MW Neutral Beam Research Articles

Numerical modelling for beam duct heat loads calculations and application to the new 1 MW neutral beam injector in the COMPASS tokamak

We introduce detailed numerical modelling of the fast neutral particles inside the duct of a new neutral beam injector (NBI) recently installed at the COMPASS tokamak (major radius , vessel midplane minor radius , toroidal field = 0.9–2.1 ). This new NBI system is able to deliver power to the plasma at nominal injection energy of . Collisions with the background neutrals inside the beam duct give birth to fast ions according to the density of the gas and tabulated cross-sections. The ion trajectories are then computed in the complete 3D magnetic field, showing the importance of the stray magnetic field and the magnitude of the field within the gap in between toroidal coils.During the experimental campaign dedicated to the new 1 MW NBI, the beam duct heating was measured by a row of thermocouples located on the top-half of the beam duct. The fast ions collisions with the duct wall cause a local temperature increase with a characteristic pattern. In COMPASS, the location of the fast ions power deposition measured experimentally is in qualitative agreement with modelling of ion losses when following them after the re-ionization process. We trace back the details of the orbits corresponding to deposition at the hot-spot inside the beam duct. Quantitative comparison between the experiment and the simulation shows that a larger than expected amount of neutral gas was inside the duct and this study will guide the design of the future NBI duct in COMPASS Upgrade.

1 MV power supplies integration issues in MITICA experiment, the ITER Heating Neutral Beam Injector prototype

Abstract MITICA, the full scale prototype of ITER Heating Neutral Beam Injector required to heat up ITER plasma with 16.5 MW injected power, is under realization at the Neutral Beam Test Facility (NBTF) in Padova (Italy) with the contributions of JApanese and EUropean Domestic Agencies (JADA and EUDA, respectively). The objective of MITICA is to produce a 16.5 MW neutral beam, obtained by accelerating negative Deuterium ions up to 1 MeV for a total ion current of 40 A and then neutralized. MITICA Power Supply (PS), installed from 2016 to 2019, includes several non-standard equipment, with ratings well beyond the present industrial standard for insulation voltage level (-1 MVdc) and dimensions: • the Acceleration Grid Power Supply (AGPS), composed of five DC Generators (DCG) rated for -200 kVdc each, connected in series to produce -1 MVdc acceleration voltage; • the Ion Source and Extraction Power Supply system (ISEPS); • the large air insulated Faraday cage (High Voltage Deck1, HVD1) hosting ISEPS and connected to the Transmission Line (TL) through an air-to-SF6 Bushing (High Voltage Bushing Assembly, HVBA); • a 100 m gas (SF6) insulated TL, connecting AGPS and ISEPS to the beam source installed inside the vacuum vessel through • the SF6-to-vacuum HV Bushing (HVB). The definition of the interfaces both between components supplied by the different DA’s and towards the buildings has been studied and finalized as far as possible during the design phase. Nevertheless, during the installation phase some issues emerged and had to be solved, minimizing modifications of the components already manufactured. The paper deals with the experience gained during the installation activities, focusing on solutions to interface the aforementioned equipment with NBTF buildings according to the stringent dimensional requirements and to the electrical insulation issues of the TL from the buildings. In particular, the solutions adopted to realize the electrical and mechanical interfaces between the TL and the HVBA are described in detail.

Beam duct for the 1 MW neutral beam injector on TCV

Abstract The 1MW power of the neutral beam injector (NBI) enters the plasma of the Tokamak a Configuration Variable (TCV) through a duct. The original beam propagation model predicted a maximal heat flux on the internal faces of the duct below 350 kW/m2, leading to an acceptable temperature rise for the 2 s nominal pulse duration. During commissioning, the NBI showed unacceptable overheating in the duct, the beam divergence being higher than expected. Several ion source grids have been tested to mitigate the beam divergence however the overheating of the duct is still problematic. Since then, the NBI operates with reduced power and duration to avoid damaging the beam duct. This paper describes the design and thermal analysis of a new beam duct capable of withstanding NBI shots at nominal regime for a range of modeled neutral beam divergences. The design drivers are the high heat flux protection and the accumulated heat dissipation. The concept features a thermal shield composed of a smooth elliptical castellated layer surrounded by an actively cooled structure.

