Showing papers in "Nuclear Fusion in 2019"

Progress of the CFETR design

Abstract: The Chinese Fusion Engineering Testing Reactor (CFETR), complementing the ITER facility, is aiming to demonstrate fusion energy production up to 200 MW initially and to eventually reach DEMO relevant power level 1 GW, to manifest a high duty factor of 0.3–0.5, and to pursue tritium self-sufficiency with tritium breeding ratio (TBR) >1. The key challenge to meet the missions of the CFETR is to run the machine in steady state (or long pulse) and high duty factor. By using a multi-dimensional code suite with physics-based models, self-consistent steady-state and hybrid mode scenarios for CFETR have been developed under a high magnetic field up to 6.5 T. The negative-ion neutral beam injection together with high frequency electron cyclotron wave and lower hybrid wave (and/or fast wave) are proposed to be used to drive the current. Subsequently the engineering design of CFETR including the magnet system, vacuum system, tritium breeding blanket, divertor, remote handling and maintenance system will be introduced. Some research and development (R&D) activities are also introduced in this paper.

Overview of first Wendelstein 7-X high-performance operation

Abstract: The optimized superconducting stellarator device Wendelstein 7-X (with major radius $R=5.5\,\mathrm{m}$, minor radius $a=0.5\,\mathrm{m}$, and $30\,\mathrm{m}^3$ plasma volume) restarted operation after the assembly of a graphite heat shield and 10 inertially cooled island divertor modules. This paper reports on the results from the first high-performance plasma operation. Glow discharge conditioning and ECRH conditioning discharges in helium turned out to be important for density and edge radiation control. Plasma densities of $1-4.5\cdot 10^{19}\,\mathrm{m}^{-3}$ with central electron temperatures $5-10\,\mathrm{keV}$ were routinely achieved with hydrogen gas fueling, frequently terminated by a radiative collapse. Plasma densities up to $1.4\cdot 10^{20}\,\mathrm{m}^{-3}$were reached with hydrogen pellet injection and helium gas fueling. Here, the ions are indirectly heated, and at a central density of $8\cdot 10^{19}\,\mathrm{m}^{-3}$ a temperature of $3.4\,\mathrm{keV}$ with $T_e/T_i=1$ was accomplished, which corresponds to $nT_i(0)\tau_E=6.4\cdot 10^{19}\,\mathrm{keVs}/\mathrm{m}^3$ with a peak diamagnetic energy of $1.1\,\mathrm{MJ}$. The discharge behaviour has further improved with boronization of the wall. After boronization, the oxygen impurity content was reduced by a factor of 10, the carbon impurity content by a factor of 5. The reduced (edge) plasma radiation level gives routinely access to higher densities without radiation collapse, e.g. well above $1\cdot 10^{20}\,\mathrm{m}^{-2}$ line integrated density and $T_e=T_i=2\,\mathrm{keV}$ central temperatures at moderate ECRH power. Both X2 and O2 mode ECRH schemes were successfully applied. Core turbulence was measured with a phase contrast imaging diagnostic and suppression of turbulence during pellet injection was observed.

Overview of the DEMO staged design approach in Europe

Abstract: This paper describes the status of the pre-conceptual design activities in Europe to advance the technical basis of the design of a DEMOnstration Fusion Power Plant (DEMO) to come in operation around the middle of this century with the main aims of demonstrating the production of few hundred MWs of net electricity, the feasibility of operation with a closed- tritium fuel cycle, and maintenance systems capable of achieving adequate plant availability. This is expected to benefit as much as possible from the ITER experience, in terms of design, licensing, and construction. Emphasis is on an integrated design approach, based on system engineering, which provides a clear path for urgent R&D and addresses the main design integration issues by taking account critical systems interdependencies and inherent uncertainties of important design assumptions (physics and technology). A design readiness evaluation, together with a technology maturation and down selection strategy are planned through structured and transparent Gate Reviews. By embedding industry experience in the design from the beginning it will ensure that early attention is given to technology readiness and industrial feasibility, costs, maintenance, power conversion, nuclear safety and licensing aspects.

