Next-step Tokamak Research Articles

The MEMENTO code for modeling of macroscopic melt motion in fusion devices

The MEMENTO (MEtallic Melt Evolution in Next-step TOkamaks) code is a new numerical implementation of the physics model originally developed for the MEMOS-U code with the objective to self-consistently describe the generation of melt and its subsequent large scale dynamics in fusion devices and to assess the damage of metallic reactor armor under powerful normal and off-normal plasma events. The model has been validated in multiple dedicated EUROfusion experiments. MEMENTO solves the heat and phase transfer problem coupled with the incompressible Navier–Stokes equations in the shallow water approximation for the thin liquid film over the solid metal and with the current propagation equations on a domain that features a time-evolving deforming metal-plasma interface. The code utilizes non-uniform and adaptive meshing along with sub-cycling in time facilitated by the AMReX open-source framework as well as AMReX’s built-in parallelization capabilities.

Plasma performance and operational space with an RMP-ELM suppressed edge

The operational space and global performance of plasmas with edge-localized modes (ELMs) suppressed by resonant magnetic perturbations (RMPs) are surveyed by comparing AUG, DIII-D, EAST, and KSTAR stationary operating points. RMP-ELM suppression is achieved over a range of plasma currents, toroidal fields, and RMP toroidal mode numbers. Consistent operational windows in edge safety factor are found across devices, while windows in plasma shaping parameters are distinct. Accessed pedestal parameters reveal a quantitatively similar pedestal-top density limit for RMP-ELM suppression in all devices of just over 3×1019 m−3. This is surprising given the wide variance of many engineering parameters and edge collisionalities, and poses a challenge to extrapolation of the regime. Wide ranges in input power, confinement time, and stored energy are observed, with the achieved triple product found to scale like the product of current, field, and radius. Observed energy confinement scaling with engineering parameters for RMP-ELM suppressed plasmas are presented and compared with expectations from established H and L-mode scalings, including treatment of uncertainty analysis. Different scaling exponents for individual engineering parameters are found as compared to the established scalings. However, extrapolation to next-step tokamaks ITER and SPARC find overall consistency within uncertainties with the established scalings, finding no obvious performance penalty when extrapolating from the assembled multi-device RMP-ELM suppressed database. Overall this work identifies common physics for RMP-ELM suppression and highlights the need to pursue this no-ELM regime at higher magnetic field and different plasma physical size.

Melt dynamics with MEMENTO — Code development and numerical benchmarks

The new numerical implementation of the MEMOS-U physics model, fully validated in multiple EUROfusion sponsored experiments, is presented. The computational tool - MEMENTO (MEtallic Melt Evolution in Next-step TOkamaks)- is able to address fusion-relevant melting scenarios that feature complex plasma-facing component geometries, involve intricate plasma wetting patterns and are characterized by vast spatio-temporal scale separations. The high level architecture of the code is discussed and numerical benchmarks of the heat transfer, fluid dynamics and current propagation solvers are presented.

Divertor power load investigations with deuterium and tritium in type-I ELMy H-mode plasmas in JET with the ITER-like wall

Divertor power load is a major challenge towards a burning plasma in a next-step tokamak. Here, the first results of a divertor power load characterisation in tritium plasmas in type-I ELMy H-mode, obtained in the JET deuterium-tritium campaign (DTE2) performed in 2021, are presented. It is demonstrated that both, transient loads due to type-I ELMs as well as the power fall-off length, do not exhibit an explicit ion mass dependence, with remarkably similar values in the tritium plasmas and in the deuterium references. This gives an improved credence to published scaling law predictions, solely based on deuterium plasma experiments. Moreover, the type-I ELM impact on the inner divertor target is studied in deuterium discharges. A slightly increased parallel energy fluence on the inner target with a factor of 1.08 compared to the outer target is observed. This is explained by the smaller major radius of the inner target.

