The roadmap for the commissioning and first operations of superconductive tokamaks envisages the possibility of running discharges with fairly elongated plasmas before the complete installation of the in-vessel components, including vertical stabilization coils, or any other specific sets of coils to be used for the magnetic control of fast transients.
In the absence of dedicated...
Steady, non-disruptive tokamak plasmas have been produced in the Madison Symmetric Torus (MST) with an electron density up to an order of magnitude above the Greenwald limit [Hurst et al., PRL, accepted for publication]. This result is made possible in part by a high-voltage, feedback-controlled power supply driving the toroidal plasma current. Also important may be the thick, stabilizing,...
A novel database study of the L-mode Density Limit (LDL) in metal- and carbon-wall devices (Alcator C-Mod, AUG, DIII-D, and TCV) identifies a two-variable, dimensionless stability boundary that predicts the LDL with significantly higher accuracy than the widely-utilized Greenwald limit. Historically, there has been broad interest in understanding the operational boundary imposed by the...
Resistive wall tearing modes (RWTM) are closely related to resistive wall modes (RWMs). RWTMs are tearing modes whose linear and nonlinear growth rate depend on the resistive wall penetration time.
The consequence for ITER, with wall penetration time of $250 ms,$ compared to $ \sim 5 ms$ in JET and DIII-D, is that the thermal quench
timescale could be much longer than previously...
The maximum allowable vertical displacement which can be recovered by the magnetic control system is a fundamental quantity for tokamak magnetic control (Gribov 2015 Nucl. Fusion 55 073021). This figure of merit is usually defined relying on a mass-less assumption, i.e. the reaction currents in wall structures are considered to vary in order to guarantee MHD equilibrium during the plasma...
Disruption prediction and avoidance is critical to maintain steady plasma operation and to avoid damage to device components in ITER and reactor-scale tokamaks. Physics-based disruption event characterization and forecasting (DECAF*) research determines the relation of events leading to disruption, and aims to provide event onset forecasts with high accuracy and sufficiently early warning to...
Vertical displacement events (VDEs) in tokamaks involve large displacements of the plasma magnetic axis from the vessel midplane, often leading to disruptions. These events are of particular concern for their potential to cause damage to plasma-facing components, as well as large forces on the vessel due to halo currents generated during the disruption that run through the plasma and vessel...
The structural integrity of the Vacuum Vessel (VV) of Pakistan's Metallic Tokamak-I (MT-I), a small spherical tokamak, was tested by simulating a 10 ms input current event on a 180° sector model. During this event, the energy from the plasma is entirely transferred to the VV's first wall. This study is based on the law of conservation of energy, demonstrating that the extent of damage is...
A pulsed hydrogen plasma stream is produced from a pulsed plasma accelerator (PPA) powered by 200 KJ Pulsed Power System (PPS). The PPS, which consists of two modules capacitor banks, is charged up to 15 kV to generates a peak discharge current of 100 kA for a half time period of 500 µs. The high voltage from the capacitor banks thus applied in between two coaxially positioned electrodes to...
Disruption of a tokamak plasma is a multi-step process in which the loss of the plasma vertical position control is often among the last events that precede the final plasma deconfinement. A flow of current between the plasma and the vessel components -the halo current [1]- is generated through the contact of vertically displaced plasma with the vacuum vessel, resulting in electromagnetic...
We present a groundbreaking multimodal neural network model designed for diagnostics resolution enhancement, which innovatively leverages inter-diagnostic correlations within a system. Traditional approaches have primarily focused on unimodal enhancement strategies, such as pixel-based image enhancement or heuristic signal interpolation. In contrast, our model employs a novel methodology by...
In the past three years, a shattered pellet injection (SPI) system designed for disruption mitigation on the EAST tokamak was successfully developed and integrated into EAST tokamak in 2022. The SPI system is capable of producing Ne pellets with diameters of ~ 5 mm and lengths ranging from 7 to 15 mm. The material gas consumption is approximately 20, 25, and 30 Pa·m3, respectively, and the...
Plasma major disruptions pose severe threats to the device integrity in future operations of International Thermonuclear Experimental Reactor (ITER). They can cause dangerous excessive electromagnetic forces, heat loads and generation of the intense beams of relativistic runaway electrons (RE). Localized interaction of intense RE beams with surrounding plasma facing components (PFC) inevitably...
In this work we report measurements of the temperature and density of the halo current region on DIII-D during disruptions using the recently upgraded Thomson scattering diagnostic allowing low-temperature measurements down to a few eV with a sub-ms repetition rate. This is done by employing deliberate downward vertical displacement events (VDEs) and relying on the expansion of the halo...
Dispersive shell pellet (DSP) injection is currently being developed as an alternative disruption mitigation technique to massive gas injection and shattered pellet injection. The main advantage of DSP injection is the core deposition of the payload which is expected to result in higher assimilation fractions and an inside-out thermal quench (TQ). DSPs have been successfully launched into...
