Plasma Disruption Research Articles

Thermal Effects on Dynamic Behavior of CFETR Vacuum Vessel

The vacuum vessel (VV) of the China Fusion Engineering Test Reactor (CFETR) was designed to withstand the variable structural loads resulting from dynamic excitation, such as seismic and plasma disruption. Modal analysis, a powerful tool, was used to evaluate the structural dynamic characteristics, such as frequency and mode shape. In addition, the CFETR VV has three different temperature conditions: room temperature (20°C), normal operation temperature (100°C), and baking temperature (200°C). In this paper, in order to investigate the influence of such different temperature conditions to the dynamic behavior of the VV, three independent finite element analysis with the same modal analysis method were performed. According to analysis results, there are obvious thermal effects on the dynamic behavior, such as nature frequency and mode shape, among the different temperature cases for the CFETR VV. Moreover, the results show that the natural frequency for each order decreases as the temperature increases, and the mode shape of the VV also changed with temperature.

Electro dynamic model of eddy currents in EU DEMO TF coil casing during major plasma disruption

The conceptual design of the EU DEMO reactor is currently ongoing within the EUROfusion consortium. Many different fault transients must be considered and carefully analyzed in the design phase; one of the most severe is the major plasma disruption (MPD), which causes several drawbacks on the magnet system. During a disruption, the plasma current decreases extremely fast, causing a fast variation of the magnetic field, which in turn induces an electric field. In presence of conductive materials, e.g., coil casing and vacuum vessel (VV), the electric field induces large eddy currents which deposit power by Joule effect. The conductive regions are tightly coupled on different timescales through the magnetic field induced by the eddy currents: the eddy currents in the VV influence the magnetic field evolution in the TF coil casing, thus affecting the power deposition in the latter. The aim of this work is the evaluation of the power deposited within the TF coil casing during a major plasma disruption due to eddy currents. The power deposition has been evaluated by means of the 3D-FOX, a finite element (FE) tool developed at Politecnico di Torino. The computed power deposition is used as input to the thermal-hydraulic (TH) simulation, performed with the 4C code, with the aim of assessing the erosion of the temperature margin given by MPD.

The advance of magnetic diagnostics system in support of EAST long-pulsed operation

In EAST long-pulsed discharge (hundreds of seconds), electric magnetic diagnosis (EMD) is very important, since EMD not only monitors tokamak security status but also provides accurate measurement accuracy for reconstruction of the plasma boundary. To avoid current measurement drift, a fiber optic current sensor, based on the Faraday effect, is developed and used for poloidal and plasma current feedback control for the first time, relative current measurement accuracy is within 0.5%. To ensure plasma boundary control accuracy, a detailed set of magnetic measurement calibration methods is developed before the plasma discharge. The maximum relative error is less than 1%, the corresponding control accuracy is within 1 cm. To minimize integrator drift error, a long-pulse integrator test is essential, the corresponding drift error needs to be subtracted in plasma control system. Besides, the saddle coil and Mirnov coil not only help to detect MHD issues, but are also utilized for plasma disruption prediction during the long-pulse discharge.

Analysis of the transient heat flux on divertor surface during plasma disruption in EAST

Analysis of the transient heat flux on divertor surface during plasma disruption in EAST

SPARC as a platform to advance tokamak science

The unique capabilities of the SPARC tokamak mean that it has the potential to contribute significantly to tokamak science and plasma physics, motivating further collaboration and broader data access beyond the CFS and MIT teams. SPARC is a compact, high-field tokamak that is currently under construction and is predicted to achieve burning plasma conditions once in operation. SPARC experimental data has the potential to advance the understanding of many aspects of tokamak physics, including but not limited to confinement and stability at high field and high density, burning plasma physics, disruption physics, and boundary physics and heat flux management in power plant-relevant conditions. The SPARC team is already a combination of members from Commonwealth Fusion Systems (CFS), a privately funded company, and the Massachusetts Institute of Technology (MIT), a non-profit university. This article describes the opportunities for the SPARC team to work with other researchers to advance toward a fusion power plant on the fastest possible time scales and to simultaneously broaden scientific understanding of plasma physics, meeting the missions of both CFS and academic partners.