High-power gyrotrons for electron cyclotron heating and current drive

Abstract: In many tokamak and stellarator experiments around the globe that are investigating energy production via controlled thermonuclear fusion, electron cyclotron heating and current drive (ECH&CD) are used for plasma start-up, heating, non-inductive current drive and MHD stability control. ECH will be the first auxiliary heating method used on ITER. Megawatt-class, continuous wave (CW) gyrotrons are employed as high-power millimeter (mm)-wave sources. The present review reports on the worldwide state-of-the-art of high-power gyrotrons. Their successful development during the past years changed ECH from a minor to a major heating method. After a general introduction of the various functions of ECH&CD in fusion physics, especially for ITER, Section 2 will explain the fast-wave gyrotron interaction principle. Section 3 discusses innovations on the components of modern long-pulse fusion gyrotrons (magnetron injection electron gun (MIG), beam tunnel, cavity, quasi-optical output coupler, synthetic diamond output window, single-stage depressed collector) and auxiliary components (superconducting magnets, gyrotron diagnostics, high-power calorimetric dummy loads). Section 4 deals with present megawatt-class gyrotrons for ITER, W7-X, LHD, EAST, KSTAR and JT-60SA, and also includes tubes for moderate pulse length machines as, ASDEX-U, DIII-D, HL-2A, TCV, QUEST and GAMMA-10. In Section 5 the development of future advanced fusion gyrotrons is discussed. These are tubes with higher frequencies for DEMO, multi-frequency (multi-purpose) gyrotrons, stepwise frequency tunable tubes for plasma stabilization, injection-locked and coaxial-cavity multi-megawatt gyrotrons, as well as sub-THz gyrotrons for collective Thomson scattering (CTS). Efficiency enhancement via multi-stage depressed collectors, fast oscillation recovery methods and reliability, availability, maintainability and inspectability (RAMI) will be discussed at the end of this section.

Recent advances in EAST physics experiments in support of steady-state operation for ITER and CFETR

Abstract: Since the last IAEA-FEC in 2016, the EAST physics experiments have been developed further in support of high-performance steady-state operation for ITER and CFETR. First demonstration of g100 seconds time scale long-pulse steady-state scenario with a good plasma performance (H98(y2) ~ 1.1) and a good control of impurity and heat exhaust with the upper tungsten divertor has been achieved on EAST using the pure radio frequency (RF) power heating and current drive. The EAST operational domain has been significantly extended towards more ITER and CFETR related high beta steady-state regime (βP ~ 2.5 & βN ~ 1.9 of using RF & NB and βP ~ 1.9 & βN ~ 1.5 of using pure RF). A large bootstrap current fraction up to 47% has been achieved with with q95~6.0-7.0. The interaction effect between the electron cyclotron resonant heating (ECRH) and two lower hybrid wave (LHW) systems has been investigated systematically, and applied for the improvement of current drive efficiency and plasma confinement quality in the steady-state scenario development on EAST. Full ELM suppression using the n= 2 RMPs has been achieved in ITER-like standard type-I ELMy H-mode plasmas with a range of the edge safety factor of q95 ≈ 3.2-3.7 on EAST. Reduction of the peak heat flux on the divertor was demonstrated using the active radiation feedback control. An increase in the total heating power and improvement of the plasma confinement are expected using a 0-D model prediction for higher bootstrap fraction. Towards long-pulse, high bootstrap current fraction operation, a new lower ITER-like tungsten divertor with active water-cooling will be installed, together with further increase and improvement of heating and current drive capability.

Properties of a new quasi-axisymmetric configuration

Abstract: A novel, compact, quasi-axisymmetric configuration is presented which exhibits low fast-particle losses and is stable to ideal magnetohydrodynamic (MHD) instabilities. The design has fast-particle loss rates below 8% for flux surfaces within the half-radius, and is shown to have a MHD stability limit of a normalized pressure of , where is volume averaged. The flux surfaces at various plasma betas and currents as calculated using the SPEC equilibrium code (Loizu et al 2016 Phys. Plasmas 23 112505) are presented. Neoclassical transport coefficients are shown to be similar to an equivalent tokamak, with a distinct banana regime at half-radius. An initial coil design study is presented to assess the feasibility of this configuration as a fusion-relevant experiment.