Disruptions are an inherent property of tokamak plasmas, which cannot be completely eliminated. The consequences of disruptions are especially dangerous for large machines like ITER and even more so for DEMO. Thermal Quench (TQ) is the initial phase of disruption followed by plasma Current Quench (CQ). Essential diagnostics for the TQ are magnetics (dB/dt), Electron Cyclotron Emission (ECE)...
The Learning Using Privileged Information (LUPI) paradigm allows training classifiers with data not available at execution time. Recently, an application of the LUPI paradigm to the prediction of disruptions with extreme data scarcity was demonstrated [J. Vega et al. Nuclear Fusion 64 (2024) 046010 (12 pp)]. The objective of the previous reference was to test the development of an adaptive...
The ITER Disruption Mitigation System (DMS) is based on Shattered Pellet Injectors (SPI), which accelerates a large protium, neon or mixture pellet with high pressure gas and shatters it prior to the entrance into the plasma, creating a plume of smaller pellet fragments. The ITER DMS Support Laboratory is part of the ITER DMS Task Force programme to establish the physics and technology basis...
Future tokamaks will require disruption mitigation systems (DMS) to prevent machine damage during the uncontrolled loss of plasma confinement. Massive impurity injection, particularly shattered pellet injection (SPI), is the leading candidate for a DMS. Validated, predictive models are needed to project these systems to future devices, which require models for the macroscopic plasma evolution...
Weibel instability due to nonlinear inverse bremsstrahlung absorption (WINLIBA) in magnetized plasma has been investigated in the frame of the relativistic kinetic theory (RKT). In this study the magnetized plasma is described by relativistic Fokker-Planck equation with an ameliorated Krook collision term which takes into account the relativistic effect and the Landau microscopic collision...
Currently machine learning disruption predictor is the most promising way of solving the disruption mitigation triggering problem. But it does need data from the target machine to be trained. However, the future machine may not be able to provide enough data both in quality and quantity to satisfy the training. In this paper we first explained why just simply mixing limited data from target...
Disruptions are one of the most critical issues in tokamak operation. In fact, the rapid termination of plasma magnetic confinement leads to significant heat and electromagnetic loads on the plasma-facing components, threatening the integrity of the reactor. Moreover, continuous early terminations of plasma discharge can cause variations in the energy production, limiting the amount of energy...
Artificial Intelligence (AI) techniques, such as Machine learning and Deep Learning, have been extensively investigated for the construction of disruptions predictive models in tokamaks. Although the excellent performance has demonstrated the applicability of the paradigm to the experimental machines currently in service, the development of cross-tokamak models is still in its infancy [1]....
Rapid plasma dynamics preceding some disruptions in tokamak devices can be inferred through the electron temperature profile evolution due to the fast thermal transport along the field lines. In particular, local collapses in the electron temperature profile are the signature of nonlinear events such as flux surface tearing, observed due to the sudden thermal transport that follows changes in...
This work explores the development and preliminary calibration of an off-normal warning system for SPARC, the aim of which is to minimize disruption loads and maximize operation time via the detection, interpretation, and pacification (i.e. avoidance and mitigation) of anomalous events. Similar systems have been implemented for existing tokamaks like DIII-D [1], NSTX [2], and TCV [3], but the...
In support of the ITER DMS development, a highly flexible SPI system [[1], [2]] was installed at ASDEX Upgrade (AUG). It offers the unique opportunity to investigate the effect of different fragment size and velocity distributions — which were characterised beforehand in extensive laboratory tests — on the disruption behaviour. The triple barrel setup with independent freezing cells, injection...
Previous investigations on JET suggest thermal stored energy ($W_{th}$) is poorly mitigated by either Massive Gas Injection (MGI) or Shattered Pellet Injection (SPI) Disruption Mitigation Systems (DMS), when measured by weighted averages of bolometer channels. A contrasting investigation on ASDEX-Upgrade found that thermal energy is well mitigated with MGI. We investigate whether the apparent...
A series of experiments is underway to explore the effect of both self-excited and externally launched plasma waves on relativistic electrons (REs) across a wide range of geometries and plasma parameters. While O and X-mode waves are routinely used for heating and current-drive in tokamaks they are incapable of directly resonating with REs since their phase velocity is much greater than the...
Runaway electrons generated during tokamak disruptions are a major concern for the safe operation of future tokamaks. These energetic electrons can carry significant current and cause severe damage to a tokamak. Therefore, mitigating runaway electrons is essential for the safety and efficiency of fusion devices. Interaction of runaway electrons with waves is one of the potential mechanisms for...
The heat flux mitigation during the Thermal Quench (TQ) by the Shattered Pellet Injection (SPI) is one of the major elements of disruption mitigation strategy for ITER. It's efficiency greatly depends on the SPI and the target plasma parameters, and is ultimately characterised by the heat deposition on to the Plasma Facing Components (PFCs). To investigate such heat deposition, JOREK...