First results of Ne shattered pellet injection for mitigating plasma disruption with full metal wall in EAST tokamak

Disruption mitigation poses a significant and unresolved challenge for ITER and future fusion reactor devices. To mitigate the effect of plasma disruption, a Shattered Pellet Injection (SPI) system has been developed and the first rapid shutdown experiments have been successfully performed in the Experimental Advanced Superconducting Tokamak with full metal wall. The experiments confirmed the system’s capability to inject 5 mm diameter neon (Ne) pellets at velocities ranging from 100 to 400 m s−1, with a maximum injected quantity of 13.2 Pa m3. Compared to unmitigated disruptions, the total radiation power was significantly higher with the respective use of SPI and Massive Gas Injection (MGI). Additionally, the radiation distribution and divertor heat flux were compared between SPI and MGI methods. The results demonstrated that SPI exhibited shorter cooling time, stronger core radiation and more uniform poloidal radiation distribution compared to MGI, indicating deeper deposition using SPI. Furthermore, during plasma shutdown, the electron temperature and peak heat flux near outer divertor strike points were reduced by 40% and 50% respectively, with SPI, in comparison to MGI. These findings serve as a valuable reference for implementing SPI technology as the baseline approach for disruption mitigation in ITER.

OpenFOAM modeling of beryllium melt motion and splashing from first wall in ITER

Beryllium (Be) is a material which will be used as a plasma facing component in ITER due to its unique properties of high thermal conductivity, low density, and high strength. However, under extreme conditions of high temperature and pressure, Be can melt at the surface of tiles and molten droplets can be ejected into the reactor leading to disruption of fusion plasma. The pressure, mass density, velocity of Be vapor, and variations of temperature at the melt layer interface can influence the splashing of Be melt. The Computational Fluid Dynamics (CFD) model based on the OpenFOAM toolbox, a free open source CFD software package, was developed to treat the coupled flow of liquid Be metal and its vapor. The vapor-melt interface is modeled using the volume of fluid (VOF) approach implemented in the interCondensatingEvaporatingFoam solver that solves the continuity, momentum, heat conduction, and VOF equations. This CFD model is capable to predict the hydrodynamic effects of Be vapor on the melt layer motion, splashing, non-linear growth of melt waves, and ejection of molten droplets. The modeling accounts for the effects of thermal, viscous, gravitational, and surface tension forces at the vapor-melt interface. In this research, we used the interCondensatingEvaporatingFoam solver to simulate the effects of Be phase change and the development of melt motion with the droplets ejected from the surface. The CFD model accounts for inter-phase change between Be liquid and Be vapor. The evaporation model was validated against the Stefan phase-change problems. The influence of heat and mass transfer across the vapor-melt interface on melt layer stability is also investigated. The results provide an understanding of how the rate of phase change affects the development of melt structures and waves at the vapor-melt interface.

Engineering design and analysis of first wall on the high-magnetic-field-side for HL-2M tokamak

The first wall is one of the key in-vessel components, which would be employed for protecting the vacuum vessel and other in-vessel components (e.g. diagnostic components, gas injection system, resonant magnetic perturbation coils). The first wall is facing the high temperature plasma, and the graphite is adopted as plasma facing material for HL-2M tokamak due to its many excellent properties including high thermal conductivity, good thermal shock, thermal fatigue resistance and good compatibility with plasma. The first wall on the high-magnetic-field-side is consisted of the graphite tile, the back plate, supports and the adjustment plate, and no active cooling structure is considered except for the residual neutral beam power region. All of the first wall parts are assembled with different bolts, and they are easy to install and maintain during their running life. All of the analysis results based on the finite element method show that the structural design of the first wall on the high-magnetic-field-side has good heat transfer performance, satisfying the thermal design requirements during the plasma operation stage (the average value of heat load in the residual neutral beam power 1.0 MW/m2, thermal load in other region 0.3 MW/m2, pulse time 5 s) for HL-2M device. The first wall can also withstand the impact of the electromagnetic load caused by 1.8 MA plasma disruption, including the major disruption and the vertical displacement events.

