ITER Research Plan Research Articles

Multi-staged ERO2.0 simulation of material erosion and deposition in recessed mirror assemblies in JET and ITER

The deposition/erosion on optical diagnostic components—mirrors—is a critical issue in reactor class devices with long-pulsed high fluence plasma operation. The paper presents results of the three-dimensional Monte–Carlo code ERO2.0 for two diagnostic aperture and first mirror geometries to be deployed in ITER, along with a separate simulation study that aims to replicate results from an experimental first-mirror study carried out on JET. Promisingly, very little plasma and impurity deposition on mirrors for the anticipated plasma durations is found in the ERO2.0 modelling taking into account the current ITER Research Plan and a material mix with beryllium first wall and a tungsten divertor. The post-mortem analysis of mirrors exposed during the experiment and the initial benchmarking efforts on the JET mirror experiment are also broadly consistent, increasing the confidence in predictions for ITER.

Power and isotope effects in the ITER baseline scenario with tungsten and tungsten-equivalent radiators in DIII-D

Experiments in DIII-D document the ITER Baseline Scenario (IBS) at q 95 ∼ 3 and P IN/P LH ∼ 1–2, in both deuterium and hydrogen utilizing Kr and Xe as Tungsten-equivalent radiators. The power threshold for H-mode operation (P LH) was determined experimentally without added impurities and found to be about a factor of two higher than the scaling law. In recent IBS experiments in deuterium, intrinsic levels of metals such as Tungsten (W) or molybdenum and inconel are present that reduce the pedestal pressure by 20%–25%. A complete radiative collapse of deuterium IBS plasmas occurs at W core concentrations C W = 10−5. Simulations show that for core temperatures expected for ITER, the plasmas would not have a radiative collapse at C W = 1 × 10−5, moreover Q = 8–10 would still be achieved for C W up to 3 × 10−5. In contrast to deuterium, the IBS in hydrogen is not affected by intrinsic high-Z impurities, indicating that hydrogen H-modes in ITER may not inform the D-T phase with respect to W accumulation and discharge survival. Compared to deuterium, the pedestal pressure in hydrogen is ∼25% lower, with much higher ELM frequency of 150 Hz, decreasing with input power. Krypton was injected in a matrix scan of input power and impurity flow in IBS hydrogen discharges. Krypton impurity density profiles in hydrogen are similar to deuterium plasmas, but at Kr flows that are 2–3 times higher for the same input power. Krypton is transported into the core and affects the whole radius; at the highest injection rates a radiative collapse occurs at core radiation fractions of 0.3–0.35, consistent with the expected maximum W radiation fraction for ITER core plasmas. Comparing the results with previous International Tokamak Physics Activity database studies of the IBS confirms that at higher radiation fraction due to high-Z impurities, a drop in H 98 of >10% is observed. On the other hand, the results using Kr as a W-equivalent radiator indicate that metal (W) devices at lower core temperatures than ITER may provide overly pessimistic performance extrapolations to ITER for deuterium-tritium operation. The new DIII-D results support a more attractive option for the ITER Research Plan with a short hydrogen phase for system commissioning, transitioning to deuterium operations as soon as possible to provide relevant conditions for deuterium-tritium operations.

Changeover between helium and hydrogen fueled plasmas in JET and WEST

The pre-fusion power operation (PFPO) phase of ITER, as described in the ITER research plan with Staged Approach,22ITER Organization, “ITER Research Plan within the Staged Approach,” ITR-18–003, Aug. 2018. includes both hydrogen (H) and helium (He) plasma operations. In preparation for PFPO, both WEST and JET ran He plasma campaigns to study plasma-wall interactions in a tungsten environment. The campaigns included a back-and-forth transition between H or deuterium (D) and He plasma operation allowing the assessment of the achievable plasma content as well as the accessible wall reservoirs for respective species. The WEST changeovers included tokamak pulses with a fixed divertor configuration. The JET changeovers applied ion cyclotron wall conditioning (ICWC) and tokamak pulses including limiter phases and four different divertor configurations. Glow discharge conditioning (GDC) was applied to complete the changeovers. The results are characterized by subdivertor optical and mass spectrometric gas analyzers and spatially resolved optical emission spectroscopy. A He content of 96–97 % after H operations is achieved by tens of ICWC pulses (JET) and several dedicated diverted plasmas (WEST and JET), while a fivefold is estimated to be required for the back transition. Effective pumping of the wall released species is a key parameter for a fast changeover. Upon applying higher heating power, the relative content of the fueled plasma species decreases. The JET gas balance analysis indicates that He operation may increase H retention. WEST divertor spectroscopy indicates a larger He inventory near the inner divertor strike line. He GDC has a clear effect on the He recycling light at the WEST divertor while D GDC did not reduce the long lasting He content observed in D pulses after the JET He campaign.