Overview of the JET preparation for deuterium–tritium operation with the ITER like-wall

Abstract: For the past several years, the JET scientific programme (Pamela et al 2007 Fusion Eng. Des. 82 590) has been engaged in a multi-campaign effort, including experiments in D, H and T, leading up to 2020 and the first experiments with 50%/50% D–T mixtures since 1997 and the first ever D–T plasmas with the ITER mix of plasma-facing component materials. For this purpose, a concerted physics and technology programme was launched with a view to prepare the D–T campaign (DTE2). This paper addresses the key elements developed by the JET programme directly contributing to the D–T preparation. This intense preparation includes the review of the physics basis for the D–T operational scenarios, including the fusion power predictions through first principle and integrated modelling, and the impact of isotopes in the operation and physics of D–T plasmas (thermal and particle transport, high confinement mode (H-mode) access, Be and W erosion, fuel recovery, etc). This effort also requires improving several aspects of plasma operation for DTE2, such as real time control schemes, heat load control, disruption avoidance and a mitigation system (including the installation of a new shattered pellet injector), novel ion cyclotron resonance heating schemes (such as the threeions scheme), new diagnostics (neutron camera and spectrometer, active Alfven eigenmode antennas, neutral gauges, radiation hard imaging systems…) and the calibration of the JET neutron diagnostics at 14 MeV for accurate fusion power measurement. The active preparation of JET for the 2020 D–T campaign provides an incomparable source of information and a basis for the future D–T operation of ITER, and it is also foreseen that a large number of key physics issues will be addressed in support of burning plasmas.

Overview of the recent experimental research on the J-TEXT tokamak

Abstract: Recent J-TEXT research has highlighted the significance of the role that non-axisymmetric magnetic perturbations, so called 3D magnetic perturbation (MP) fields, play in fundamentally 2D concept, i.e. tokamak. This paper presents the J-TEXT results achieved over the last two years, especially on the impacts of 3D MP fields on MHD instabilities, plasma disruptions and plasma turbulence transport. On J-TEXT, the resonant MPs (RMPs) system, capable of providing either a static or a high frequency (up to 8 kHz) rotating RMP field, has been upgraded by adding a new set of 12 in-vessel saddle coils. The shattered pellet injection (SPI) system has been built in J-TEXT in the spring of 2018. The new capabilities advance J-TEXT to be a forefront of international magnetic fusion facilities, allow a flexible study of 3D effects and disruption mitigation in a tokamak. The fast rotating RMP field has been successfully applied for avoidance of mode locking and the prevention of plasma disruption. A new control strategy, which applies pulsed RMP to the tearing mode only during the accelerating phase region, was proved by nonlinear numerical modelling to be efficient in accelerating mode rotation and even completely suppresses the mode. Remarkably, the rotating tearing mode was completely suppressed by the electrode biasing. The impacts of 3D magnetic topology on the turbulences have been investigated on J-TEXT. It is found that the fluctuations of electron density, electron temperature, and plasma potential can be significantly modulated by the island structure, and a larger fluctuation level appears at the X-point of islands. The suppression of runaway electrons during disruptions is essential to the operation of ITER, and it has been reached by utilizing the 3D magnetic perturbations on J-TEXT. This may provide an alternative mechanism of runaway suppression for large-scale tokamak and ITER.