Disruption prediction on EAST with different wall conditions based on a multi-scale deep hybrid neural network

Plasma disruption is a very dangerous event for future tokamaks and fusion reactors. Therefore, predicting disruption is crucial for ensuring the safety and performance of reactors. In this study, the features of two deep learning algorithms are integrated to establish a multi-scale hybrid network disruption predictor. Firstly, 43 diagnostic signals are extracted by a convolutional neural network (CNN), and the time information is learned by a long short-term memory network. The predictor is trained and tested on a database containing non-all-metal wall discharges. Its area under the receiver operator characteristic curve (AUC), which is a common performance metric for deep learning algorithms, reaches 0.97, and the true positive rate is , while the false positive rate is . Since EAST was upgraded in 2020, the wall condition has been upgraded from non-all-metal to all-metal. To examine the robustness of the predictor, the EAST disruption predictor is migrated to the all-metal wall experiment for the first time. It is again tested under the all-metal wall experimental data, and its warning performance decreases significantly, with an AUC of only 0.79. To improve the robustness and sensitivity of the predictor against disruptions, the convolutional attention mechanism is introduced into the CNN. After training and testing with the same data set, the warning performance for the all-metal wall data is improved, with the AUC value increasing to 0.84. To further improve the robustness of the predictor, the t-distributed stochastic neighbor embedding algorithm is employed to explore the difference of the sample distribution before and after EAST upgradation, and the transfer learning algorithm is used to reduce this difference. By applying transfer learning to a small amount of discharges from the all-metal wall data, the warning performance in the all-metal wall experiment is further improved, and the AUC value increases to 0.93. Overall, the proposed data-driven predictor is promising for applications in future devices such as ITER.

ORNL Progress in Disruption Mitigation Technology in Support of ITER

The mitigation of plasma disruptions in tokamaks has become a very important topic in magnetic fusion research, motived by the potential challenges that may occur in ITER disruptions due to the high magnetic field and high plasma current. Such disruptions can have a deleterious effect on the internal components due to the fast dissipation of the plasma thermal energy and the magnetic stored energy leading to large forces, as well as the possible formation of several megaamperes of energetic runaway electrons during the current quench. Oak Ridge National Laboratory has been developing and deploying technology to inject material into the plasma to rapidly radiate the thermal energy and start a fast plasma current ramp down to dissipate the magnetic stored energy. The choice of materials to inject and the injection technology have evolved over the past decades to arrive at the present systems planned for ITER based on cryogenic pellets of hydrogen-neon mixtures for thermal mitigation and hydrogen pellets for runaway electron mitigation. This scheme injects shattered cryogenic material into the plasma from pellets formed in situ in a pipe gun and fired onto angled metal surfaces at the end of the injection line just before entering the plasma. In this paper, we describe the evolution of schemes and technologies that have been employed for disruption mitigation and runaway electron prevention and dissipation, discuss how they have performed in present-day experiments, and give the outlook for the use of this technology in a burning plasma and how it may continue to evolve in the future.

Damages of TZM as plasma facing material under transient heat load

Titanium-Zirconium-Molybdenum (TZM) is applied as the first wall material of EAST due to its excellent performance. To investigate the service behavior of TZM under transient heat fluxes, the postmortem inspection of in-situ damaged TZM tiles by plasma disruption heat load in EAST was analyzed. Meanwhile, the EMS-60 electron gun has been also used to test the transient heat load behaviors of TZM. There are some similar characteristics between the in-situ melting TZM and the ex-situ melting TZM tiles, but there are also some obvious differences. Specifically, in some melting zone, the surface morphology of partial in-situ TZM melting region is like that of ex-situ TZM melting areas, and their convex surface is relatively smooth, while the concave surface is relatively rough, and the solidification layer undergoes recrystallization. In other melting areas, there are many small pits on the surface of the TZM tile that was melted in-situ, but there are no obvious pits on the surface of the ex-situ TZM sample, and part of the surface of the TZM tile that was melted in-situ is like a big pit, and the melted TZM droplets in this melting area tend to splash towards the periphery. The capillary waves around the castle structure in the ex-situ test samples are the result of wetting effect of liquid metal and capillary force, and the holes in the resolidified layer is caused by the overheating boiling of TZM liquid. On the other hand, the number and propagation direction of cracks in the in-situ TZM tile are more complex than those in the ex-situ TZM sample. Firstly, the number of cracks on the in-situ TZM tile is large and dense, while the number of cracks on the surface of the ex-situ melting sample is small and scattered. The crack width of the in-situ tile is greater than that of the ex-situ tile. Secondly, the crack extension direction on the in-situ melted TZM is diverse, while the extension direction of the surface crack on the ex-situ TZM sample is relatively simple, and there is only one kind of crack on the tile surface. Such results provide reference for understanding the thermal behavior of TZM under transient heat flux in fusion devices.