PFPO plasma scenarios for exploration of long pulse operation in ITER

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.

New modelling capabilities to support the ITER EC H&CD System optimisation and preparation of plasma operation

The ITER Electron Cyclotron Resonance Heating (ECRH) & Current Drive (ECCD) system is planned to be progressively installed and commissioned following the four stage approach of the ITER Research Plan. Starting with an injection of up to 5.8 MW from one Upper Launcher (UL) in the vacuum vessel to assist the plasma breakdown during First Plasma (FP) operation, the system will then be extended to achieve a capability of 20 MW injected power in Pre-Fusion Plasma Operation (PFPO) and Fusion Power Operation (FPO) phases. Development of optical modelling was required to characterize the optical performance of the FP configuration with the so-called First Plasma Protection Components. An optical 3D model using Zemax OpticStudio® has been developed and extended to the UL. Effects of higher order modes, thermal deformations and tolerances on the UL functionality have been characterized and are presented. Finally, in preparation of plasma operation and in the frame of the EC system upgrade layout optimisation, ECRH-ECCD modelling is being undertaken within the ITER Integrated Modelling and Analysis (IMAS) suite.

Prediction of divertor heat flux width for ITER pre-fusion power operation using BOUT++ transport code

Prediction of divertor heat flux width is performed for the first and the second pre-fusion power operation (PFPO) phases specified in the new ITER research plan using BOUT++ transport code (Li et al 2018 Comput. Phys. Commun. 228 69–82). The initial plasma profiles inside the separatrix are taken from CORSICA scenario studies. Transport coefficients in transport code are calculated by inverting the plasma profiles inside the separatrix and are assumed to be constants in the scrape-off-layer. An anomalous thermal diffusivity scan is performed with E × B and magnetic drifts. The results in two scenarios identifying two distinct regimes: a drift-dominant regime when diffusivity is smaller than the respective critical thermal diffusivity χ c and a turbulence-dominant regime when diffusivity is larger than it. The Goldston heuristic drift model and the ITPA multi-machine experimental scaling yield a lower limit of the width λ q . From transport simulations, we obtain the critical thermal diffusivity χ c = 0.5 m2 s−1 for the PFPO-1 scenario with toroidal magnetic field B = 1.77 T and plasma current I p = 5 MA, and χ c = 0.3 m2 s−1 for the PFPO-2 scenario with toroidal magnetic field B = 2.65 T and plasma current I p = 7.5 MA. Separatrix temperature and collisionality also have a significant impact on the heat flux width in the drift-dominant regime. The investigation clearly yields a scaling for critical thermal diffusivity using ITER scenarios with fixed safety factor q 95, major radius R, aspect ratio R/a, and the separatrix temperature T sep, and establishes the connection with CFETR and C-Mod discharges. This scaling implies that for a given tokamak device with q 95, R, R/a, and T sep fixed, a reduction of poloidal magnetic field by a factor of 3 leads to a 9 times higher critical value of thermal diffusivity χ c, possibly yielding a transition from turbulence- to drift-dominant regime.