Results of the first plasma campaign in a start configuration of GOL-NB multiple-mirror trap

The first plasma campaign was completed in a start configuration of the GOL-NB multiple-mirror trap. The device included a 4 m long high-field section, two end sections of magnetic flux expanders and a short temporary section for the commissioning of two 25 keV, 0.75 MW neutral beams. Low-temperature start plasma was created using a coaxial arc plasma gun. The main objective of the first experiments is to develop and characterize the low-temperature plasma stream that will be used later for filling the central trap of GOL-NB with the start plasma. The plasma parameters and operation regimes are reported.

New 50-keV neutral beam injector for the Globus-M2 spherical tokamak

A design of the new neutral beam injector and the process of generation of a high-energy atomic beam are described in detail. The injector is fully prepared for experiments on auxiliary heating of the Globus-M2 tokamak plasma. The docking of injector with tokamak vacuum vessel is completed. The predictions for non-inductive current drive by 50 keV 1 MW neutral beam in the Globus-M2 are presented.

Physics research on the TCV tokamak facility: from conventional to alternative scenarios and beyond

The research program of the TCV tokamak ranges from conventional to advanced-tokamak scenarios and alternative divertor configurations, to exploratory plasmas driven by theoretical insight, exploiting the device’s unique shaping capabilities. Disruption avoidance by real-time locked mode prevention or unlocking with electron-cyclotron resonance heating (ECRH) was thoroughly documented, using magnetic and radiation triggers. Runaway generation with high-Z noble-gas injection and runaway dissipation by subsequent Ne or Ar injection were studied for model validation. The new 1 MW neutral beam injector has expanded the parameter range, now encompassing ELMy H-modes in an ITER-like shape and nearly non-inductive H-mode discharges sustained by electron cyclotron and neutral beam current drive. In the H-mode, the pedestal pressure increases modestly with nitrogen seeding while fueling moves the density pedestal outwards, but the plasma stored energy is largely uncorrelated to either seeding or fueling. High fueling at high triangularity is key to accessing the attractive small edge-localized mode (type-II) regime. Turbulence is reduced in the core at negative triangularity, consistent with increased confinement and in accord with global gyrokinetic simulations. The geodesic acoustic mode, possibly coupled with avalanche events, has been linked with particle flow to the wall in diverted plasmas. Detachment, scrape-off layer transport, and turbulence were studied in L- and H-modes in both standard and alternative configurations (snowflake, super-X, and beyond). The detachment process is caused by power ‘starvation’ reducing the ionization source, with volume recombination playing only a minor role. Partial detachment in the H-mode is obtained with impurity seeding and has shown little dependence on flux expansion in standard single-null geometry. In the attached L-mode phase, increasing the outer connection length reduces the in–out heat-flow asymmetry. A doublet plasma, featuring an internal X-point, was achieved successfully, and a transport barrier was observed in the mantle just outside the internal separatrix. In the near future variable-configuration baffles and possibly divertor pumping will be introduced to investigate the effect of divertor closure on exhaust and performance, and 3.5 MW ECRH and 1 MW neutral beam injection heating will be added.

Extension of the operating space of high-bN fully non-inductive scenarios on TCV using neutral beam injection

The fully non-inductive sustainment of high normalized beta plasmas () is a crucial challenge for the steady-state operation of a tokamak reactor. In order to assess the difficulties facing such scenarios, steady-state regimes have been explored on the tokamak à configuration variable (TCV) using the newly available 1 MW neutral beam injection (NBI) system. The operating space is extended towards plasmas that are closer to those expected in JT-60SA and ITER, i.e. with significant NBI and electron cyclotron resonance heating and current drive (ECRH/CD), bootstrap current and fast ion (FI) fraction. values up to 1.4 and 1.7 are obtained in lower single null L-mode () and H-mode () plasmas, respectively, at zero time averaged loop voltage and . Fully non-inductive operation is not achieved with NBI alone, whose injection can even increase the loop voltage in the presence of EC waves. A strong contribution to the total plasma pressure of thermal and FIs from NBI is experimentally evidenced and confirmed by interpretative ASTRA and NUBEAM modeling, which further predicts that FI charge-exchange reactions are the main loss channel for NBH/CD efficiency. Internal transport barriers, which are expected to maximize the bootstrap current fraction, are not formed in either the electron or the ion channel in the plasmas explored to date, despite a significant increase in the toroidal rotation and FI fraction with NBI, which are known to reduce turbulence. First results on scenario development of high- fully non-inductive H-mode plasmas are also presented.