Initial analytical theory of plasma disruption and experimental evidence

It is a great physical challenge to achieve controlled nuclear fusion in magnetic confinement tokamak and solve energy shortage problem for decades. In tokamak plasma, large-scale plasma instability called disruption will halt power production of reactor and damage key components. Prediction and prevention of plasma disruption is extremely urgent and important. However, there is no analytical theory can elucidate plasma disruption physical mechanism yet. Here we show an analytical theory of tokamak plasma disruption based on nonextensive geodesic acoustic mode theory, which can give the physical mechanism of disruption. The proposed theory has not only been confirmed by experimental data of disruption on T-10 device, but also can explain many related phenomena around plasma disruption, filling the gap in physical mechanism of tokamak plasma disruption.

Structural behaviour characterization of ST40 Inner Vacuum Chamber (IVC2) during a plasma VDE using ANSYS Workbench

The paper aims to define the structural behavior of the ST40 Inner Vacuum Chamber (IVC2) under the action of electromagnetic loads from a previously analysed plasma disruption scenario, namely an Upward Vertical Displacement Event (UVDE) scenario. This spherical tokamak is owned by Tokamak Energy Ltd, and it is the highest field device of its kind. The considered loads are obtained from a previous electromagnetic analysis [1] and interpolated on the structural model. Indeed, the study represents the second and final part of the IVC2 electromagnetic and structural behavior characterization. The maximum load due to eddy current and the maximum load generated by halo current (occurring in two different time steps) are studied in two separated analyses. These were carried out considering the entire ST40 model with high degree of detail, allowing to accurately approximate its real behavior and to extrapolate the stress and deformation conditions on all the device elements. Specifically, the completeness of the model has guaranteed to study the global and the local stress state generated on the IVC2, on its supports and on all the components that weigh on it.

Real-time control applications of JET heterodyne radiometer electron temperature profiles using ICRH

The Real Time ECE data processing has been recently implemented at JET, by using a cross calibration between the Martin-Puplett interferometer (also known as Michelson) measurements and the Heterodyne faster data. The applicability of RT control with ICRH has been discussed in recent works. In particular, in case of a hollow electron temperature profile, central ICRF heating could be applied to restore temperature peaking. In this work a study about the possibility of controlling the electron temperature profile by using ICRHas an actuator, with inputs from ECE measurements, is described. The ICRH is here used for the optimization of the plasma ramp-down, in order to correct the end of the discharge and avoid the plasma disruption, though a similar approach could also be proposed to improve the plasma performances during the flat top. Preliminary results have been obtained on the physical and logical feasibility of using ICRH in a real-time control loop. A database of good pulses in identifiable conditions has been compiled and validated by covering the main experiments both in the baseline and hybrid scenarios. Such scenarios have been simulated, for a selected numbers of discharges, using the full-wave propagation solver TORIC-SSQLFP, in order to establish the ICRH power deposition profiles to be used to locally restore a peaked electron temperature profile.

Progress of a gas-electron-multiplier based camera (GEM) in investigating EAST macroscopic instabilities

In recent years magneto-hydrodynamic instabilities have been observed in different experimental scenarios by a gas electron multiplier (GEM) camera, since it was installed on the Experimental Advanced Superconducting Tokamak (EAST). The GEM camera is set up outside the EAST vacuum chamber to record x-ray photons through a beryllium window, which has obtained valuable data for analysis: (1) neoclassical tearing mode (NTM) onset and (2) plasma disruption dynamics in the presence of magnetic islands. Although not precisely compared with theory and simulations, the GEM, in synergy with other diagnostics, reveals the significance of electron dynamics in NTM and plasma disruption. This is meaningful for understanding and modeling the disruption instability of fusion plasmas.

Simulation and developments for large pellet formation and acceleration for shattered pellet injection of the ITER DMS