Preparation for assembly and commissioning of ITER

Progress is described on the construction of infrastructure and plant systems, and on the manufacturing of the ITER tokamak components and their assembly since the IAEA FEC 2018. Major achievements in this period have been the completion of the tokamak building and installation of the cryostat base and lower section that mark the start of tokamak assembly in 2020. This has been followed by the arrival of key core tokamak components to enable assembly of the first tokamak sector, which is made up of one vacuum vessel (VV) sector, 2 toroidal field (TF) coils and the corresponding thermal shield (TS), and the completion of manufacturing and testing of the first two poloidal field (PF) coils (PF5 and PF6). In parallel with this, the tools required for assembly have been commissioned and plant systems required for operations have progressed in construction and have started or will start commissioning in the near future. The remaining core tokamak components (e.g. the remaining eight VV sectors, TF coils, PF coils and central solenoid (CS) modules) are in advanced manufacturing state or already completed and under testing, as required for the assembly plan to first plasma (FP). In parallel with machine assembly, systems to support plasma operation during the FP and later phases of the ITER research plan (IRP) have made significant progress as well. This includes the plasma control system, for which the final design for FP has been completed, the electron cyclotron heating (ECH) system, diagnostic systems, plasma protection components for FP and the disruption mitigation system (DMS). The IRP itself has also been consolidated and specific issues identified where R&D support by the ITER Members’ fusion communities may refine it further. The consolidation of the IRP has involved changes to the upgrade paths considered in the baseline by the identification of MHD stable plasma scenarios that can meet ITER’s Q = 5 steady-state fusion power goal with neutral beam heating (NBI) and ECH thus removing the need for the installation of a lower hybrid current drive (LHCD) system previously included as an upgrade in the baseline for this specific purpose.

Energy deposition and melt deformation on the ITER first wall due to disruptions and vertical displacement events

An analysis workflow has been developed to assess energy deposition and material damage for ITER vertical displacement events (VDEs) and major disruptions (MD). This paper describes the use of this workflow to assess the melt damage to be expected during unmitigated current quench (CQ) phases of VDEs and MDs at different points in the ITER research plan. The plasma scenarios are modeled using the DINA code with variations in plasma current I p, disruption direction (upwards or downwards), Be impurity density n Be, and diffusion coefficient χ. Magnetic field line tracing using SMITER calculates time-dependent, 3D maps of surface power density q ⊥ on the Be-armored first wall panels (FWPs) throughout the CQ. MEMOS-U determines the temperature response, macroscopic melt motion, and final surface topology of each FWP. Effects of Be vapor shielding are included. Scenarios at the baseline combination of I p and toroidal field (15 MA/5.3 T) show the most extreme melt damage, with the assumed n Be having a strong impact on the disruption duration, peak q ⊥ and total energy deposition to the first wall. The worst-cases are upward 15 MA VDEs and MDs at lower values of n Be, with q ⊥,max = 307 MW m−2 and maximum erosion losses of ∼2 mm after timespans of ∼400–500 ms. All scenarios at 5 MA avoided melt damage, and only one 7.5 MA scenario yields a notable erosion depth of 0.25 mm. These results imply that disruptions during 5 MA, and some 7.5 MA, operating scenarios will be acceptable during the pre-fusion power operation phases of ITER. Preliminary analysis shows that localized melt damage for the worst-case disruption should have a limited impact on subsequent stationary power handling capability.

Simulation of heating and current drive sources for scenarios of the ITER research plan

Predicting the impact of heating and current drive (H&CD) sources is essential to evaluate the performance of ITER plasmas and to subsequently optimise the scenarios for the four stages of the ITER research plan. This should be done in the context of global transport calculations of complete plasma discharges. For this purpose, a dedicated workflow has been developed in the ITER integrated modelling and analysis suite as a modular component to be used together with transport solvers to quantify the dynamics of H&CD sources for the different phases of a plasma discharge, including possible synergetic effects between the heating sources. This paper presents the results of the combined modelling of H&CD sources for the ITER DT baseline 15 MA/5.3 T scenario including the synergy between neutral beam injection (NBI) of deuterium, fusion-born alpha particles and ion cyclotron resonance heating (ICRH) at the fundamental frequency of deuterium, showing modest synergetic effects. The results of the combined H&CD sources for an ITER 7.5 MA/2.65 T helium plasma of the second pre-fusion power operation phase (PFPO-2) are also shown, exhibiting more significant synergetic effects between the fundamental ICRH minority hydrogen heating and NBI hydrogen beams. Finally, a study of electron cyclotron heating absorption for an ITER helium PFPO scenario at 7.5 MA/2.65 T is also presented with a discussion on the edge parasitic absorption that arises under specific conditions.