A real-time machine learning-based disruption predictor in DIII-D

Abstract: A disruption prediction algorithm, named DPRF (Disruption Prediction using Random Forests), has run in real-time in the DIII-D plasma control system (PCS) over more than 900 discharges. DPRF naturally provides a probability mapping associated to its predictions, i.e. the disruptivity signal, now incorporated in the DIII-D PCS. This paper discusses disruption prediction accomplishments in terms of shot-by-shot performances, by simulating alarms on each discharge as in the PCS framework. Depending on the optimised performance metric chosen to evaluate DPRF, we find that almost all disruptive discharges are detected in average with few hundred of milliseconds of warning time, but this comes at a high cost of false alarms produced. Performances are not satisfying ITER requirements, where the success rate has to be higher than 95%, but this is not completely unexpected. DPRF is trained on many years of major disruptions occurring during the flattop phase of the plasma current on DIII-D, but without any differentiation by cause. Furthermore, we find that DPRF is affected by a relatively high fraction of false alarms occurring during the first 500 milliseconds from the flattop onset. This subtle effect, more evident on the discharges where DPRF ran in real-time, can be marginalized by taking specific precautions on the validity range of the predictions, and performances do improve. Even if presently burdened by some limitations, DPRF provides an incredible and novel advantage. Thanks to the feature contribution analysis (e.g., the identification of which signals contributed to triggering an alarm), it is possible to interpret and explain DPRF predictions. It is the first time that such interpretability features is exploited by a disruption predictor: by uncovering the disruption events' causes, a better understanding of disruption dynamics is achieved, and a clear path toward the design of disruption avoidance strategies can be provided.

Machine learning for disruption warnings on Alcator C-Mod, DIII-D, and EAST

Abstract: This paper reports on disruption prediction using a shallow machine learning method known as a random forest, trained on large databases containing only plasma parameters that are available in real-time on Alcator C-Mod, DIII-D, and EAST. The database for each tokamak contains parameters sampled ~106 times throughout ~104 discharges (disruptive and non-disruptive) over the last four years of operation. It is found that a number of parameters (e.g. , , , ) exhibit changes in aggregate as a disruption is approached on one or more of these tokamaks. However, for each machine, the most useful parameters, as well as the details of their precursor behaviors, are markedly different. When the prediction problem is framed using a binary classification scheme to discriminate between time slices 'close to disruption' and 'far from disruption', it is found that the prediction algorithms differ substantially in performance among the three machines on a time slice-by-time slice basis, but have similar disruption detection rates (~80%–90%) on a shot-by-shot basis after appropriate optimisation. This could have important implications for disruption prediction and avoidance on ITER, for which development of a training database of disruptions may be infeasible. The algorithm's output is interpretable using a method that identifies the most strongly contributing input signals, which may have implications for avoiding disruptive scenarios. To further support its real-time capability, successful applications in inter-shot and real-time environments on EAST and DIII-D are also discussed.

The European approach to the fusion-like neutron source: the IFMIF-DONES project

Abstract: The need of a neutron source for the qualification of materials to be used in future fusion power reactors have been recognized in the European (EU) fusion programme for many years. The construction and exploitation of this facility is presently considered to be in the critical path of DEMO. This issue prompted the EU to launch activities for the design and engineering of the IFMIF-DONES (International Fusion Materials Irradiation Facility-DEMO Oriented Neutron Source) facility based on and taking profit of the results obtained in the IFMIF/EVEDA (Engineering Validation and Engineering Design Activities) project, presently conducted in the framework of the EU-Japan Bilateral Agreement on the Broader Approach to Fusion. These activities and R&D work for the IFMIF-DONES Plant are presently taking place in the framework of a work package of the EUROfusion Consortium, in direct collaboration with the Fusion for Energy Organization. The main objective of these activities is to consolidate the design and the underlying technology basis in order to be ready for IFMIF-DONES construction as early as possible. The paper presents the main engineering results for a generic site obtained during the first years of design work, as indicated in the recently released IFMIF-DONES Preliminary Engineering Design Report, making emphasis on the design evolution from previous phases and on the critical issues to be further developed in the near future. The proposed European site to host the facility (Granada, Spain) is briefly introduced as well.