Disruption event characterization and forecasting in tokamaks

Disruption prediction and avoidance is a critical need for next-step tokamaks, such as ITER. Disruption Event Characterization and Forecasting (DECAF) research fully automates analysis of tokamak data to determine chains of events that lead to disruptions and to forecast their evolution allowing sufficient time for mitigation or complete avoidance of the disruption. Disruption event chains related to local rotating or global magnetohydrodynamic (MHD) modes and vertical instability are examined with warnings issued for many off-normal physics events, including density limits, plasma dynamics, confinement transitions, and profile variations. Along with Greenwald density limit evaluation, a local radiative island power balance theory is evaluated and compared to the observation of island growth. Automated decomposition and analysis of rotating tearing modes produce physical event chains leading to disruptions. A total MHD state warning model comprised of 15 separate criteria produces a disruption forecast about 180 ms before a standard locked mode detector warning. Single DECAF event analyses have begun on KSTAR, MAST, and NSTX/-U databases with thousands of shot seconds of device operation using from 0.5 to 1 × 106 tested sample times per device. An initial multi-device database comparison illustrates a highly important result that plasma disruptivity does not need to increase as βN increases. Global MHD instabilities, such as resistive wall modes (RWMs), can give the briefest time period of warning before disruption compared to other physics events. In an NSTX database with unstable RWMs, the mode onset, loss of boundary and current control, and disruption event warnings are found in all cases and vertical displacement events are found in 91% of cases. An initial time-dependent reduced physics model of kinetic RWM stabilization created to forecast the disruption chain predicts instability 84% of the time for experimentally unstable cases with a relatively low false positive rate. Instances of the disruption event chain analysis illustrate dynamics including H–L back transitions for rotating MHD and global RWM triggering events. Disruption warnings are issued with sufficient time before the disruption (on transport timescales) to potentially allow active profile control for disruption avoidance, active mode control, or mitigation.

NSTX-U theory, modeling and analysis results

The mission of the low aspect ratio spherical tokamak NSTX-U is to advance the physics basis and technical solutions required for optimizing the configuration of next-step steady-state tokamak fusion devices. NSTX-U will ultimately operate at up to 2 MA of plasma current and 1 T toroidal field on axis for 5 s, and has available up to 15 MW of neutral beam injection power at different tangency radii and 6 MW of high harmonic fast wave heating. With these capabilities NSTX-U will develop the physics understanding and control tools to ramp-up and sustain high performance fully non-inductive plasmas with large bootstrap fraction and enhanced confinement enabled via the low aspect ratio, high beta configuration. With its unique capabilities, NSTX-U research also supports ITER and other critical fusion development needs. Super-Alfvénic ions in beam-heated NSTX-U plasmas access energetic particle (EP) parameter space that is relevant for both α-heated conventional and low aspect ratio burning plasmas. NSTX-U can also generate very large target heat fluxes to test conventional and innovative plasma exhaust and plasma facing component solutions. This paper summarizes recent analysis, theory and modelling progress to advance the tokamak physics basis in the areas of macrostability and 3D fields, EP stability and fast ion transport, thermal transport and pedestal structure, boundary and plasma material interaction, RF heating, scenario optimization and real-time control.

NSTX-U theory, modeling and analysis results

Abstract The mission of the low aspect ratio spherical tokamak NSTX-U is to advance the physics basis and technical solutions required for optimizing the configuration of next-step steady-state tokamak fusion devices. NSTX-U will ultimately operate at up to 2 MA of plasma current and 1 T toroidal field on axis for 5 seconds, and has available up to 15 MW of Neutral Beam Injection (NBI) power at different tangency radii and 6 MW of High Harmonic Fast Wave (HHFW) heating. With these capabilities NSTX-U will develop the physics understanding and control tools to ramp-up and sustain high performance fully non-inductive plasmas with large bootstrap fraction and enhanced confinement enabled via the low aspect ratio, high beta configuration. With its unique capabilities, NSTX-U research also supports ITER and other critical fusion development needs. Super-Alfvénic ions in beam-heated NSTX-U plasmas access energetic particle parameter space that is relevant for both -heated conventional and low aspect ratio burning plasmas. NSTX-U can also generate very large target heat fluxes to test conventional and innovative plasma exhaust and plasma facing component (PFC) solutions. This paper summarizes recent analysis, theory and modelling progress to advance the tokamak physics basis in the areas of macrostability and 3D fields, energetic particle stability and fast ion transport, thermal transport and pedestal structure, boundary and plasma material interaction, RF heating, scenario optimization and real-time control.