Implications of parasitic absorption of electron cyclotron waves on ITER operation around half-field

The ITER Research Plan envision operation around half of the nominal magnetic field (i.e. around B = 2.65 T) as a path to baseline operation. This work discusses constraints on the optimal range of magnetic field, which is bounded in the lower limit by the presence of the third-harmonic electron cyclotron resonance at half field, and on the upper limit by the loss of core heating and current drive. It will be shown that increasing the magnetic field by only 3%, i.e. to 2.75 T, eliminates the third harmonic parasitic absorption without compromising demonstration of access to H-mode, while operating at a magnetic field of 3.0 T—previously proposed for optimal use of the ion cyclotron system—would impair the use of the electron cyclotron system for core-heating and current drive. Operation at 2.65 T would still be possible if the polarization of the equatorial launcher is changed from X-mode to O-mode in the current flattop phase.

First demonstration of full ELM suppression in low input torque plasmas to support ITER research plan using n = 4 RMP in EAST

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 T NBI ∼ 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 q 95 is around 3.6 and normalized plasma beta β N ∼ 1.5–1.8, similar to that in the ITER Q = 10 scenario. Suppression windows in both q 95 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%n GW (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%n GW 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 q 95 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 q 95 ∼ 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.

Integrated ELM and divertor power flux control using RMPs with low input torque in EAST in support of the ITER research plan

Experiments have been carried out at the EAST tokamak to study ITER-relevant scenario integration issues, related to edge localized mode (ELM) control in H-mode plasmas by the application of three-dimensional (3D) resonant magnetic perturbations (RMPs), which have a large impact on the execution of the ITER research plan. The EAST experiments have successfully demonstrated ELM suppression at normalized torque inputs similar to ITER. The application of RMP fields with high toroidal mode number (n = 4) reduces the impact of ELM control on energy and particle confinement compared to those use lower n (n = 1, 2) RMPs. Injection of successive pellets is found to be effective in increasing the plasma density in ELM-suppressed H-modes and reducing the divertor power without triggering large ELMs at EAST. Access to high recycling and radiative divertor conditions while maintaining ELM suppression has been demonstrated in EAST by the use of gas fuelling and neon impurity seeding. Both approaches have been found to be effective in reducing power fluxes to the divertor strike points in near-separatrix lobes for both n = 2 and n = 4 RMPs. However, reduction of power fluxes in off-separatrix lobes is only effective for n = 4 RMP application, which is consistent with magnetic topology modelling (including plasma response) results showing a shallow penetration into the confined plasma region of field lines connected to these lobes compared to n = 2. The EAST results support the use of high n 3D fields for ELM suppression in ITER high Q DT scenarios since they provide optimum integration features regarding energy and particle confinement, pellet fuelling, radiative divertor operation while eliminating ELM transient power loads and being compatible with low torque input.

H-mode plasmas in the pre-fusion power operation 1 phase of the ITER research plan

The optimum conditions for access to and sustainment of H-mode plasmas and their expected plasma parameters in the pre-fusion power operation 1 (PFPO-1) phase of the ITER research plan, where the additional plasma heating will be provided by 20 MW of electron cyclotron heating, are assessed in order to identify key open R&D issues. The assessment is performed on the basis of empirical and physics-based scalings derived from present experiments and integrated modelling of these plasmas including a range of first-principle transport models for the core plasma. The predictions of the integrated modelling of ITER H-mode plasmas are compared with ITER-relevant experiments carried out at JET (low-collisionality high-current H modes) and ASDEX Upgrade (significant electron heating) for both global H-mode properties and scale lengths of density and temperature profiles finding reasonable agreement. Specific integration issues of the PFPO-1 H-mode plasma scenarios are discussed taking into account the impact of the specificities of the ITER tokamak design (level of ripple, etc).

Impact of suprathermal ions on neutron yield in the pre-DT phase of ITER operation

An assessment of neutron production during the pre-fusion power-operation (PFPO) phase has been carried out for a representative set of plasma scenarios predicted by the ITER Research Plan. A range of heating systems, namely neutral beam injection (NBI) (hydrogen), electron cyclotron resonance heating (ECRH), and ion cyclotron resonance heating (ICRH) are planned to be used for PFPO studies in helium, hydrogen, and mixed hydrogen–helium plasmas. Fast ions (protons and 3He) originating from NBI and ICRH systems can increase neutron production in PFPO plasmas by directly interacting with intrinsic Be impurities or through secondary processes, as also evidenced at JET. The generation of fast ions in ITER PFPO scenarios has been modelled using the ASTRA-NBI and TORIC-SSFPQL codes. A significant impact of the synergy between hydrogen NBI and hydrogen-minority ICRH on neutron production in helium plasmas is reported. In addition, the stability of the toroidicity-induced Alfvén eigenmodes (TAE) is analyzed for PFPO plasmas with a high pressure of suprathermal ions and a weak reversed shear. The possible impact of sawtooth oscillations and TAEs on neutron production is discussed, based on a linear stability analysis.