Formation of hot, stable, long-lived field-reversed configuration plasmas on the C-2W device

Abstract: TAE Technologies' research is devoted to producing high temperature, stable, long-lived field-reversed configuration (FRC) plasmas by neutral-beam injection (NBI) and edge biasing/control. The newly constructed C-2W experimental device (also called "Norman") is the world's largest compact-toroid (CT) device, which has several key upgrades from the preceding C-2U device such as higher input power and longer pulse duration of the NBI system as well as installation of inner divertors with upgraded electrode biasing systems. Initial C-2W experiments have successfully demonstrated a robust FRC formation and its translation into the confinement vessel through the newly installed inner divertor with adequate guide magnetic field. They also produced dramatically improved initial FRC states with higher plasma temperatures (Te ~250+ eV; total electron and ion temperature g1.5 keV, based on pressure balance) and more trapped flux (up to ~15 mWb, based on rigid-rotor model) inside the FRC immediately after the merger of collided two CTs in the confinement section. As for effective edge control on FRC stabilization, a number of edge biasing schemes have been tried via open field-lines, in which concentric electrodes located in both inner and outer divertors as well as end-on plasma guns are electrically biased independently. As a result of effective outer-divertor electrode biasing alone, FRC plasma diamagnetism duration has reached up to ~9 ms which is equivalent to C-2U plasma duration. Magnetic field flaring/expansion in both inner and outer divertors plays an important role in creating a thermal insulation on open field-lines to reduce a loss rate of electrons, which leads to improvement of the edge and core FRC confinement properties. Experimental campaign with inner-divertor magnetic-field flaring has just commenced and early result indicates that electron temperature of the merged FRC stays relatively high and increases for a short period of time, presumably by NBI and ExB heating.

Integrated operation of steady-state long-pulse H-mode in Experimental Advanced Superconducting Tokamak

Abstract: Recent Experimental Advanced Superconducting Tokamak (EAST) experiments have successfully demonstrated a long-pulse steady-state scenario with improved plasma performance through integrated operation since the last IAEA FEC in 2016. A discharge with a duration over 100 s using pure radio frequency (RF) power heating and current drive has been obtained with the required characteristics for future long-pulse tokamak reactors such as good energy confinement quality (H98y2 ~ 1.1) with electron internal transport barrier inside ρ < 0.4, small ELMs (frequency ~100–200 Hz), and good control of impurity and heat exhaust with the tungsten divertor. The optimization of X-point, plasma shape, the outer gap and local gas puffing near the low hybrid wave (LHW) antenna were integrated with global parameters of B T and line-averaged electron density for higher current drive efficiency of LHW and on-axis deposition of electron cyclotron heating in the long-pulse operation. More recently, a high β P RF-only discharge (β P ~ 1.9 and β N ~ 1.5, /n GW ~ 0.80, f bs ~ 45% at q 95 ~ 6.8) was successfully maintained over 24 s with improved hardware capabilities, demonstrating performance levels needed for the China Fusion Engineering Test Reactor steady-state operation. A higher energy confinement is observed at higher β P and with favorable toroidal field direction. Towards the next goal (≥400 s long-pulse H-mode operations with ~50% bootstrap current fraction) on EAST, an integrated control of the current density profile, pressure profile and radiated divertor will be addressed in the near future.

Determination of the tolerable impurity concentrations in a fusion reactor using a consistent set of cooling factors

Abstract: In the present work, the tolerable impurity level and composition for a reactor plasma using several sets of model assumptions are evaluated. Special care was taken to evaluate a comprehensive and consistent set of atomic data for 35 different elements, such that the impurity level for various elements may be studied as a function of their nuclear charge. The data set may not only be useful for the presented work or for system codes which design fusion reactors, but also for interpretation of bolometric measurements. Additionally, the predictions of the spectral distribution of the radiated power is of high quality such that soft x-ray broadband measurements may be interpreted. In the present work the data is used for predicting the radiated power in a reactor plasma, using a 0D, several variants of a 0.5D model and a realistic 1D ASTRA modelling of a DEMO plasma, i.e. the EU DEMO1 2015 design. The maximal or appropriate impurity content of a reactor plasma for all models can be determined, such that the predictions from a simplistic 0D model can be compared to less simplistic models and a proper reactor simulation. These comparisons suggest that with the simplistic models the impurity content may be estimated within a factor of about 1.5, independent of the realization of the reactor plasma. At the same time this study underlines the sensitivity of the reactor performance on the impurity mixture and especially of the He content of the plasma. Additionally, an extended 0.5D model is presented which is able to predict variations of the fusion yield Q and the He concentration, when both is known for a reference scenario. These predictions prove to be of high accuracy when compared to the 1D ASTRA modelling and thus, allow the net impact of an increased dilution and a simlutaneous temperature rise at constant plasma pressure to be evaluated. Furthermore, the parameter space is scanned with more than 105 model reactor plasmas demonstrating that the use of a low-Z impurity diminishes the possibility of an economical feasible reactor plasma. The main results of the parameter scan are made available via scaling formulae.