Hot-Tail Runaway Seed Landscape during the Thermal Quench in Tokamaks.

Runaway electron populations seeded from the hot tail generated by the rapid cooling in plasma-terminating disruptions are a serious concern for next-step tokamak devices such as ITER. Here, we present a comprehensive treatment of the thermal quench, including the superthermal electron dynamics, heat and particle transport, atomic physics, and radial losses due to magnetic perturbations: processes that are strongly linked and essential for the evaluation of the runaway seed in disruptions mitigated by material injection. We identify limits on the injected impurity density and magnetic perturbation level for which the runaway seed current is acceptable without excessive thermal energy being lost to the wall via particle impact. The consistent modeling of generation and losses shows that runaway beams tend to form near the edge of the plasma, where they could be deconfined via external perturbations.

Evaluation of tritium burnup fraction for CFETR scenarios with core-edge coupling simulations

A key mission for the next-step fusion tokamak device China Fusion Engineering Test Reactor (CFETR) is demonstrating tritium self-sufficiency, which requires a sufficiently high tritium burnup fraction (fburnup) in order to match a practically achievable tritium breeding ratio (TBR) with the blanket design constraints. Core-edge coupling simulations are performed to investigate the dependence of fburnup on different controlling parameters for CFETR scenarios. Core plasma profiles with a range of pedestal densities are simulated by consistent iterative calculations of equilibrium, transport, auxiliary heating and current drives within the OMFIT framework. The core-SOL integrated COREDIV code is then used to evaluate fburnup with the OMFIT modelled core plasma parameters as input. According to the simulations, fburnup can be effectively increased with a higher pedestal density on account of the fusion power increasing faster than the fueling source required to maintain steady-state. Higher can also increase the fburnup due to increase of fuel recycling. Deeper pellet fueling deposition and lower ratio of particle to thermal diffusivities D/χ can both increase the effective particle confinement time and thus fburnup. However, the effect of helium and other impurities (Ar and W) is shown to reduce fburnup for comparable impurity and main ion transport. Based on our analysis, using present pellet fueling technology, achieving fburnup > 3% for CFETR will be very challenging. This is a lower limit for the required TBR (>1) to match the achievable TBR for tritium self-sufficiency. Furthermore, our study suggests that if fueling can penetrate deeper than r/a < 0.8 under optimistic conditions, the required burnup fraction could be attainable. The modelling results thus provide important suggestions and implications for the optimization of CFETR scenarios and development of advanced fueling systems.

Filamentary plasma eruptions and their control on the route to fusion energy

The tokamak is the most advanced approach to fusion and is approaching operation under power-plant conditions, promising sustainable, low-emission, baseload power to the grid. As the heating power of a tokamak is increased above a threshold, the plasma suddenly bifurcates to a state of high confinement, creating a region of plasma with a large pressure gradient at its edge. This bifurcation results in a repetitive sequence of explosive filamentary plasma eruptions called edge-localized modes (ELMs). ELMs on next-step tokamaks, such as ITER, will likely cause excessive erosion to plasma-facing components and must be controlled. We present what is understood about how ELMs form, their filamentary nature and the mechanisms that transport heat and particles to the first wall of the tokamak. We also discuss methods to control ELMs, including magnetic perturbations. Substantial progress has been made towards making fusion a source of commercial energy. One challenge is the damage that filamentary plasma eruptions cause to plasma-facing components. This Perspective summarizes what is understood about these eruptions and strategies for controlling them.

