profilometric measurements and electron microscopy. The CFC erosion is determined mainly by damages to the PAN-fibers. While the energy increases from 0.2 to 2.4 MJ/m2 the removed layer of PAN-fibers area increases from 0.01 to 10 μm per pulse. The erosion of tungsten (pure and lanthanum oxide-doped tungsten) is shown to be determined mainly by crack formation, melt layer movement and droplets ejection.

Edge localized mode (ELM) triggering by pellet injection in the DIII-D tokamak has been simulated with the non-linear MHD code JOREK with a view to validating its physics models. JOREK has been subsequently applied to evaluate the requirements for ELM control by pellet injection in ITER. JOREK modelling results for DIII-D show that the key parameter for the triggering of ELMs by pellets is the value of the localized pressure perturbation caused by pellet injection which leads to a threshold minimum pellet size for a given injection velocity, injection geometry and H-mode plasma characteristics. The minimum pellet size for ELM triggering is found to depend on injection geometry with the largest value being required for injection at the outer midplane, intermediate for injection near the X-point and the smallest one for injection at the high-field side. The first results of studies for ELM triggering by pellet injection in ITER 15 MA Q = 10 plasmas with the foreseen injection geometry in ITER are presented.

Self-consistent transport simulation of ITER scenarios is a very important tool for the exploration of the operational space and for scenario optimisation. It also provides an assessment of the compatibility of developed scenarios (which include fast transient events) with machine constraints, in particular with the poloidal field (PF) coil system, heating and current drive (H&CD), fuelling and particle and energy exhaust systems. This paper discusses results of predictive modelling of all reference ITER scenarios and variants using two suite of linked transport and equilibrium codes. The first suite consisting of the 1.5D core/2D SOL code JINTRAC [1] and the free boundary equilibrium evolution code CREATE-NL [2,3], was mainly used to simulate the inductive D-T reference Scenario-2 with fusion gain Q=10 and its variants in H, D and He (including ITER scenarios with reduced current and toroidal field). The second suite of codes was used mainly for the modelling of hybrid and steady state ITER scenarios. It combines the 1.5D core transport code CRONOS [4] and the free boundary equilibrium evolution code DINA-CH [5].

A path to a new high performance regime has been discovered in tokamaks that could improve the attractiveness of a fusion reactor. Experiments on DIII-D using a quiescent H-mode edge have navigated a valley of improved edge peeling-ballooning stability that opens up with strong plasma shaping at high density, leading to a doubling of the edge pressure over the standard H mode with edge localized modes at these parameters. The thermal energy confinement time increases as a result of both the increased pedestal height and improvements in the core transport and reduced low-k turbulence. Calculations of the pedestal height and width as a function of density using constraints imposed by peeling-ballooning and kinetic-ballooning theory are in quantitative agreement with the measurements.

The magnetic energy balance and magnetic energy flows for plasma disruptions in which runaway plateau plasmas are formed and terminated at JET has been analysed and compared with that of runaway-free disruptions. The analysis shows that the energy loss processes during runaway plateau plasma termination are qualitatively different from those of a runaway-free disruption because of the pre-existence of a runaway population in the first case. As a consequence, a significant fraction of the runaway plateau plasma magnetic energy is directly converted into runaway electron kinetic energy during the runaway plateau termination phase. This leads to the fluxes being deposited by runaway electrons onto in-vessel components during the termination of runaway plateaus to be significantly larger than those expected from the initial kinetic energy of the runaway electrons in the runaway plateau plasma.

For the first time, DIII-D experiments have achieved stationary quiescent H-mode (QH-mode) operation for many energy confinement times at simultaneous ITER-relevant values of beta, confinement, and safety factor, in an ITER-like shape. QH-mode provides excellent energy confinement, even at very low plasma rotation, while operating without edge localized modes (ELMs) and with strong impurity transport via the benign edge harmonic oscillation (EHO). By tailoring the plasma shape to improve the edge stability, the QH-mode operating space has also been extended to densities exceeding 80% of the Greenwald limit, overcoming the long-standing low-density limit of QH-mode operation. In the theory, the density range over which the plasma encounters the kink-peeling boundary widens as the plasma cross-section shaping is increased, thus increasing the QH-mode density threshold. The DIII-D results are in excellent agreement with these predictions, and nonlinear magnetohydrodynamic analysis of reconstructed QH-mode equilibria shows unstable low n kink-peeling modes growing to a saturated level, consistent with the theoretical picture of the EHO. Furthermore, high density operation in the QH-mode regime has opened a path to a new, previously predicted region of parameter space, named “Super H-mode” because it is characterized by very high pedestals that can be more than a factor of two above the peeling-ballooning stability limit for similar ELMing H-mode discharges at the same density.

An extensive study of intrinsic and controlled non-axisymmetric field (δB) impacts in KSTAR has enhanced the understanding about non-axisymmetric field physics and its implications, in particular, on resonant magnetic perturbation (RMP) physics and power threshold (Pth) for L–H transition. The n = 1 intrinsic non-axisymmetric field in KSTAR was measured to remain as low as δB/B0 ~ 4 × 10−5 even at high-beta plasmas (βN ~ 2), which corresponds to approximately 20% below the targeted ITER tolerance level. As for the RMP edge-localized-modes (ELM) control, robust n = 1 RMP ELM-crash-suppression has been not only sustained for more than ~90 τE, but also confirmed to be compatible with rotating RMP. An optimal window of radial position of lower X-point (i.e. Rx = m) proved to be quite critical to reach full n = 1 RMP-driven ELM-crash-suppression, while a constraint of the safety factor could be relaxed (q95 = 5 0.25). A more encouraging finding was that even when Rx cannot be positioned in the optimal window, another systematic scan in the vicinity of the previously optimal Rx allows for a new optimal window with relatively small variations of plasma parameters. Also, we have addressed the importance of optimal phasing (i.e. toroidal phase difference between adjacent rows) for n = 1 RMP-driven ELM control, consistent with an ideal plasma response modeling which could predict phasing-dependent ELM suppression windows. In support of ITER RMP study, intentionally misaligned RMPs have been found to be quite effective during ELM-mitigation stage in lowering the peaks of divertor heat flux, as well as in broadening the 'wet' areas. Besides, a systematic survey of Pth dependence on non-axisymmetric field has revealed the potential limit of the merit of low intrinsic non-axisymmetry. Considering that the ITER RMP coils are composed of 3-rows, just like in KSTAR, further 3D physics study in KSTAR is expected to help us minimize the uncertainties of the ITER RMP coils, as well as establish an optimal 3D configuration for ITER and future reactors.