A measure of fast ion beta at marginal stability in the reversed field pinch

Energetic ions are well confined in the core of the reversed field pinch and a substantial population develops rapidly during tangential neutral beam injection. A saturated fast ion density is observed within several milliseconds and coincides with the onset of energetic particle mode activity. First measurements of the fusion product profile determine the fast ion density and pressure profiles, the latter of which when combined with neutral particle analysis of the energy distribution. These are key measurements in explaining the limiting behavior of the fast ion profile in response to a combination of effects. Tearing mode activity affects the fast ion profile, and the fast ions noticeably influence the core-most tearing mode. The nonlinear situation settles into a marginally stable state where the EPM transports fast ions down a steep gradient to a region where their confinement is limited by stochastic orbit and charge exchange losses. In the state at marginal stability (achieved in one set of experimental conditions with full 1 MW neutral beam power), a critical fast ion beta gradient is measured that limits the core , or nearly four times the core thermal , but less than a strictly classical calculation predicts.

Scenario development for D–T operation at JET

The JET exploitation plan foresees D–T operations in 2020 (DTE2). With respect to the first D–T campaign in 1997 (DTE1), when JET was equipped with a carbon wall, the experiments will be conducted in presence of a beryllium–tungsten ITER-like wall and will benefit from an extended and improved set of diagnostics and higher additional heating power (32 MW neutral beam injection + 8 MW ion cyclotron resonance heating). There are several challenges presented by operations with the new wall: a general deterioration of the pedestal confinement; the risk of heavy impurity accumulation in the core, which, if not controlled, can cause the radiative collapse of the discharge; the requirement to protect the divertor from excessive heat loads, which may damage it permanently. Therefore, an intense activity of scenario development has been undertaken at JET during the last three years to overcome these difficulties and prepare the plasmas needed to demonstrate stationary high fusion performance and clear alpha particle effects. The paper describes the status and main achievements of this scenario development activity, both from an operational and plasma physics point of view.

Secondary electron emission detectors for neutral beam characterization on C-2W.

Heating, current drive, and partial fueling from neutral beam injection are essential to sustainment of C-2W field-reversed configuration plasmas. C-2W has eight 2.1 MW neutral beams (16.8 MW of total electrical power), capable of providing a beam of 15 keV hydrogen neutrals for 30 ms. To maximize the effectiveness of neutral beam injection, duct losses must be minimized by maintaining beam alignment and optimizing beam current for minimum divergence. Each beam terminates on a vertical and horizontal array of secondary electron emission detectors (nine in the vertical, seven in the horizontal, and sharing one in the middle). The molybdenum detectors are spatially separated to characterize the beam size and alignment. With knowledge of the geometry of the vacuum ducts and horizontal and vertical beam profiles from test stand measurements, the focal length, divergence, and power loss were calculated. Through characterization, the set of neutral beams are optimized to inject up to 12 MW of power into the confinement vessel throughout the plasma discharge.

Simulation of the beamline thermal measurements to derive particle beam parameters in the ITER neutral beam test facility.

Injection of high energy neutral beam particles will be used in the ITER experiment for plasma heating and current drive. In a ITER heating beam injector, a 40 MW electrostatically accelerated negative beam will be neutralised and filtered along the beamline, obtaining a nominal 16.5 MW neutral beam power to be injected in the tokamak plasma or intercepted during conditioning and commissioning. The beam will heat the actively cooled panels of the beamline components with up to 13 MW/m2 surface power density and 18 MW power. These extreme conditions require testing in a ITER full scale neutral beam test facility under construction in Padova where the temperature of the beamline components will be monitored by 610 embedded thermocouples for protection against critical conditions, for recognising beam conditioning, and for deriving beam parameters. Power density maps of the expected beam-component interactions are applied on a parametric non-linear finite element model to simulate fields of expected temperatures. Such thermal maps are analyzed to derive the beam parameters during operation: divergence of 3-7 mrad and misalignment of 0-3 mrad. The sensibility of the temperature measurements is discussed considering a minimum 10% fraction of the nominal beam power.