Non-plasma diagnostic instrumentation for a next-step tokamak with a burning plasma

Any device to succeed ITER and demonstrate successful engineering implementation of a first wall and tritium generation will require very extensive instrumentation of the device components. Measurements of slow and fast movements of components, strains and stresses on them, temperatures and fluid flow parameters will all be necessary. While it is hoped that the physics of the plasma will become fully known during the life of such a device (enabling a major reduction in the complexity of plasma diagnostics) qualification of the engineered components will be the primary output of the operation. Survival of the first wall for long-pulse full-power operational periods, breeding sufficient tritium to maintain operation, reliable behavior of coils and the auxiliary systems of fueling, heating and remote maintenance equipment, must all be demonstrated. This paper puts together a preliminary estimate of the instrumentation that will be necessary to meet this requirement for a specific example of a tokamak design.

Comparison benchmark between tokamak simulation code and TokSys for Chinese Fusion Engineering Test Reactor vertical displacement control design**Project supported by the National Natural Science Foundation of China (Grant Nos. 11305216, 11305209, and 11375191), the National Magnetic Confinement Fusion Research Program of China (Grant Nos. 2014GB103000 and 2014GB110003), and External Cooperation Program of BIC, Chinese Academy of Sciences (Grant No. GJHZ201303).

Vertical displacement event (VDE) is a big challenge to the existing tokamak equipment and that being designed. As a Chinese next-step tokamak, the Chinese Fusion Engineering Test Reactor (CFETR) has to pay attention to the VDE study with full-fledged numerical codes during its conceptual design. The tokamak simulation code (TSC) is a free boundary time-dependent axisymmetric tokamak simulation code developed in PPPL, which advances the MHD equations describing the evolution of the plasma in a rectangular domain. The electromagnetic interactions between the surrounding conductor circuits and the plasma are solved self-consistently. The TokSys code is a generic modeling and simulation environment developed in GA. Its RZIP model treats the plasma as a fixed spatial distribution of currents which couple with the surrounding conductors through circuit equations. Both codes have been individually used for the VDE study on many tokamak devices, such as JT-60U, EAST, NSTX, DIII-D, and ITER. Considering the model differences, benchmark work is needed to answer whether they reproduce each other’s results correctly. In this paper, the TSC and TokSys codes are used for analyzing the CFETR vertical instability passive and active controls design simultaneously. It is shown that with the same inputs, the results from these two codes conform with each other.

Overview of physics results from the conclusive operation of the National Spherical Torus Experiment