A decade-long operation of the Korean Superconducting Tokamak Advanced Research (KSTAR) has contributed significantly to the operation of superconducting tokamak devices and the advancement of tokamak physics which will be beneficial for the ITER and K-DEMO programs. Even with limited heating capability, various conventional as well as new operating regimes have been explored and have achieved improved performance. As examples, a long pulse high-confinement mode operation with and without an edge-localized mode (ELM) crash was well over 70 and 30 s, respectively. The unique capabilities of KSTAR allowed it to improve the capability of controlling harmful instabilities, and they have been instrumental in uncovering much new physics. The highlights are that the L/H transition threshold power is sensitive to the resonant magnetic perturbation (RMP) and insensitive to non-resonant magnetic perturbation. Co-Ip offset rotation dominated by an electron channel predicted by general neoclassical toroidal viscosity theory was confirmed. Improved heat dispersal in a divertor system using three rows of rotating RMP was demonstrated and predictive control of the ELM-crash with a priori modeling was successfully tested. In magnetohydrodynamic physics, validation of the full reconnection model (i.e. q0 > 1 right after the sawtooth crash) and self-consistent validation of the anisotropic distribution of turbulence amplitude and flow in the presence of the 2/1 island with theoretical models were achieved. The turbulence amplitude induced by RMP was linearly increased with the slow RMP coil current ramp-up time (i.e. the magnetic diffusion time scale). The Dα spikes (i.e. ELM-crash amplitude) was linearly decreased with the turbulence amplitude and not correlated with the perpendicular electron flow. In the turbulence area, a non-diffusive 'avalanche' transport event and the role of a quiescent coherent mode in confinement were studied. To accommodate the anticipation of a higher performance of the KSTAR plasmas with the increased heating powers, a new divertor/internal interface with a full active cooling system will be implemented after a full test of the new heating (neutral beam injection II and electron cyclotron heating) and current drive (CD) (Helicon and lower hybrid CD) systems. An upgrade plan for the internal hardware, heating systems and efficient CD system may allow for a long pulse operation of higher performance plasmas at βN > 3.0 with f bs ~ 0.5 and Ti > 10 keV.

The electron temperature and density pedestals tend to vary in their relative radial positions, as observed in DIII-D (Beurskens et al 2011 Phys. Plasmas 18 056120) and ASDEX Upgrade (Dunne et al 2017 Plasma Phys. Control. Fusion 59 14017). This so-called relative shift has an impact on the pedestal magnetohydrodynamic (MHD) stability and hence on the pedestal height (Osborne et al 2015 Nucl. Fusion 55 063018). The present work studies the effect of the relative shift on pedestal stability of JET ITER-like wall (JET-ILW) baseline low triangularity (δ) unseeded plasmas, and similar JET-C discharges. As shown in this paper, the increase of the pedestal relative shift is correlated with the reduction of the normalized pressure gradient, therefore playing a strong role in pedestal stability. Furthermore, JET-ILW tends to have a larger relative shift compared to JET carbon wall (JET-C), suggesting a possible role of the plasma facing materials in affecting the density profile location. Experimental results are then compared with stability analysis performed in terms of the peeling-ballooning model and with pedestal predictive model EUROPED (Saarelma et al 2017 Plasma Phys. Control. Fusion). Stability analysis is consistent with the experimental findings, showing an improvement of the pedestal stability, when the relative shift is reduced. This has been ascribed mainly to the increase of the edge bootstrap current, and to minor effects related to the increase of the pedestal pressure gradient and narrowing of the pedestal pressure width. Pedestal predictive model EUROPED shows a qualitative agreement with experiment, especially for low values of the relative shift.

Optimal strategies for disruption mitigation benefit from the understanding of details both spatially and temporally. Beyond the assessment of the efficacy of a particular proposed Disruption Mitigation System (DMS), ITER's longevity will require accounting of both mitigated and unmitigated disruptions. Accurate models and validated simulations that detail multiple ITER scenarios with mitigated and unmitigated disruptions are essential for accurate estimates of load damage. The primary candidate for ITER's DMS is Shattered Pellet Injection (SPI); its efficacy must be evaluated within the next several years. To perform critical time dependent 3-D nonlinear simulations, we have developed a particle based SPI model in the NIMROD code coupled to its modified single fluid equations with impurity and radiation [Izzo, Nucl. Fusion 46(5), 541 (2006)]. SPI validation simulations of the thermal quench and comparisons to DIII-D impurity scan experiments [Shiraki et al., Phys. Plasmas 23(6), 062516 (2016)] are presented. We also present an initial ITER Q = 10 pure neon SPI simulation and compare it with the DIII-D SPI simulations. NIMROD SPI simulations demonstrate that the ablating fragment drives strong parallel flows that transport the impurities and governs the thermal quench. Analysis of SPI simulations shows that the mixed deuterium/neon SPI results in a more benign thermal quench due to the enhanced transport caused by the additional deuterium. These results suggest that an optimal pellet mixture exists for the SPI system.

Full suppression of type-I edge localized modes (ELMs) using n = 4 resonant magnetic perturbations (RMPs) as planned for ITER has been demonstrated for the first time (n is the toroidal mode number of the applied RMP). This is achieved in EAST plasmas with low input torque and tungsten divertor, and the target plasma for these experiments in EAST is chosen to be relevant to the ITER Q = 10 operational scenario, thus also addressing significant scenario issues for ITER. In these experiments the lowest neutral beam injection (NBI) input torque is around TNBI ∼ 0.44 Nm, which extrapolates to around 14 Nm in ITER (compared to a total torque input of 35 Nm when 33 MW of NBI are used for heating). The q95 is around 3.6 and normalized plasma beta βN ∼ 1.5–1.8, similar to that in the ITER Q = 10 scenario. Suppression windows in both q95 and plasma density are observed; in addition, lower plasma rotation is found to be favourabe to access ELM suppression. ELM suppression is maintained with line averaged density up to 60%nGW (Greenwald density limit) by feedforward gas fuelling after suppression is achieved. It is interesting to note that in addition to an upper density, a low density threshold for ELM suppression of 40%nGW is also observed. In these conditions energy confinement does not significantly drop (<10%) during ELM suppression when compared to the ELMy H-mode conditions, which is much better than previous results using low n (n = 1 and 2) RMPs in higher q95 regimes. In addition, the core plasma tungsten concentration is clearly reduced during ELM suppression demonstrating an effective impurity exhaust. MHD response modelling using the MARS-F code shows that edge magnetic field stochasticity has a peak at q95 ∼ 3.65 for the odd parity configuration, which is consistent to the observed suppression window around 3.6–3.75. These results expand the physical understanding of ELM suppression and demonstrate the effectiveness of n = 4 RMPs for reliable control ELMs in future ITER high Q plasma scenarios with minimum detrimental effects on plasma confinement.