Kinetic-fluid coupling simulations of ITER Type I ELM

Edge localized modes (ELMs)-induced transient heat loads on the divertor targets represent an important threat to target lifetime and can lead to the need to replace them with a frequency that has a major impact in the execution of the ITER Research Plan. Predicting the impact of such large transient heat loads through modeling is especially challenging and is often attempted through the use of fluid plasma boundary modeling codes, such as SOLPS-ITER, in which the ELM is crudely approximated as a fixed large, but limited in time, increase in anomalous cross-field transport coefficients for particles and heat to mimic a specified total ELM energy loss. However, one problem with this approach is that the boundary conditions at the target sheath entrance are expected to vary strongly in time through the ELM transient, whilst fixed kinetic target sheath heat transmission factors (SHTF), and more generally, constant heat flux limiters, are typically applied in the fluid codes.This contribution describes the first results of efforts to address ELMs issues for ITER simulations under high performance conditions using the 1D3V electrostatic parallel Particle-in-Cell (PIC) code BIT1, to study the kinetic effects and to provide time-dependent kinetic target sheath heat transmission factors. In a second step of the work, these are used in the formulation of fluid boundary conditions for calculations of ELM target heat loads using the SOLPS-ITER code.The BIT1-SOLPS-ITER coupling allows us to investigate the kinetic effects on the targets, by comparing power and particle fluxes from time-dependent simulations of ITER Type I ELMs.

ITER assembly phase: Progress toward first plasma

By the end of 2019, ITER had completed about 67% of the overall work required to achieve First Plasma. Significant progress has been achieved since SOFT 2018, most visible in the completion of key buildings and infrastructure on the worksite. The tokamak building will be ready for equipment as of Spring 2020. Commissioning of the connection to the EU grid and the substation for the steady-state electric network is complete. Planning for commissioning of other key plant systems (e.g., secondary cooling water, cryoplant, pulsed power for magnet systems) is well underway. On the manufacturing front, progress is equally impressive. The base and lower cylinder of the cryostat were completed and turned over to ITER in July 2019. The first two poloidal field coils will be turned over to ITER in early 2020, and two more are being wound. Factory tests of the first central solenoid module will be completed in 2020, and five more modules are in the manufacturing process. The first toroidal field coil is to arrive onsite in early 2020, and progress is rapid on additional TF coils. The first sections of the thermal shield are already onsite. The first vacuum vessel sector is complete and will arrive onsite in early 2020. This progress means that, for most of the major first-of-a-kind components, the capability has now been demonstrated to fabricate according to ITER’s demanding specifications. With this progress achieved, ITER will enter Assembly Phase in 2020, remaining on schedule for First Plasma in 2025. The sequence of ITER operation from First Plasma (FP) to the achievement of the Q = 10 project goal has been consolidated in a Staged Approach, with all systems to be installed before the start of full fusion power operation in 2035. The ITER Research Plan, issued publicly in 2018, is consistent with the systems available in each phase. An adjustment of the machine baseline configuration was made in 2019, with two equatorial ports allocated to the Test Blanket Systems and the development of a better performing Disruption Mitigation System. Highlights from each of these areas (manufacturing, commissioning, tokamak assembly) will be presented along with the updated status and plans.