Erosion, screening, and migration of tungsten in the JET divertor

Abstract: The erosion of tungsten (W), induced by the bombardment of plasma and impurity particles, determines the lifetime of plasma-facing components as well as impacting on plasma performance by the influ ...

Progress toward ITER’s First Plasma

Abstract: 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.

Progress in disruption prevention for ITER

Abstract: E.J. Strait, J.L. Barr, M. Baruzzo, J.W. Berkery, R.J. Buttery, P.C. de Vries, N.W. Eidietis, R.S. Granetz, J.M. Hanson, C.T. Holcomb, D.A. Humphreys, J.H. Kim, E. Kolemen, M. Kong, M.J. Lanctot, M. Lehnen, E. Lerche, N.C. Logan, M. Maraschek, M. Okabayashi, J.K. Park, A. Pau, G. Pautasso, F.M. Poli, C. Rea, S.A. Sabbagh, O. Sauter, E. Schuster, U.A. Sheikh, C. Sozzi, F. Turco, A.D. Turnbull, Z.R. Wang, W.P. Wehner, L. Zeng

Ion beam analysis of fusion plasma-facing materials and components : facilities and research challenges

Abstract: Following the IAEA Technical Meeting on “Advanced Methodologies for the Analysis of Materials in Energy Applications Using Ion Beam Accelerators”, this paper reviews the current status of ion beam analysis techniques and some aspects of ion-induced radiation damage in materials for the field of materials relevant to fusion. Available facilities, apparatus development and future research options and challenges are presented and discussed. The analysis of beryllium and radioactivity-containing samples from future experiments in JET or ITER represents not only an analytical but also a technical challenge. A comprehensive list of the facilities, their current status, and analytical capabilities comes alongside detailed descriptions of the labs. A discussion of future issues of sample handling and the current status of facilities at JET complete the technical section. To prepare the international ion beam analysis community for these challenges, the IAEA technical meeting concludes the necessity for determining new nuclear reaction cross-sections and improving the inter-laboratory comparability by defining international standards and testing these via a round-robin test.

Real-time capable modeling of neutral beam injection on NSTX-U using neural networks

Abstract: A new model of heating, current drive, torque and other effects of neutral beam injection on NSTX-U that uses neural networks has been developed. The model has been trained and tested on the results of the Monte Carlo code NUBEAM for the database of experimental discharges taken during the first operational campaign of NSTX-U. By projecting flux surface quantities onto empirically derived basis functions, the model is able to efficiently and accurately reproduce the behavior of both scalars, like the total neutron rate and shine through, and profiles, like beam current drive and heating. The model has been tested on the NSTX-U real-time computer, demonstrating a rapid execution time orders of magnitude faster than the Monte Carlo code that is well suited for the iterative calculations needed to interpret experimental results, optimization during scenario development activities, and real-time plasma control applications. Simulation results of a proposed design for a nonlinear observer that embeds the neural network calculations to estimate the poloidal flux profile evolution, as well as and fast ion diffusivity, are presented.

Role of the pedestal position on the pedestal performance in AUG, JET-ILW and TCV and implications for ITER

Abstract: The role of the pedestal position on the pedestal performance has been investigated in AUG, JET-ILW and TCV. When the pedestal is peeling-ballooning (PB) limited, the three machines show a similar ...