Understanding and predicting divertor heat-load width ${\lambda}_q$ is a critically important problem for an easier and more robust operation of ITER with high fusion gain. Previous predictive simulation data for ${\lambda}_q$ using the extreme-scale edge gyrokinetic code XGC1 in the electrostatic limit under attached divertor plasma conditions in three major US tokamaks [C.S. Chang et al., Nucl. Fusion 57, 116023 (2017)] reproduced the Eich and Goldston attached-divertor formula results [formula #14 in T. Eich et al., Nucl. Fusion 53, 093031 (2013); R.J. Goldston, Nucl. Fusion 52, 013009 (2012)], and furthermore predicted over six times wider ${\lambda}_q$ than the maximal Eich and Goldston formula predictions on a full-power (Q = 10) scenario ITER plasma. After adding data from further predictive simulations on a highest current JET and highest-current Alcator C-Mod, a machine learning program is used to identify a new scaling formula for ${\lambda}_q$ as a simple modification to the Eich formula #14, which reproduces the Eich scaling formula for the present tokamaks and which embraces the wide ${\lambda}_q^X{GC}$ for the full-current Q = 10 ITER plasma. The new formula is then successfully tested on three more ITER plasmas: two corresponding to long burning scenarios with Q = 5 and one at low plasma current to be explored in the initial phases of ITER operation. The new physics that gives rise to the wider ${\lambda}q_^{XGC} is identified to be the weakly-collisional, trapped-electron-mode turbulence across the magnetic separatrix, which is known to be an efficient transporter of the electron heat and mass. Electromagnetic turbulence and high-collisionality effects on the new formula are the next study topics for XGC1.

Abstract We present an overview of results from a series of L–H transition experiments undertaken at JET since the installation of the ITER-like-wall (JET-ILW), with beryllium wall tiles and a tungsten divertor. Tritium, helium and deuterium plasmas have been investigated. Initial results in tritium show ohmic L–H transitions at low density and the power threshold for the L–H transition ( P LH ) is lower in tritium plasmas than in deuterium ones at low densities, while we still lack contrasted data to provide a scaling at high densities. In helium plasmas there is a notable shift of the density at which the power threshold is minimum ( <?CDATA ${\bar{n}}_{\mathrm{e},\mathrm{min}}$?> <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" display="inline" overflow="scroll"> <mml:msub> <mml:mrow> <mml:mover accent="true"> <mml:mrow> <mml:mi>n</mml:mi> </mml:mrow> <mml:mo>¯</mml:mo> </mml:mover> </mml:mrow> <mml:mrow> <mml:mi mathvariant="normal">e</mml:mi> <mml:mo>,</mml:mo> <mml:mi>min</mml:mi> </mml:mrow> </mml:msub> </mml:math> ) to higher values relative to deuterium and hydrogen references. Above <?CDATA ${\bar{n}}_{\mathrm{e},\mathrm{min}}$?> <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" display="inline" overflow="scroll"> <mml:msub> <mml:mrow> <mml:mover accent="true"> <mml:mrow> <mml:mi>n</mml:mi> </mml:mrow> <mml:mo>¯</mml:mo> </mml:mover> </mml:mrow> <mml:mrow> <mml:mi mathvariant="normal">e</mml:mi> <mml:mo>,</mml:mo> <mml:mi>min</mml:mi> </mml:mrow> </mml:msub> </mml:math> (He) the L–H power threshold at high densities is similar for D and He plasmas. Transport modelling in slab geometry shows that in helium neoclassical transport competes with interchange-driven transport, unlike in hydrogen isotopes. Measurements of the radial electric field in deuterium plasmas show that E r shear is not a good indicator of proximity to the L–H transition. Transport analysis of ion heat flux in deuterium plasmas show a non-linearity as density is decreased below <?CDATA ${\bar{n}}_{\mathrm{e},\mathrm{min}}$?> <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" display="inline" overflow="scroll"> <mml:msub> <mml:mrow> <mml:mover accent="true"> <mml:mrow> <mml:mi>n</mml:mi> </mml:mrow> <mml:mo>¯</mml:mo> </mml:mover> </mml:mrow> <mml:mrow> <mml:mi mathvariant="normal">e</mml:mi> <mml:mo>,</mml:mo> <mml:mi>min</mml:mi> </mml:mrow> </mml:msub> </mml:math> . Lastly, a regression of the JET-ILW deuterium data is compared to the 2008 ITPA scaling law.

Complete 3D simulations of the current quench phase of ITER disruptions are key to predict asymmetric forces acting into the ITER wall. We present for the first time such simulations for ITER mitigated disruptions at realistic Lundquist numbers. For these strongly mitigated disruptions, we find that the edge safety factor remains above 2 and the maximal integral horizontal forces remain below 1 MN. The maximal integral vertical force is found to be 13 MN and arises in a time scale given by the resistive wall time as expected from theoretical considerations. In this respect, the vertical force arises after the plasma current has completely decayed, showing the importance of continuing the simulations also in the absence of plasma current. We conclude that the horizontal wall force rotation is not a concern for these strongly mitigated disruptions in ITER, since when the wall forces form, there are no remaining sources of rotation.