The disruption mitigation system (DMS) for ITER is based on the shattered pellet injection (SPI) technology. The principle of operation is to form cylindrical cm-sized cryogenic pellets and accelerate them to high speeds towards a shattering chamber, where the pellets disintegrate into a plume of fragments of different sizes and velocities, which then enter the plasma for the mitigation...
The ITER disruption mitigation support laboratory is part of the ITER Disruption Mitigation System (DMS) Task Force programme to establish the physics and technology basis for the ITER DMS. The laboratory is located at the HUN-REN Centre for Energy Research (CER), Budapest Hungary. The aims include production, launching and shattering of 28.5x57 mm (d x L) H, D, Ne and mixture pellets,...
Localized wall damage from post-disruption runaway electron (RE) wall impact is a significant concern for future large tokamaks. One possible method for reducing this wall damage in the event of an unavoidable RE-wall impact is massive injection of low-Z (H2 or D2) gas. This injection can have the effect of partially recombining the cold thermal background plasma, resulting in a greatly...
Runaway electrons of MeV and higher energies can dominate the plasma
current during ITER startup and the current quench phase of a major
disruption. The plasma regime spans from reasonably low-density and
high-temperature (startup) to high-density and low-temperature
(disruption mitigated by high-Z impurities), and somewhere in between
(disruption mitigation by low-Z injection). Here we...
Realizing tokamak power plants requires reducing the frequency and impact of disruptions sufficiently to accept them as a part of operations. SPARC is a high field tokamak [1] ($B_o=12.2$ T, $I_p=8.7$ MA) designed to demonstrate Q>1 and to explore divertor and disruption solutions for the ARC power plant. Significant disruption work is ongoing in hardware, software, operational planning, and...
Safe termination of the plasma discharge using injection of intense gas flows and macroparticles (pellets) is considered as the main system for preventing development of the runaway electron beams in tokamak-reactor (ITER) [1]. One of the main limitations in the use of these systems in large-scale tokamaks is the weak penetration of injected gas and particles into the central zones of the high...
The potential of localized heat loads under disruptions to cause considerable melting of plasma-facing components (PFC) has been extensively investigated. Two distinctive regimes exist, which lead to different types of PFC damage and necessitate different modelling approaches; surface loading and volumetric loading.
Surface heating is caused by electrons and ions with energies in the...
Extreme high transient heat flux up to thousands of MW/m2 in a short pulse (~ms) during disruption in future large scale tokamak imposes great challenge on plasma facing components (PFCs), which is very concerned and worried by the ITER. Currently, understanding the consequence of thermal damage behaviors on W PFC by is an critical issue. EAST, as a superconducting tokamak, is installed metal...
During tokamak disruptions, the magnetic surfaces are broken creating large regions of chaotic magnetic field lines. The physics associated with post-disruption chaotic magnetic fields needs be understood to address force, heat, and runaway electron loading on the walls. Direct simulations are too challenging to allow parameter scans and have uncertainties that can only be addressed by a...
Since the RE generation during tokamak disruption is exponentially sensitive to initial plasma current, highly energetic RE beams pose a critical challenge for future tokamaks. Accurate simulations of tokamak disruptions are therefore essential for the development of successful mitigation strategies and safe operation. However, when simulating such disruptions, fluid plasma models are often...
In 1996, during a disruption, 3 hundred tonnes of JET vacuum vessel moved by 7 mm sideways. In spite of significant efforts to understand the phenomena, the horizontal force on the tokamak wall during plasma disruptions still remains poorly understood. For example, the predictions for ITER vary greatly, from 2 to 60 MN, with the upper estimate exceeding the design margin of 48 MN. To resolve...
The talk begins with an overview of several consequent steps in evaluation of the pulsed EM loads in tokamaks and concentrates on calculation of AVDE-induced loads in ITER. The presented practical EM model uses the superposition of two patterns of the halo current: one perfectly symmetric and another perfectly anti-symmetric. It combines the following features of two recent trial models: (a)...
Electromagnetic (and thermal) loads during vertical displacement events (VDEs) are of major concerns in tokamaks, and in future pilot plants that are based on this concept. In particular, Alcator C-Mod has been used to study disruptions from planned as well as unexpected VDEs, and where halo currents were well analyzed in the disrupting plasma [1,2]. Therefore, it offers comprehensive hot and...
In tokamaks, eddy currents and associated forces, driven by rapid current quenches during disruptions are important drivers for structural engineering requirements. Additionally, recent interest in disruption-driven 3D currents, such as the Runaway Electron Mitigation Coil (REMC) concept, further motivates the need to capture currents in passive conducting structures early in and throughout...
During the thermal quench (TQ), the stored thermal energy is released with a short timescale and might cause serious damage to plasma-facing components (PFCs), especially in future large-scale tokamaks. Here presents the detailed description of the TQ database consisting of 164 disruption discharges, including both major disruptions (MDs) and hot vertical displacement events (VDEs), on EAST...