Research on the National Spherical Torus Experiment, NSTX, targets physics understanding needed for extrapolation to a steady-state ST Fusion Nuclear Science Facility, pilot plant, or DEMO. The unique ST operational space is leveraged to test physics theories for next-step tokamak operation, including ITER. Present research also examines implications for the coming device upgrade, NSTX-U. An energy confinement time, τE, scaling unified for varied wall conditions exhibits a strong improvement of BTτE with decreased electron collisionality, accentuated by lithium (Li) wall conditioning. This result is consistent with nonlinear microtearing simulations that match the experimental electron diffusivity quantitatively and predict reduced electron heat transport at lower collisionality. Beam-emission spectroscopy measurements in the steep gradient region of the pedestal indicate the poloidal correlation length of turbulence of about ten ion gyroradii increases at higher electron density gradient and lower Ti gradient, consistent with turbulence caused by trapped electron instabilities. Density fluctuations in the pedestal top region indicate ion-scale microturbulence compatible with ion temperature gradient and/or kinetic ballooning mode instabilities. Plasma characteristics change nearly continuously with increasing Li evaporation and edge localized modes (ELMs) stabilize due to edge density gradient alteration. Global mode stability studies show stabilizing resonant kinetic effects are enhanced at lower collisionality, but in stark contrast have almost no dependence on collisionality when the plasma is off-resonance. Combined resistive wall mode radial and poloidal field sensor feedback was used to control n = 1 perturbations and improve stability. The disruption probability due to unstable resistive wall modes (RWMs) was surprisingly reduced at very high βN/li > 10 consistent with low frequency magnetohydrodynamic spectroscopy measurements of mode stability. Greater instability seen at intermediate βN is consistent with decreased kinetic RWM stabilization. A model-based RWM state-space controller produced long-pulse discharges exceeding βN = 6.4 and βN/li = 13. Precursor analysis shows 96.3% of disruptions can be predicted with 10 ms warning and a false positive rate of only 2.8%. Disruption halo currents rotate toroidally and can have significant toroidal asymmetry. Global kinks cause measured fast ion redistribution, with full-orbit calculations showing redistribution from the core outward and towards V∥/V = 1 where destabilizing compressional Alfvén eigenmode resonances are expected. Applied 3D fields altered global Alfvén eigenmode characteristics. High-harmonic fast-wave (HHFW) power couples to field lines across the entire width of the scrape-off layer, showing the importance of the inclusion of this phenomenon in designing future RF systems. The snowflake divertor configuration enhanced by radiative detachment showed large reductions in both steady-state and ELM heat fluxes (ELMing peak values down from 19 MW m−2 to less than 1.5 MW m−2). Toroidal asymmetry of heat deposition was observed during ELMs or by 3D fields. The heating power required for accessing H-mode decreased by 30% as the triangularity was decreased by moving the X-point to larger radius, consistent with calculations of the dependence of E × B shear in the edge region on ion heat flux and X-point radius. Co-axial helicity injection reduced the inductive start-up flux, with plasmas ramped to 1 MA requiring 35% less inductive flux. Non-inductive current fraction (NICF) up to 65% is reached experimentally with neutral beam injection at plasma current Ip = 0.7 MA and between 70–100% with HHFW application at Ip = 0.3 MA. NSTX-U scenario development calculations project 100% NICF for a large range of 0.6 < Ip(MA) < 1.35.

Electrostatic dust detector for fusion devices with improved sensitivity

Methods to measure the inventory of dust particles and to remove dust if it approaches safety limits will be required in next-step tokamaks such as ITER. An electrostatic dust detector, based on a fine grid of interlocking circuit traces, biased to 30 or 50 V, has been developed for the detection of dust on remote surfaces in air and vacuum environments. Gaining operational experience of dust detection on surfaces in tokamaks is important, however the level of dust generated in contemporary short-pulse tokamaks is comparatively low and high sensitivity is necessary to measure dust on a shot-by-shot basis. We report on modifications in the detection electronics that have increased the sensitivity of the electrostatic dust detector by a factor of up to 120, – a level suitable for measurements on contemporary tokamaks.

Controlling surface dust in a tokamak

Methods to measure the inventory of dust particles and to remove dust if it approaches safety limits will be required in next-step tokamaks. A novel electrostatic dust detector, based on a fine grid of interlocking circuit traces, biased to 30 or 50V, has been developed for the detection of dust on remote surfaces in air and vacuum environments. Impinging dust particles are detected when they create a short circuit between the traces, however this short circuit is temporary suggesting the device may be useful for the removal of dust from specific areas. The fate of the dust particles has been tracked by measurements of mass gain/loss. Heating by the current pulse caused up to 90% of the particles to be ejected from the grid or vaporized, the removal efficiency depending on the experimental geometry. We also report the first attempt at real-time dust detection in NSTX.

Dust-particle transport in tokamak edge plasmas

Dust particulates in the size range of 10nm–100μm are found in all fusion devices. Such dust can be generated during tokamak operation due to strong plasma∕material-surface interactions. Some recent experiments and theoretical estimates indicate that dust particles can provide an important source of impurities in the tokamak plasma. Moreover, dust can be a serious threat to the safety of next-step fusion devices. In this paper, recent experimental observations on dust in fusion devices are reviewed. A physical model for dust transport simulation and a newly developed code DUSTT are discussed. The DUSTT code incorporates both dust dynamics due to comprehensive dust-plasma interactions as well as the effects of dust heating, charging, and evaporation. The code tracks test dust particles in realistic plasma backgrounds as provided by edge-plasma transport codes. The results are presented for dust transport in current and next-step tokamaks. The effect of dust on divertor plasma profiles and core plasma contamination is examined.