The prediction of power fluxes and plasma-wall interactions impacted by MHD processes during ITER operation [disruption, Edge Localized Modes (ELMs), 3D magnetic fields applied for ELM control, etc.] requires models that include an accurate description of the MHD processes themselves, as well as of the edge plasma and plasma-wall interaction processes. In this paper, we report progress on improving the edge plasma physics models in the nonlinear extended MHD code JOREK, which has capabilities to simulate the MHD response of the plasma to the applied external 3D fields, disruptions and ELMs. The extended MHD model includes E × B drifts, diamagnetic drifts, and neoclassical flows. These drifts can have large influences, on e.g., divertor asymmetries. Realistic divertor conditions are important for impurity sputtering, transport, and their effect on the plasma. In this work, we implemented kinetic and fluid neutral physics modules, investigated the influence of poloidal flows under divertor conditions in the ITER PFPO-1 (1.8T/5MA) H-mode plasma scenario, and compared the divertor plasma conditions and heat flux to the wall for both the fluid and kinetic neutral model (in JOREK) to the well-established 2D boundary plasma simulation code suite SOLPS-ITER. As an application of the newly developed model, we investigated time-dependent divertor solutions and the transition from attached to partially detached plasmas. We present the formation of a high-field-side high-density-region and how it is driven by poloidal E × B drifts.

Abstract The recent deuterium–tritium campaign in JET-ILW (DTE2) has provided a unique opportunity to study the isotope dependence of the L-H power threshold in an ITER-like wall environment (Be wall and W divertor). Here we present results from dedicated L-H transition experiments at JET-ILW, documenting the power threshold in tritium and deuterium–tritium plasmas, comparing them with the matching deuterium and hydrogen datasets. From earlier experiments in JET-ILW it is known that as plasma isotopic composition changes from deuterium, through varying deuterium/hydrogen concentrations, to pure hydrogen, the value of the line averaged density at which the threshold is minimum, <?CDATA ${\bar n_{{\text{e}},{\text{min}}}}$?> <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" overflow="scroll"><mml:mrow><mml:msub><mml:mover><mml:mi>n</mml:mi><mml:mo>ˉ</mml:mo></mml:mover><mml:mrow><mml:mrow><mml:mtext>e</mml:mtext></mml:mrow><mml:mo>,</mml:mo><mml:mrow><mml:mtext>min</mml:mtext></mml:mrow></mml:mrow></mml:msub></mml:mrow></mml:math> , increases, leading us to expect that <?CDATA ${\bar n_{{\text{e}},{\text{min}}}}$?> <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" overflow="scroll"><mml:mrow><mml:msub><mml:mover><mml:mi>n</mml:mi><mml:mo>ˉ</mml:mo></mml:mover><mml:mrow><mml:mrow><mml:mtext>e</mml:mtext></mml:mrow><mml:mo>,</mml:mo><mml:mrow><mml:mtext>min</mml:mtext></mml:mrow></mml:mrow></mml:msub></mml:mrow></mml:math> (T) &lt; <?CDATA ${\bar n_{{\text{e}},{\text{min}}}}$?> <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" overflow="scroll"><mml:mrow><mml:msub><mml:mover><mml:mi>n</mml:mi><mml:mo>ˉ</mml:mo></mml:mover><mml:mrow><mml:mrow><mml:mtext>e</mml:mtext></mml:mrow><mml:mo>,</mml:mo><mml:mrow><mml:mtext>min</mml:mtext></mml:mrow></mml:mrow></mml:msub></mml:mrow></mml:math> (DT) &lt; <?CDATA ${\bar n_{{\text{e}},{\text{min}}}}$?> <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" overflow="scroll"><mml:mrow><mml:msub><mml:mover><mml:mi>n</mml:mi><mml:mo>ˉ</mml:mo></mml:mover><mml:mrow><mml:mrow><mml:mtext>e</mml:mtext></mml:mrow><mml:mo>,</mml:mo><mml:mrow><mml:mtext>min</mml:mtext></mml:mrow></mml:mrow></mml:msub></mml:mrow></mml:math> (D) &lt; <?CDATA ${\bar n_{{\text{e}},{\text{min}}}}$?> <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" overflow="scroll"><mml:mrow><mml:msub><mml:mover><mml:mi>n</mml:mi><mml:mo>ˉ</mml:mo></mml:mover><mml:mrow><mml:mrow><mml:mtext>e</mml:mtext></mml:mrow><mml:mo>,</mml:mo><mml:mrow><mml:mtext>min</mml:mtext></mml:mrow></mml:mrow></mml:msub></mml:mrow></mml:math> (H). The new power threshold data confirms these expectations in most cases, with the corresponding ordering of the minimum power thresholds. We present a comparison of this data to power threshold scalings, used for extrapolation to future devices such as ITER and DEMO.

Results of experiments investigating the performance of the JET Mark IIA divertor are reported and compared with the performance of its Mark I predecessor. The principal effect of reducing the divertor width (increasing closure) was to increase pumping for both deuterium and impurities while reducing upstream neutral pressure. Neither the orientation of the divertor target relative to the divertor plasma nor the width of the divertor had a major influence on core plasma performance in ELMy H modes. Changing the core triangularity and thus the edge magnetic shear modifies the ELM frequency in ELMy H mode plasmas, thereby changing the peak divertor power loading. The integrated performance of the core and divertor plasmas is reviewed with a view to extrapolation to the requirements of ITER. The confinement of JET ELMy H modes with hot, medium density edges is good (H97 ≈ 1) and follows a gyro-Bohm scaling. The impurity content of these discharges is low and within the ITER requirements. When an attempt is made to raise the density with deuterium gas fuelling, the ELM frequency increases and the confinement, especially in the edge, decreases. Good confinement can be achieved in JET either by producing a large edge pedestal, typically in discharges with NB heating or by centrally peaked heating with ICRH schemes. Large amplitude type I ELMs, which are present in all discharges with a large edge pedestal, would result in unacceptable divertor plate erosion when scaled to ITER. Since the power deposition profile due to α heating in ITER is calculated to be intermediate between the JET NB and RF heating profiles, it is likely that operation in ITER with small ELMs in order to reduce first wall loading will result in degraded confinement compared with present day scaling laws.

A comparative analysis of the spatial and temporal characteristics of transient energy loads (ELMs and disruptions) on plasma facing components (PFCs) in present tokamak devices and their extrapolation to next step devices is presented. Type I ELMs lead to the expulsion of energy by the plasma in helical structures with ballooning-like features and toroidal numbers in the range n = 10–15. The plasma energy is transported towards the divertor and the main chamber PFCs leading to significant transient energy loads at these two locations on small wetted area. The largest transient energy fluxes onto PFCs in tokamaks are measured during the thermal quench of disruptions. These fluxes do not exceed greatly those of large Type I ELMs, due to the much larger wetted area for energy flux during the thermal quench compared to Type I ELMs. The implications of these findings for the next step devices are discussed.