Modelling one-third field operation in the ITER pre-fusion power operation phase

In the four-stage approach of the new ITER Research Plan, the first pre-fusion power operation (PFPO) phase will only have limited power available from external heating and current drive (H&CD) systems: 20–30 MW provided by the electron cyclotron resonance heating (ECRH) system. Accessing the H-mode confinement regime at such low auxiliary power requires operating at low magnetic field, plasma current and density, i.e. 1.8 T and 5 MA for a density between 40% and 50% of the Greenwald density. H-mode plasmas at 5 MA/1.8 T will also be investigated in the second PFPO phase when ITER will have its full complement of H&CD capabilities installed, i.e. 20–30 MW of ECRH, 20 MW of ion cyclotron resonance heating and 33 MW of neutral beam injection. This paper describes the operational constraints and the H&CD capabilities for such scenarios in hydrogen and helium plasmas, to assess their viability and the issues it will be possible to address with them. The modelling results show that 5 MA/1.8 T scenarios are viable and will allow the exploration of the H-mode physics and control issues foreseen in the ITER Research Programme in the PFPO phases.

Progress toward ITER’s First Plasma

ITER reached 50% completion of the work required to achieve First Plasma in November 2017. Progress has been made on ITER infrastructure since the 2016 Fusion Energy Conference, most visibly the construction of many key buildings. The key parts of the tokamak assembly building and the tokamak bioshield have been completed. The tokamak building itself will be ready for equipment in 2020. The cryogenic plant and the magnet power supply buildings are complete, and these systems begin commissioning in 2020. The power conversion and distribution area is complete and in operation, and construction has started on the component cooling water system building. Manufacturing of the basic components of the ITER tokamak is also proceeding well. The base and lower cylinder of the cryostat have been assembled on the ITER site. The first modules of the central solenoid and of the six poloidal field coils have been wound. The first winding packs of the toroidal field magnets are complete, as are the first casings, which have been verified to meet the high tolerances required. The first vacuum vessel sector is near completion and demonstrated to meet strict tolerances. The heating and current drive systems (neutral beams, electron cyclotron heating and ion cyclotron heating) are in the final design phase. The sequence of ITER operation from First Plasma to the achievement of the Q = 10 and Q = 5 project goals has been adapted to the Staged Approach to construction, a stepwise installation of all systems. The ITER Research Plan has been revised in 2017 to be consistent with the systems available in each phase. Physics studies focus on the disruption mitigation system, design of the ITER tungsten divertor, and modelling of ITER plasma scenarios. Modelling concentrates on the initial phases of the Research Plan and on the Q = 10 scenario, especially plasma termination.

ITER construction and manufacturing progress toward first plasma

In November 2017, the ITER Project completed 50% of the work required for First Plasma. Key buildings, such as the tokamak assembly building, the cryogenic plant, and the magnet power supply building, have been completed. The tokamak building will be ready for equipment in 2020. Thus, the physical plant is moving toward completion, and systems are entering the commissioning phase. Manufacturing progress is equally impressive. The base and lower cylinder of the cryostat have been assembled on the ITER site. Three of the six poloidal field coils and the first of the six modules of the central solenoid are being wound. The winding packs and casings for the first toroidal field magnets are complete, and the parts for the first vacuum vessel sector have been fabricated and are on progress to be assembled. The Heating and Current Drive systems are in the final design phase. The ITER operational timeline from First Plasma at the end of 2025 to achievement of the project goals has been revised in a stepwise installation of components and auxiliary systems followed by commissioning and operations; all systems will be installed before the start of the fusion power operational phase at the end of 2035. The ITER Research Plan has been made consistent in 2017 with this Staged Approach.

Overview of the JET results in support to ITER

The 2014–2016 JET results are reviewed in the light of their significance for optimising the ITER research plan for the active and non-active operation. More than 60 h of plasma operation with ITER first wall materials successfully took place since its installation in 2011. New multi-machine scaling of the type I-ELM divertor energy flux density to ITER is supported by first principle modelling. ITER relevant disruption experiments and first principle modelling are reported with a set of three disruption mitigation valves mimicking the ITER setup. Insights of the L–H power threshold in Deuterium and Hydrogen are given, stressing the importance of the magnetic configurations and the recent measurements of fine-scale structures in the edge radial electric. Dimensionless scans of the core and pedestal confinement provide new information to elucidate the importance of the first wall material on the fusion performance. H-mode plasmas at ITER triangularity (H = 1 at βN ~ 1.8 and n/nGW ~ 0.6) have been sustained at 2 MA during 5 s. The ITER neutronics codes have been validated on high performance experiments. Prospects for the coming D–T campaign and 14 MeV neutron calibration strategy are reviewed.