Radio-Frequency Current Drive in DIII-D

Two methods of radio-frequency (rf) current drive that are well suited to controlling and sustaining the current profile in burning plasma experiments have been studied in the DIII-D tokamak. Fast-wave current drive (FWCD) gave centrally peaked current densities that increased linearly with central electron temperature. While high harmonic absorption of the fast waves on energetic beam ions could reduce the available power for current drive, FWCD figures of merit as high as γFW = 0.5 × 1019 A/m2·W were still achieved. Electron cyclotron current drive (ECCD) was shown to be localized to the region of power deposition, with a current drive efficiency that decreased as the magnetic well depth increased. The detrimental effect of the magnetic well could be mitigated by raising the electron beta. ECCD figures of merit as high as γEC = 0.5 × 1019 A/m2·W were measured for central deposition. The experimental FWCD and ECCD were both extensively tested against theoretical models and were found to be in excellent agreement. Validation of these predictive models of rf current drive aids in scenario development for next-step tokamaks.

Suppression of large edge-localized modes in high-confinement DIII-D plasmas with a stochastic magnetic boundary.

A stochastic magnetic boundary, produced by an applied edge resonant magnetic perturbation, is used to suppress most large edge-localized modes (ELMs) in high confinement (H-mode) plasmas. The resulting H mode displays rapid, small oscillations with a bursty character modulated by a coherent 130 Hz envelope. The H mode transport barrier and core confinement are unaffected by the stochastic boundary, despite a threefold drop in the toroidal rotation. These results demonstrate that stochastic boundaries are compatible with H modes and may be attractive for ELM control in next-step fusion tokamaks.

Heating and current drive by electron cyclotron waves

The physics model of electron cyclotron heating (ECH) and current drive (ECCD) is becoming well validated through systematic comparisons of theory and experiment. This work has shown that ECH and ECCD can be highly localized and robustly controlled in toroidal plasma confinement systems, leading to applications including stabilization of magnetohydrodynamic instabilities like neoclassical tearing modes, control and sustainment of desired profiles of current density and plasma pressure, and studies of localized transport in laboratory plasmas. The experimental work was supported by a broad base of theory based on first principles which is now well encapsulated in linear ray tracing codes describing wave propagation, absorption, and current drive and in fully relativistic quasilinear Fokker–Planck codes describing in detail the response of the electrons to the energy transferred from the wave. The subtle balance between wave-induced diffusion and Coulomb relaxation in velocity space provides an understanding of the effects of trapping of current-carrying electrons in the magnetic well. Strong quasilinear effects and radial transport of electrons, which may broaden the driven current profile, have also been observed under some conditions and appear to be consistent with theory, but in large devices these are usually insignificant. The agreement of theory and experiment, the wide range of established applications, and the technical advantages of ECH support the application of ECH in next-step tokamaks and stellarators.

ELMy H-modes in JET helium-4 plasmas

ELMy H-modes in helium-4 plasmas provide valuable information on ELMy H-mode physics as well as a possible early low activation operational phase for next-step tokamaks, such as ITER. With this in mind, a series of helium-4 H-mode experiments were performed on JET with pure helium-4 NBI auxiliary heating (up to 12 MW). A set of ELMy H-mode plasmas were produced, in both the Type I ELM regime and a second regime, which showed characteristics similar to the deuterium Type III regime, but with a reverse ELM frequency dependence on power. Sawteeth were also observed, and had similar behaviour to those seen in deuterium. Compared with deuterium plasmas, Type I ELMy H-mode confinement is seen to be 28 ± 6% poorer in helium-4 plasmas and the L–H power threshold 42 ± 10% larger. This is the opposite of the behaviour predicted by experimental isotope mass scalings from hydrogenic plasmas. Comparison with a wider hydrogenic database, enables the effects of isotopic charge and mass to be studied independently.