Experimental results on divertor heat load measurements from ASDEX Upgrade and JET are discussed. Thereby three topics are considered: (i) parameter dependence of steady state heat flux profiles, (ii) spatial distribution of the heat flux profile during type I edge localised modes (ELMs), and (iii) temporal evolution of the energy deposition during type I ELMs. No clear scaling of steady state heat flux profiles with plasma parameters is found. For different data sets a broadening of the heat flux profiles, a constant profile width, as well as a steepening with heating power is found. Extrapolation to ITER requires a review of the data. The heat flux profile is not significantly broadened during type I ELMs. Advantageous is the change of the in/out symmetry. The temporal behaviour of the energy deposition shows a strong increase of the heat flux on time scales of the ion sound speed and an exponential decay with about twice the rise time.

This paper presents the results of JET experiments aimed at studying the operational space of plasmas with a Type III ELMy edge, in terms of both local and global plasma parameters. In JET, the Type III ELMy regime has a wide operational space in the pedestal ne – Te diagram, and Type III ELMs are observed in standard ELMy H-modes as well as in plasmas with an internal transport barrier (ITB). The transition from an H-mode with Type III ELMs to a steady state Type I ELMy H-mode requires a minimum loss power, PTypeI. PTypeI decreases with increasing plasma triangularity. In the pedestal ne – Te diagram, the critical pedestal temperature for the transition to Type I ELMs is found to be inversely proportional to the pedestal density (Tcrit ∝ 1/n) at a low density. In contrast, at a high density, Tcrit, does not depend strongly on density. In the density range where Tcrit ∝ 1/n, the critical power required for the transition to Type I ELMs decreases with increasing density. Experimental results are presented suggesting a common mechanism for Type III ELMs at low and high collisionality. A single model for the critical temperature for the transition from Type III to Type I ELMs, based on the resistive interchange instability with magnetic flutter, fits well the density and toroidal field dependence of the JET experimental data. On the other hand, this model fails to describe the variation of the Type III ne – Te operational space with isotopic mass and q95. Other results are instead suggestive of a different physics for Type III ELMs. At low collisionality, plasma current ramp experiments indicate a role of the edge current in determining the transition from Type III to Type I ELMs, while at high collisionality, a model based on resistive ballooning instability well reproduces, in term of a critical density, the experimentally observed q95 dependence of the transition from Type I to Type III ELMs. Experimental evidence common to Type III ELMs in standard ELMy H-modes and in plasmas with ITBs indicates that they are driven by the same instability.

Material erosion, transport and deposition in the divertor of ITER are modelled with the Monte-Carlo impurity transport code ERO taking into account chemical erosion, physical sputtering, enhanced chemical erosion of redeposited carbon and a beryllium influx from main chamber erosion. The continuous deposition of beryllium leads to reduced carbon erosion along the divertor plates with increasing exposure time. With 1% beryllium in the edge plasma an upper value of the long-term tritium retention rate can be estimated to about 15.9 mg T/s. For 0.1% beryllium this number decreases to about 6.4 mg T/s. These numbers do not change significantly with the sticking assumption for hydrocarbons. The erosion of the divertor plates is less critical. Maximal erosion rates of 0.4 nm/s with 1% beryllium and 1.8 nm/s with 0.1% beryllium occur at the outer target. Erosion due to transient heat loads is not yet included in the modelling.

The Experimental Advanced Superconducting Tokamak (EAST) has demonstrated, for the first time, long-pulse divertor plasmas over 400 s, entirely driven by lower hybrid current drive (LHCD), and further extended high-confinement plasmas, i.e. H-modes, over 30 s with predominantly LHCD and advanced lithium wall conditioning. Many new and exciting physics results have been obtained in the quest for long-pulse operations. The key findings are as follows: (1) access to H-modes in EAST favours the divertor configuration with the ion ∇B drift directed away from the dominant X-point; (2) divertor asymmetry during edge-localized modes (ELMs) also appears to be dependent on the toroidal field direction, with preferential particle flow opposite to the ion ∇B drift; (3) LHCD induces a striated heat flux (SHF), enhancing heat deposition away from the strike point, and the degree of SHF can be modified by supersonic molecule beam injection; (4) the long-pulse H-modes in EAST exhibit a confinement quality between type-I and type-III ELMy H-modes, with H98(y,2) ∼ 0.9, similar to type-II ELMy H-modes.

In this paper, the nature of the primary instability present in the pedestal forming region prior to the transition into H mode is analysed using a gyrokinetic code on JET-ILW profiles. The linear analysis shows that the primary instability is of resistive nature, and can therefore be stabilized by increased temperature, hence power. The unstable modes are identified as being resistive ballooning modes. Their growth rates decrease for temperatures increasing towards the experimentally measured temperature at the L–H transition. The growth rates are larger for lower effective charge Zeff. This dependence is shown to be in qualitative agreement with recent and past experimental observations of reduced Zeff associated with lower L–H power thresholds.

Edge Localised Modes (ELMs) in ITER Q = 10 H-mode plasmas are likely to lead to large transient heat loads to the divertor. To avoid an ELM induced reduction of the divertor lifetime, the large ELM energy losses need to be controlled. In ITER, ELM control is foreseen using magnetic field perturbations created by in-vessel coils and the injection of small D2 pellets. ITER plasmas are characterised by low collisionality at a high density (high fraction of the Greenwald density limit). These parameters cannot simultaneously be achieved in current experiments. Therefore, the extrapolation of the ELM properties and the requirements for ELM control in ITER relies on the development of validated physics models and numerical simulations. In this paper, we describe the modelling of ELMs and ELM control methods in ITER. The aim of this paper is not a complete review on the subject of ELM and ELM control modelling but rather to describe the current status and discuss open issues.

This paper describes the progress in analysis of the ITER plasma vertical stabilisation (VS) system since its design review in 2007–2008. Two indices characterising plasma VS were studied. These are (1) the maximum value of plasma vertical displacement due to free drift that can be stopped by the VS system and (2) the maximum root mean square value of low frequency noise in the dZ/dt measurement signal used in the VS feedback loop. The first VS index was calculated using the PET code for 15 MA plasmas with the nominal position and shape. The second VS index was studied with the DINA code in the most demanding simulations for plasma magnetic control of 15 MA scenarios with the fastest plasma current ramp-up and early X-point formation, the fastest plasma current ramp-down in a divertor configuration, and an H to L mode transition at the current flattop. The studies performed demonstrate that the VS in-vessel coils, adopted recently in the baseline design, significantly increase the range of plasma controllability in comparison with the stabilising systems VS1 and VS2, providing operating margins sufficient to achieve ITER's goals specified in the project requirements.