Drift effects and up–down asymmetry in balanced double‐null DIII‐D divertor configurations

In a preliminary design study for a small‐angle slot (SAS) divertor configuration suitable for DIII‐D double‐null advanced tokamak (AT) plasmas, a conceptual “SAS 2” slot divertor configuration has been modelled with SOLPS‐ITER. A balanced double‐null equilibrium, representing an H‐mode AT plasma, is employed, with 5 MW neutral beam input power. In simulations with fully recycling walls and no pumping, an up–down‐symmetric open divertor geometry is compared with the SAS 2 configuration. In density scans without electric or magnetic drifts, the upstream separatrix electron density needed to achieve detachment at both target strike points (ne, sep, det.) is ∼35% lower in SAS 2 than in the open configuration, with ne, sep, det. = 1.66 and 2.59 × 1019 m−3, respectively. With drifts, the same ∼35% difference remains, but detachment density for each configuration is shifted up by ∼15%, giving ne, sep, det. = 1.94 and 2.99 × 1019 m−3, respectively. Cryopumping is modelled for the SAS 2 divertor, and with a pump duct deep in the upper slot on the common‐flux side, detachment density is shifted up by ∼15–20%. Strong pumping in the upper slot delays detachment there compared with the lower slot, regardless of drift direction. This research sets the stage for future modelling that would explore more sophisticated cryopumping configurations, and sensitivity of divertor performance to changes in magnetic balance and input power.

Cryopump development of the 5 MW neutral beam injector system on HL-2M tokamak.

In order to carry out the long pulse and high power neutral beam injector (NBI) heating experiment on the HL-2M tokamak and satisfy its high vacuum pumping speed in the NBI, two built-in cryopumps with large area and based on a three-stage adsorption structure have been developed. The design idea and size of the three stage structure, the manufacturing technique, and the test experiment for cryopumps have been described in this paper. The experimental result shows that the cooldown time of liquid nitrogen pipes and liquid helium pipes are 1.5 h and 10 min, respectively. With activated charcoal bonded on the condensation plates, the pumping speed reaches 1.13 × 106 l/s, and the consumption of liquid helium is 2.53 g/s and 3.11 g/s for the 4-component pump and the 5-component pump, respectively. The conductance probability of the pump inlet is 0.25 for hydrogen.

Beam duct for the 1 MW neutral beam heating injector on TCV

The Tokamak à Configuration Variable (TCV) has been recently equipped with a 1MW neutral beam heating (NBH) injector A. Fasoli for the TCV team, 2015 [1]. Two new stainless steel ports with rectangular aperture of 170×220mm have been manufactured and installed for this purpose. The NBH injector is connected to one of them via a stainless steel port extension. The port and its extension together form the beam duct between the TCV vacuum vessel (VV) and the NBH injector.A preliminary thermal analysis of the beam duct showed no expectation of thermal events such as overheating. Indeed, although the beam power density near the internal faces of the beam duct reaches ∼10MW/m2, the very grazing incidence angle between the beam axis and the wall was expected to lead to a maximal effective heat flux of ∼350kW/m2 for a maximal duration of 2s, resulting in acceptable temperature rise.As a result, the beam duct did not include any provision for cooling. However, early in 2016 the commissioning of the NBH injector showed high overheating of the port extension, resulting in local melting and ultimately vacuum leak. This paper describes the design and analysis of an actively cooled beam duct and the status of the beam duct.

The GOL-NB program: further steps in multiple-mirror confinement research

Physical and technical details of the GOL-NB project are presented. GOL-NB is a medium-scale multiple-mirror trap that is under development in the Budker Institute, Novosibirsk, Russia. This device will be created in several years as a deep conversion of the existing GOL-3 facility. It will consist of a central trap with two 0.75 MW neutral beams, two multiple-mirror solenoids, two expander tanks and a plasma gun that creates the start plasma. The central trap with the neutral beam injection-heated plasma is a compact gas-dynamic system. The multiple-mirror sections should decrease the power and particle losses along the magnetic field. The confinement improvement factor depends on plasma parameters and on the magnetic configuration in the multiple mirrors. The main physical task of GOL-NB is direct demonstration of the performance of multiple-mirror sections that will change equilibrium plasma parameters in the central trap. In this paper we discuss results of the scenario modeling and progress in the hardware.