High fusion performance in Super H-mode experiments on Alcator C-Mod and DIII-D

Abstract: The 'Super H-Mode' regime is predicted to enable pedestal height and fusion performance substantially higher than standard H-Mode operation. This regime exists due to a bifurcation of the pedestal pressure, as a function of density, that is predicted by the EPED model to occur in strongly shaped plasmas above a critical pedestal density. Experiments on Alcator C-Mod and DIII-D have achieved access to the Super H-Mode (and Near Super H) regime, and obtained very high pedestal pressure, including the highest achieved on a tokamak (p ped ∼ 80 kPa) in C-Mod experiments operating near the ITER magnetic field. DIII-D Super H experiments have demonstrated strong performance, including the highest stored energy in the present configuration of DIII-D (W ∼ 2.2-3.2 MJ), while utilizing only about half of the available heating power (P heat ∼ 7-12 MW). These DIII-D experiments have obtained the highest value of peak fusion gain, Q DT,equiv ∼ 0.5, achieved on a medium scale (R < 2 m) tokamak. Sustained high performance operation (β N ∼ 2.9, H98 ∼ 1.6) has been achieved utilizing n = 3 magnetic perturbations for density and impurity control. Pedestal and global confinement has been maintained in the presence of deuterium and nitrogen gas puffing, which enables a more radiative divertor condition. A pair of simple performance metrics is developed to assess and compare regimes. Super H-Mode access is predicted for ITER and expected, based on both theoretical prediction and observed normalized performance, to allow ITER to achieve its goals (Q = 10) at I p < 15 MA, and to potentially enable more compact, cost effective pilot plant and reactor designs.

Progress of indirect drive inertial confinement fusion in the United States

Abstract: Indirect drive converts high power laser light into x-rays using small high-Z cavities called hohlraums. X-rays generated at the hohlraum walls drive a capsule filled with deuterium– tritium (DT) fuel to fusion conditions. Recent experiments have produced fusion yields exceeding 50 kJ where alpha heating provides ~3× increase in yield over PdV work. Closing the gaps toward ignition is challenging, requiring optimization of the target/implosions and the laser to extract maximum energy. The US program has a three-pronged approach to maximize target performance, each closing some portion of the gap. The first item is optimizing the hohlraum to couple more energy to the capsule while maintaining symmetry control. Novel hohlraum designs are being pursued that enable a larger capsule to be driven symmetrically to both reduce 3D effects and increase energy coupled to the capsule. The second issue being addressed is capsule stability. Seeding of instabilities by the hardware used to mount the Nuclear Fusion Progress of indirect drive inertial confinement fusion in the United States

Progress in the ITER neutral beam test facility

Abstract: The ITER Heating Neutral Beam (HNB) injectors, one of the tools necessary both to achieve burning conditions and to control plasma instabilities, are characterized by such demanding parameters as to require the construction of a Neutral Beam Test Facility (NBTF) dedicated to their development and optimization. The NBTF was realized in Padua (Italy), with the direct contribution of the Italian government, through the Consorzio RFX as the host entity, IO, the in kind contributions of three Domestic Agencies (F4E, JADA, INDA) and the technical and scientific support of various European laboratories and universities. The NBTF hosts two experiments: SPIDER and MITICA. The former is devoted to the optimization of the HNB and DNB ion sources and to the achievement of the required source performances. MITICA is the full size prototype of the ITER HNB, with an ion source identical to the one used in SPIDER. IAEA-CN-FIP/1-3Rb [Right hand page running head is the paper number in Times New Roman 8 point bold capitals, centred] This paper gives an overview of the progress of the NBTF realization with particular emphasis on issues discovered during this phase of activities and to the adopted solutions in order to minimize the impact on the schedule while maintaining the goals of the facilities. The realization of MITICA is well advanced; it is expected to enter into operation in 2023 due to the long procurement times of the in-vessel mechanical components. The power supply designed to operate at 1MV are in an advanced phase of realization; all the high voltage components have been installed and the complex insulation test phase has begun in 2018. The construction and installation of SPIDER plant systems was successfully completed with their integration into the facility. The mechanical components of the ion source were installed inside the vessel and connected to the plants. The integrated commissioning with control (CODAS), protection and safety systems ended positively and the first experimental phase began. Finally, the first results of the SPIDER experimentation and of the 1MV insulation tests on the MITICA high voltage components are presented.