Abstract Fuelling requirements for ITER are analysed in relation to pellet fuelling and ELM pacing, and a divertor power load control consistent with the ITER pumping and fuel throughput capabilities. The plasma parameters at the separatrix and the particle sources are derived from scalings based on SOLPS simulations. Effective transport coefficients in the H-mode pedestal are derived from EPED1 + SOLPS scalings for the pedestal height and width. 1.5D transport is simulated in the ASTRA framework. The operating window for ITER DT plasmas with the required fusion performance and level of ELM, and divertor power load control compatible with ITER fuelling and pumping capabilities, is determined. It is shown that the flexibility of the ITER fuelling systems, comprising pellet and gas injection systems, enables operation with Q = 10, which was found to be marginal in previous studies following a similar approach but with different assumptions. The present assessment shows that a reduction of <?CDATA $\langle {{n}_{e}}\rangle $ ?> <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" overflow="scroll"> <mml:mstyle displaystyle="false"> <mml:mrow> <mml:mrow> <mml:mo>〈</mml:mo> <mml:mrow> <mml:mstyle displaystyle="false"> <mml:msub> <mml:mrow> <mml:mi>n</mml:mi> </mml:mrow> <mml:mi>e</mml:mi> </mml:msub> </mml:mstyle> </mml:mrow> <mml:mo>〉</mml:mo> </mml:mrow> </mml:mrow> </mml:mstyle> </mml:math> by a factor ~2 (from 9 to 5 × 10 19 m −3 ) in 15 MA H-mode plasmas leads to a reduction in the required pellet fuelling rate by a factor of four. Results of the analysis of the fuelling requirements for a range of ITER scenarios are found to be similar to those obtained with the JINTRAC code that included 2D modelling of the edge plasma.

Abstract Long Pulse Scenarios (LPS) in ITER foreseen during the Pre-Fusion Power Operation (PFPO) phase of the ITER Research Plan (IRP) are assessed using 1.5D transport simulations within the ASTRA framework. Such assessment is required to predict the operational space for LPS operation in PFPO, as well as to evaluate which physics processes for LPS operation during Fusion Power Operation (FPO) could be studied during PFPO. An important aspect in the development of LPSs in PFPO is to minimize lifetime consumption of the Central Solenoid (CS) for these scenarios. The maximum pulse length achievable for LPSs in PFPO with no consumption of CS lifetime (currents in CS coils ⩽30 kA per turn) has been assessed for a range of heating schemes and heating mixes, confinement regimes (L-mode and H-mode) and for helium and hydrogen plasmas. The operational space of LPS and pulse length has been explored through density scans with the Heating and Current Drive mix required for the FPO Q ⩾ 5 steady-state plasma scenario (namely Neutral Beam Injection and Electron Cyclotron Heating) including acceptable shine through losses on the first wall for both helium and hydrogen plasmas. Fast particle physics aspects that are common between FPO plasmas and LPS PFPO H-mode plasmas at low densities are studied including MHD stability analysis with the KINX code and non-perturbative critical gradient model based on high-n Toroidal Alfven Eigenmodes (TAE) stability kinetic ballooning code HINST calculations.

In current large tokamaks, non-intrinsic seeded impurities have been used to produce divertor power loads which would be considered acceptable when extrapolated to ITER. Many devices have achieved the goals of high fractional radiated powers, small frequent ELMs and detachment which are characteristic of radiative H mode regimes. The influence of divertor geometry on these characteristics is described. It has been a matter of concern that the Zeff associated with the seeded impurities may exceed that allowable in ITER and also that the degradation in energy confinement may be unacceptable. Confidence can only be built in the prediction of these parameters in ITER if reliable scalings are available for impurity content and energy confinement which have a sound physics basis. Work is described at JET in this area whilst using multimachine data to characterize the size scaling and provide a context for the JET data. Predicted levels for the impurity content of seeded ITER plasmas appear to be of marginal acceptability. Discharges run in the JET Mark I, Mark IIA and Mark IIAP divertors are compared and indicate that increased divertor closure has brought relatively minor benefits in highly radiative discharges. The acceptability of the energy confinement of radiation for ITER remains unclear. Dimensionless parameter scaling experiments have been conducted in which β, q25, fractional radiated power and Zeff are held constant for a range of ρ*. The price paid for high edge radiation and small ELMs appears to be a 25% loss in total stored energy as a result of edge pedestal degradation. However, the underlying energy confinement scaling may still be consistent with gyro-Bohm scaling, which would give an adequate margin for ITER. This conclusion is, however, sensitive to the scaling of confinement with collisionality, which is difficult to determine due to the coupling between ρ* and ν* which is a consequence of radiation dominated regimes.

Nine type-I ELMy H-mode discharges in diagnostic optimized configuration in JET are analysed with the EDGE2D/NIMBUS package. EDGE2D solves the fluid equations for the conservation of particles, momentum and energy for hydrogenic and impurity ions, while neutrals are followed with the two-dimensional Monte Carlo module NIMBUS. Using external boundary conditions from the experiment, the perpendicular heat conductivities χi,e and the particle transport coefficients D, v are varied until good agreement between code result and measured data is obtained. A step-like ansatz is used for the edge transport parameters for the outer core region, the edge transport barrier and the outer scrape-off layer. The time-dependent effect of edge localized modes on the edge profiles is simulated with an ad hoc ELM model based on the repetitive increase of the transport coefficients χi,e and D. The values of the transport coefficients are matched to experimental data mapped to the outer midplane, in the course of which radial shifts of experimental profiles of the order of 1 cm caused by the accuracy limit of the equilibrium reconstruction are taken into account. Simulated divertor profiles obtained from the upstream transport ansatz and the experimental boundary conditions agree with measurements, except a small region localized at the separatrix strike points which is supposed to be affected by direct ion losses. The integrated analysis using EDGE2D modelling, although still limited by the marginal spatial resolution of individual diagnostics, allows the characterization of profiles in the edge/pedestal region and supplies additional information on the separatrix position. The steep density gradient zone inside the separatrix shrinks compared to the electron temperature with increasing density, indicating the effect of the neutral penetration depth becoming shorter than the region of reduced transport.