Electron cyclotron plasma startup in the GDT experiment

We report on a new plasma startup scenario in the gas dynamic trap (GDT) magnetic mirror device. The primary 5 MW neutral beam injection (NBI) plasma heating system fires into a sufficiently dense plasma target (‘seed plasma’), which is commonly supplied by an arc plasma generator. In the reported experiments, a different approach to seed plasma generation is explored. One of the channels of the electron cyclotron resonance (ECR) heating system is used to ionize the neutral gas and build up the density of plasma to a level suitable for NBI capture. After a short transition of approximately 1 ms the discharge becomes essentially similar to a standard one initiated by the plasma gun. This paper presents the discharge scenario and experimental data on the seed plasma evolution during ECRH, along with the dependencies on incident microwave power, magnetic configuration and pressure of a neutral gas. The characteristics of the consequent high-power NBI discharge are studied and differences from the conventional scenario are discussed. A theoretical model describing the ECR breakdown and the seed plasma accumulation in a large-scale mirror trap is developed on the basis of the GDT experiment.

Conceptual design of a camera system for neutron imaging in low fusion power tokamaks

The basic principles for designing a camera system for neutron imaging in low fusion power tokamaks are illustrated for the case of the HL-2A tokamak device. HL-2A has an approximately circular cross section, with total neutron yields of about 1012 n/s under 1 MW neutral beam injection (NBI) heating. The accuracy in determining the width of the neutron emission profile and the plasma vertical position are chosen as relevant parameters for design optimization. Typical neutron emission profiles and neutron energy spectra are calculated by Monte Carlo method. A reference design is assumed, for which the direct and scattered neutron fluences are assessed and the neutron count profile of the neutron camera is obtained. Three other designs are presented for comparison. The reference design is found to have the best performance for assessing the width of peaked to broadened neutron emission profiles. It also performs well for the assessment of the vertical position.

Simulation Analysis of Charge-Exchange Losses During Neutral Beam Injection on EAST

Neutral beam injection (NBI) is one of the most effective heating methods on modern tokamak. As one of the two main losses during the NBI discharge, the charge-exchange (CX) losses are hard to measure due to their scattering. In this paper, through the simulation analysis with NUBEAM and ONETWO, CX losses have been investigated to obtain a rough evaluation by comparing to the shinethrough losses or the beam power. Firstly, according to the simulation results, the ratio of CX losses to the shinethrough losses varies from 0.28 to 1.87 in the scanning of plasma and NBI parameters. Thus, CX losses should not be ignored in the evaluations of the Low–High power threshold and beam torque in beam-heated discharges. Secondly, when Te(0) > 1 keV, CX losses have little relationship with the increasing temperature and can be roughly estimated as 17 % of the beam power or 85 % of the shinethrough losses. Finally, in the high collisional plasma, CX losses and shinethrough losses increase with the injected 80 keV neutral beam power. However, the ratio of them is almost similar. Thus, CX losses can be estimated as about 30 % of the shinethrough losses. On the other hand, scans of injected 4 MW neutral beam energy show that the beam energy has little effect on CX losses, which can be approximately estimated as 5 % of the beam power. Therefore, based on the simulation results, for EAST-NBI-1, when the line-average density \(\bar{n}_{e} = 3 \times 10^{19} /{\text{m}}^{3}\), PNBI = 2 MW and ENBI = 50 keV, the total beam losses (CX losses and shinethrough losses) are about 30 % of the beam power. These conclusions will be valuable for the experimental analysis of beam-heated discharges on EAST.

The science program of the TCV tokamak: exploring fusion reactor and power plant concepts