Overview of physics studies on ASDEX upgrade

Abstract: The ASDEX Upgrade (AUG) programme, jointly run with the EUROfusion MST1 task force, continues to significantly enhance the physics base of ITER and DEMO. Here, the full tungsten wall is a key asset for extrapolating to future devices. The high overall heating power, flexible heating mix and comprehensive diagnostic set allows studies ranging from mimicking the scrape-off-layer and divertor conditions of ITER and DEMO at high density to fully non-inductive operation (q 95 = 5.5, ) at low density. Higher installed electron cyclotron resonance heating power 6 MW, new diagnostics and improved analysis techniques have further enhanced the capabilities of AUG. Stable high-density H-modes with MW m-1 with fully detached strike-points have been demonstrated. The ballooning instability close to the separatrix has been identified as a potential cause leading to the H-mode density limit and is also found to play an important role for the access to small edge-localized modes (ELMs). Density limit disruptions have been successfully avoided using a path-oriented approach to disruption handling and progress has been made in understanding the dissipation and avoidance of runaway electron beams. ELM suppression with resonant magnetic perturbations is now routinely achieved reaching transiently . This gives new insight into the field penetration physics, in particular with respect to plasma flows. Modelling agrees well with plasma response measurements and a helically localised ballooning structure observed prior to the ELM is evidence for the changed edge stability due to the magnetic perturbations. The impact of 3D perturbations on heat load patterns and fast-ion losses have been further elaborated. Progress has also been made in understanding the ELM cycle itself. Here, new fast measurements of and E r allow for inter ELM transport analysis confirming that E r is dominated by the diamagnetic term even for fast timescales. New analysis techniques allow detailed comparison of the ELM crash and are in good agreement with nonlinear MHD modelling. The observation of accelerated ions during the ELM crash can be seen as evidence for the reconnection during the ELM. As type-I ELMs (even mitigated) are likely not a viable operational regime in DEMO studies of 'natural' no ELM regimes have been extended. Stable I-modes up to have been characterised using -feedback. Core physics has been advanced by more detailed characterisation of the turbulence with new measurements such as the eddy tilt angle - measured for the first time - or the cross-phase angle of and fluctuations. These new data put strong constraints on gyro-kinetic turbulence modelling. In addition, carefully executed studies in different main species (H, D and He) and with different heating mixes highlight the importance of the collisional energy exchange for interpreting energy confinement. A new regime with a hollow profile now gives access to regimes mimicking aspects of burning plasma conditions and lead to nonlinear interactions of energetic particle modes despite the sub-Alfvenic beam energy. This will help to validate the fast-ion codes for predicting ITER and DEMO.

Influence of massive material injection on avalanche runaway generation during tokamak disruptions

Abstract: In high-current tokamak devices such as ITER, a runaway avalanche can cause a large amplification of a seed electron population. We show that disruption mitigation by impurity injection may significantly increase the runaway avalanche growth rate in such devices. This effect originates from the increased number of target electrons available for the avalanche process in weakly ionized plasmas, which is only partially compensated by the increased friction force on fast electrons. We derive an expression for the avalanche growth rate in partially ionized plasmas and investigate the effects of impurity injection on the avalanche multiplication factor and on the final runaway current for ITER-like parameters. For impurity densities relevant for disruption mitigation, the maximum amplification of a runaway seed can be increased by tens of orders of magnitude compared to previous predictions. This motivates careful studies to determine the required densities and impurity species to obtain tolerable current quench parameters, as well as more detailed modeling of the runaway dynamics including transport effects.

Application of Gaussian process regression to plasma turbulent transport model validation via integrated modelling

Abstract: This paper outlines an approach towards improved rigour in tokamak turbulence transport model validation within integrated modelling. Gaussian process regression (GPR) techniques were applied for p ...