There is an increasing body of evidence that the energy lost from diverted tokamak plasmas due to edge localized mode (ELM) activity may not be confined solely to deposition on divertor components. Plasma-facing surfaces in the main confinement chamber also appear to intercept significant fluxes. Whilst this is of no practical consequence for the operation of present day facilities, concerns are being raised over the possible impact on future devices, in which ELMS carrying higher energies are expected. A key parameter required in this assessment is the energy transported by ions in the ELM as it moves through the scrape-off layer (SOL). This contribution presents the first known direct experimental demonstration that ELM events can convect ions with considerable energies to regions in the far SOL. These measurements, obtained on the JET tokamak with an ion energy analyser probe, are combined with a recently developed SOL transient model to show that the ions can, indeed, reach first limiting surfaces with energies that are a considerable fraction (similar to 50%) of those found in the H-mode edge pedestal region. This experiment-theory comparison supports a picture of the ELM in which filaments of hot plasma originating in the pedestal region dissipate energy primarily through parallel losses to the divertor targets during their radial propagation across the SOL.

The modelling studies being performed for steady state divertor operation of the ITER-FEAT design are summarized. Optimization of the divertor geometry reveals the importance of the proper target shape for a reduction of the peak power loads. A high gas conductance between the divertor legs is also essential for maintaining acceptable conditions in the outer divertor, which receives a higher power loading than the inner divertor. Impurity seeding, which would be necessary if tritium co-deposition concerns precluded the use of carbon as the plasma facing material, can ensure the required high radiation level at acceptable Zeff, and the divertor performance is not very sensitive to the choice of radiating impurity.

Washboard (WB) modes (Smeulders P et al 1999 Plasma Phys. Control. Fusion 41 1303) are a very common edge instability regularly observed in the H-mode regime in JET. They are detected as (normally several) bands of continuously fluctuating magnetic activity rotating in the direction of the electron diamagnetic drift with typical frequencies in the range of 10–90 kHz. The time evolution of the WB mode frequency is found to follow qualitatively the evolution of the electron temperature measured near the pedestal top, probably due to the strong diamagnetic drift associated with the large pedestal gradients. Evidence for their involvement in the pedestal and ELM dynamics will be presented. Increasing WB mode amplitude is correlated with an increase in the time between consecutive type-I ELMs. In situations in which a sudden increase (decrease) of WB mode activity is observed, the build-up of the pedestal temperature (and, linked to this, also of the pedestal pressure) of the electrons is seen to become slower (faster). This is a strong indication that the WB mode activity has a regulating effect on the pedestal and that it is responsible for an enhanced transport of energy across the separatrix. The occurrence of a class of type-I ELM precursor modes commonly observed in JET in discharges with low to moderate collisionality ( , roughly) (Perez C P et al EFDA-JET Preprint EFD-P(02)11) is found to be associated with a weakening of the WB modes. The underlying mechanism for this interaction has not been yet identified. In contrast to low triangularity discharges, WB activity is found to increase with gas puffing at high triangularity. This can provide an explanation for the regime recently identified on JET that has been called the mixed type-I/type-II ELM regime (Saibene G et al 2002 Plasma Phys. Control. Fusion 44 1769). A modified version of the peeling–ballooning cycle for type-I ELMs on JET that takes into account the WB mode phenomenon and is consistent with the experimental observations is proposed.

The first burning fusion plasma experiment based on the tokamak principle, international tokamak experimental reactor (ITER) is now ready for construction. Based on the continuous progress of many years of fusion research, the design relies upon a large and robust set of experimental data. The focus of present day fusion research is therefore shifting towards the issues of ITER plasma operation and machine availability. The latter is governed mainly by plasma–wall interaction issues, in particular the lifetime of plasma-facing components and long-term tritium retention. To coordinate the research activities in this area a task force for plasma–wall interaction (EU-PWI-TF) has been initiated by the European fusion research programme under EFDA. This contribution describes the experimental database in these areas and outlines the task force strategy and further research that will be needed to address the critical issues.

In this paper we analyse and discuss the thermal energy loss dynamics before and during JET disruptions that occurred between 2002 and 2004 in discharges which reached >4.5 MJ of thermal energy. We observe the slow thermal energy transients with diamagnetic loops and the fast ones with electron cyclotron emission and soft x-ray diagnostics. For most disruption types in JET, the plasma thermal energy at the time of the thermal quench is substantially less than that of the full performance plasma, typically in the range of 10–50% depending on plasma conditions and disruption type. The exceptions to this observation are disruptions in plasmas with a strong internal transport barrier (ITB) and in discharges terminating in a pure vertical displacement event, in which the plasma conserves a very high energy content up to the thermal quench. These disruption types are very sudden, leaving little scope for the combined action of soft plasma landing strategies and intrinsic performance degradation, both requiring >500 ms to be effective, to decrease the available thermal energy. The characteristic time for the loss of energy from the main plasma towards the PFCs in the thermal quench of JET disruptions is in the range 0.05–3.0 ms. The shortest timescales are typical of disruptions caused by excessive pressure peaking in ITB discharges. The available thermal energy fraction and thermal quench duration observed in JET can be processed (with due caution) into estimates for the projected PFC lifetime of the ITER target.

A number of designs for external or in-vessel coils generating Resonant Magnetic Perturbations (RMPs) for type I ELMs control in ITER are analyzed. The RMPs generated by the I-coils in the successful DIII-D ELMs control experiments are taken as a reference. The three ITER characteristic scenarios (H-mode, hybrid and steady-state) are studied. In a second part of the paper, the first results of a self-consistent non-linear MHD modelling of the plasma response to the RMPs with the JOREK code, for a DIII-D case, are presented.

The damage threshold and damage mechanisms of divertor armor materials, i.e. CFC and tungsten, were studied under the impact of ITER relevant ELM-like loads. These experiments were carried out in a Quasi-Stationary Plasma Accelerator applying repetitive pulses of 500 μs up to 100 cycles. CFC showed preferential erosion of the PAN fiber-bundles above 0.6 MJ/m2 and cracking of pitch fiber-bundles. Tungsten showed cracking already at 0.2 MJ/m2 and melting at flat surfaces above 1 MJ/m2. Cracks in tungsten were identified as primary and secondary cracks which all propagated in the vertical direction, which was considered to be less critical. At an energy density of 1.5 MJ/m2, the melt-layer completely covered the surface and bridged the castellation slots.