TCV is acquiring a new 1 MW neutral beam and 2 MW additional third-harmonic electron cyclotron resonance heating (ECRH) to expand its operational range. Its existing shaping and ECRH launching versatility was amply exploited in an eclectic 2013 campaign. A new sub-ms real-time equilibrium reconstruction code was used in ECRH control of NTMs and in a prototype shape controller. The detection of visible light from the plasma boundary was also successfully used in a position-control algorithm. A new bang-bang controller improved stability against vertical displacements. The RAPTOR real-time transport simulator was employed to control the current density profile using electron cyclotron current drive. Shot-by-shot internal inductance optimization was demonstrated by iterative learning control of the current reference trace. Systematic studies of suprathermal electrons and ions in the presence of ECRH were performed. The L–H threshold power was measured to be ∼50–75% higher in both H and He than D, to increase with the length of the outer separatrix, and to be independent of the current ramp rate. Core turbulence was found to decrease from positive to negative edge triangularity deep into the core. The geodesic acoustic mode was studied with multiple diagnostics, and its axisymmetry was confirmed by a full toroidal mapping of its magnetic component. A new theory predicting a toroidal rotation component at the plasma edge, driven by inhomogeneous transport and geodesic curvature, was tested successfully. A new high-confinement mode (IN-mode) was found with an edge barrier in density but not in temperature. The edge gradients were found to govern the scaling of confinement with current, power, density and triangularity. The dynamical interplay of confinement and magnetohydrodynamic modes leading to the density limit in TCV was documented. The heat flux profile decay lengths and heat load profile on the wall were documented in limited plasmas. In the snowflake (SF) divertor configuration the heat flux profiles were documented on all four strike points. SF simulations with the EMC3-EIRENE code, including the physics of the secondary separatrix, underestimate the flux to the secondary strike points, possibly resulting from steady-state E × B drifts. With neon injection, radiation in a SF was 15% higher than in a conventional divertor. The novel triple-null and X-divertor configurations were also achieved in TCV.

Progress in preparing scenarios for operation of the International Thermonuclear Experimental Reactor

The development of operating scenarios is one of the key issues in the research for ITER which aims to achieve a fusion gain (Q) of ∼10, while producing 500 MW of fusion power for ≥300 s. The ITER Research plan proposes a success oriented schedule starting in hydrogen and helium, to be followed by a nuclear operation phase with a rapid development towards Q ∼ 10 in deuterium/tritium. The Integrated Operation Scenarios Topical Group of the International Tokamak Physics Activity initiates joint activities among worldwide institutions and experiments to prepare ITER operation. Plasma formation studies report robust plasma breakdown in devices with metal walls over a wide range of conditions, while other experiments use an inclined EC launch angle at plasma formation to mimic the conditions in ITER. Simulations of the plasma burn-through predict that at least 4 MW of Electron Cyclotron heating (EC) assist would be required in ITER. For H-modes at q95 ∼ 3, many experiments have demonstrated operation with scaled parameters for the ITER baseline scenario at ne/nGW ∼ 0.85. Most experiments, however, obtain stable discharges at H98(y,2) ∼ 1.0 only for βN = 2.0–2.2. For the rampup in ITER, early X-point formation is recommended, allowing auxiliary heating to reduce the flux consumption. A range of plasma inductance (li(3)) can be obtained from 0.65 to 1.0, with the lowest values obtained in H-mode operation. For the rampdown, the plasma should stay diverted maintaining H-mode together with a reduction of the elongation from 1.85 to 1.4. Simulations show that the proposed rampup and rampdown schemes developed since 2007 are compatible with the present ITER design for the poloidal field coils. At 13–15 MA and densities down to ne/nGW ∼ 0.5, long pulse operation (>1000 s) in ITER is possible at Q ∼ 5, useful to provide neutron fluence for Test Blanket Module assessments. ITER scenario preparation in hydrogen and helium requires high input power (>50 MW). H-mode operation in helium may be possible at input powers above 35 MW at a toroidal field of 2.65 T, for studying H-modes and ELM mitigation. In hydrogen, H-mode operation is expected to be marginal, even at 2.65 T with 60 MW of input power. Simulation code benchmark studies using hybrid and steady state scenario parameters have proved to be a very challenging and lengthy task of testing suites of codes, consisting of tens of sophisticated modules. Nevertheless, the general basis of the modelling appears sound, with substantial consistency among codes developed by different groups. For a hybrid scenario at 12 MA, the code simulations give a range for Q = 6.5–8.3, using 30 MW neutral beam injection and 20 MW ICRH. For non-inductive operation at 7–9 MA, the simulation results show more variation. At high edge pedestal pressure (Tped ∼ 7 keV), the codes predict Q = 3.3–3.8 using 33 MW NB, 20 MW EC, and 20 MW ion cyclotron to demonstrate the feasibility of steady-state operation with the day-1 heating systems in ITER. Simulations using a lower edge pedestal temperature (∼3 keV) but improved core confinement obtain Q = 5–6.5, when ECCD is concentrated at mid-radius and ∼20 MW off-axis current drive (ECCD or LHCD) is added. Several issues remain to be studied, including plasmas with dominant electron heating, mitigation of transient heat loads integrated in scenario demonstrations and (burn) control simulations in ITER scenarios.