Tungsten in form of macrobrush structure is foreseen as one of candidate materials for the ITER divertor and the dome. Melting of tungsten and the following melt motion and melt splashing are expected to be the main mechanisms of damage which determine the lifetime of plasma facing components. New experimental investigations of droplet emission from the W melt layer for the Edge Localised Mode (ELM)-like heat loads have been carried out at the plasma gun facility quasistationary plasma accelerators (QSPA-T). In these experiments the threshold for droplet emission and the distributions of velocity on emission angles and amplitude of the ejected droplets were determined. In the paper the main physical mechanism (the Kelvin–Helmholtz instability) of the melt splashing under the heat loads being applied at QSPA-T and those anticipated after the ITER transients is analyzed. These numerical simulations demonstrated a reasonable agreement with the experimental data on the droplet sizes and droplet velocities and allowed the projections upon the W melt splashing at ITER conditions.

The chapter summarizes the physics issues of the demonstration toroidal fusion power plant (Demo) that can be addressed by ITER operation. These include burning plasma specific issues, i.e. energetic particle behaviour and plasma self-heating effects, and a broader class of power-plant scale physics issues that cannot be fully resolved in present experiments. A critical issue for Demo is whether MHD and energetic particle modes driven by fast particles will become unstable and affect plasma performance. Self-heating effects are expected to be especially important for control of steady-state plasmas with internal transport barriers (ITBs) and high bootstrap current fractions. Experimental data from ITER will improve strongly the physics basis of projections to Demo of major plasma parameters such as the energy confinement time, beta and density limits, edge pedestal temperature and density, and thermal loads on in-vessel components caused by ELMs and disruptions. ITER will also serve as a test bed for fusion technology studies, such as power plant plasma diagnostics, heating and current drive systems, plasma facing components, divertor and blanket modules. Finally, ITER is expected to provide benefits for the understanding of burning plasma behaviour in other magnetic confinement schemes.

The dependence of the H-mode edge transport barrier width on normalized ion gyroradius (ρ * = ρ/a) in discharges with type I ELMs was examined in experiments combining data for the JET and DIII-D tokamaks.The plasma configuration as well as the local normalized pressure (β), collisionality (ν * ), Mach number and the ratio of ion and electron temperature at the pedestal top were kept constant, while ρ * was varied by a factor of four.The width of the steep gradient region of the electron temperature (T e ) and density (n e ) pedestals

Tungsten in form of a macrobrush structure is foreseen as one of two candidate materials for the ITER divertor. The main mechanisms of metallic target damage are surface melting and melt motion erosion, which determines the lifetime of plasma facing components (PFC). The damage to W-macrobrush targets under repetitive ELM-like heat loads corresponding to the conditions of the plasma gun QSPA-T and ITER is numerically investigated with the three-dimensional melt motion code MEMOS. The calculations revealed a significant damage to brush edges caused by the interaction of impacting plasma with the lateral surfaces. In addition, experimentally observed overlapping of brush gaps by molten tungsten was numerically confirmed. These 3D effects of the repetitive transient loads may significantly influence the PFC lifetime.

Power requirements for maintaining sufficiently high confinement (i.e. normalized energy confinement time H 98 ⩾ 1) in H-mode and its relation to H-mode threshold power scaling, P th , are of critical importance to ITER. In order to better characterize these power requirements, recent experiments on the Alcator C-Mod tokamak have investigated H-mode properties, including the edge pedestal and global confinement, over a range of input powers near and above P th . In addition, we have examined the compatibility of impurity seeding with high performance operation, and the influence of plasma radiation and its spatial distribution on performance. Experiments were performed at 5.4 T at ITER relevant densities, utilizing bulk metal plasma facing surfaces and an ion cyclotron range of frequency waves for auxiliary heating. Input power was scanned both in stationary enhanced D α (EDA) H-modes with no large edge localized modes (ELMs) and in ELMy H-modes in order to relate the resulting pedestal and confinement to the amount of power flowing into the scrape-off layer, P net , and also to the divertor targets. In both EDA and ELMy H-mode, energy confinement is generally good, with H 98 near unity. As P net is reduced to levels approaching that in L-mode, pedestal temperature diminishes significantly and normalized confinement time drops. By seeding with low- Z impurities, such as Ne and N 2 , high total radiated power fractions are possible, along with substantial reductions in divertor heat flux (&gt;4×), all while maintaining H 98 ∼ 1. When the power radiated from the confined versus unconfined plasma is examined, pedestal and confinement properties are clearly seen to be an increasing function of P net , helping to unify the results with those from unseeded H-modes. This provides increased confidence that the power flow across the separatrix is the correct physics basis for ITER extrapolation. The experiments show that P net / P th of one or greater is likely to lead to H 98 ⩾ 1 operation, and also that such a condition can be made compatible with a low- Z radiative impurity solution for reducing divertor heat loads to levels acceptable for ITER.

Experiments at DIII-D investigated the effects of magnetic error fields similar to those expected from proposed ITER test blanket modules (TBMs) containing ferromagnetic material. Studied were effects on: plasma rotation and locking, confinement, L–H transition, the H-mode pedestal, edge localized modes (ELMs) and ELM suppression by resonant magnetic perturbations, energetic particle losses, and more. The experiments used a purpose-built three-coil mock-up of two magnetized ITER TBMs in one ITER equatorial port. The largest effect was a reduction in plasma toroidal rotation velocity v across the entire radial profile by as much as Δ v / v ∼ 60% via non-resonant braking. Changes to global Δ n / n , Δβ/β and ΔH 98 /H 98 were ∼3 times smaller. These effects are stronger at higher β. Other effects were smaller. The TBM field increased sensitivity to locking by an applied known n = 1 test field in both L- and H-mode plasmas. Locked mode tolerance was completely restored in L-mode by re-adjusting the DIII-D n = 1 error field compensation system. Numerical modelling by IPEC reproduces the rotation braking and locking semi-quantitatively, and identifies plasma amplification of a few n = 1 Fourier harmonics as the main cause of braking. IPEC predicts that TBM braking in H-mode may be reduced by n = 1 control. Although extrapolation from DIII-D to ITER is still an open issue, these experiments suggest that a TBM-like error field will produce only a few potentially troublesome problems, and that they might be made acceptably small.