Amongst the technological developments needed for ITER, the disruption mitigation system (DMS) is one of the most challenging since it has to mitigate the consequences of high-energy plasma disruptions to ensure the lifetime of ITER's structure and in-vessel components. ITER DMS is using Shattered Pellet Injection, whose technology needs to be developed for ITER specific requirements.In the framework of the “Technology Fundamental Studies for DMS” contract with ITER Organization, as part of the ITER DMS Task Force activities, and CEA, the DSBT (Low Temperature Systems Department) designed and operates a pellet injection test bench to produce and accelerate large cryogenic pellets with diameters from 10 to 28.5 mm and length/diameter ratios up to two. The objective of this activity is to study and to optimise the pellet formation and acceleration process using pure Protium, Deuterium, Neon and mixtures of these gases.Analytic and numerical studies were used to define the geometrical and cryogenic parameters for the pellet desublimation in the in-situ condensation cell of the cryostat. An analytical model has also been developed for the pellet acceleration in the single stage gas gun to validate the propellant valve design aiming at achieving velocities of up to 500 m/s for 28.5 mm protium pellets and 200 m/s for neon pellets.The complete test bench is composed of the gas propulsion system, the cryostat with its key component the pellet formation cell, the flight tube and the target chamber with viewing ports. The mechanical assembly is completed with specific instrumentation and fast data acquisition systems for the observation of the pellet formation and acceleration. The pellet integrity and trajectory are characterised by shadowgraphy with high resolution, high-speed cameras located at 1 m and 3 m after the barrel exit and with an accelerometer and a target foil in the target chamber.This document details the simulation studies, the selected test bench design and first obtained results.

Impurity Behavior in Plasma Recovery after a Vacuum Failure in the Experimental Advanced Superconducting Tokamak

After a vacuum failure in a tokamak, plasma runaway or plasma disruptions frequently occur during plasma recovery, causing difficulties in rebuilding a well-confined collisional plasma. In this work, the impurity behavior during plasma recovery after a vacuum failure in the 2019 spring campaign of the Experimental Advanced Superconducting Tokamak (EAST) was studied by analyzing the spectra recorded by fast-time-response extreme ultraviolet (EUV) spectrometers with 5 ms/frame. During the plasma current ramp-up in recovery discharges, a high content of the low-Z impurities of oxygen and carbon was found, i.e., dozens of times higher than that of normal discharges, which may have caused the subsequent runaway discharges. The electron temperature in the recovery discharge may have dropped to less than 75 eV when the collisional plasma quenched to the runaway status, based on the observable impurity ions in the two cases. Therefore, the lifetime of collisional plasma in the recovery discharge, τc, was deduced from the lifetime of H- and He-like oxygen and carbon ions identified from EUV spectra. It was found that, after several discharges with real-time lithium granule injection, the runaway electron flux and O+ influx reduced to 45% and 20%, respectively. Meanwhile, the lifetime of confined plasma was extended from 113 ms to 588 ms, indicating the effective suppression of impurities and runaway electrons and improvement in plasma performance by real-time lithium granule injection. The results in this work provide valuable references for the achievement of first plasma in future superconducting fusion devices such as ITER and CFETR.

MHD instability dynamics and turbulence enhancement towards the plasma disruption at the HL-2A tokamak

The evolutions of MHD instability behaviors and enhancement of both electrostatic and electromagnetic turbulence towards the plasma disruption have been clearly observed in the HL-2A plasmas. Two types of plasma disruptive discharges have been investigated for similar equilibrium parameters: one with a distinct stage of a small central temperature collapse (sim 5–10%) around 1 millisecond before the thermal quench (TQ), while the other without. For both types, the TQ phase is preceded by a rotating 2/1 tearing mode, and it is the development of the cold bubble from the inner region of the 2/1 island O-point along with its inward convection that causes the massive energy loss. In addition, the micro-scale turbulence, including magnetic fluctuations and density fluctuations, increases before the small collapse, and more significantly towards the TQ. Also, temperature fluctuations measured by electron cyclotron emission imaging enhances dramatically at the reconnection site and expand into the island when approaching the small collapse and TQ, and the expansion is more significant close to the TQ. The observed turbulence enhancement near the X-point cannot be fully interpreted by the linear stability analysis by GENE. Evidences suggest that nonlinear effects, such as the reduction of local E_rtimes B shear and turbulence spreading, may play an important role in governing turbulence enhancement and expansion. These results imply that the turbulence and its interaction with the island facilitate the stochasticity of the magnetic flux and formation of the cold bubble, and hence, the plasma disruption.

Estimation of the power dissipation on the casing of DTT toroidal field coil during a fast plasma disruption

The aim of this paper is the computation of the power dissipated on the Toroidal Field Coil (TFC) casing of the Divertor Tokamak Test facility (DTT). The DTT Tokamak will be installed in Fascati (Italy), and is in an advanced design stage. The magnetic system will be realized with superconducting coils so that the correct dimensioning of the cooling system is mandatory. As a consequence, a detailed knowledge of the main sources of dissipation is required. Our attention will here be focused on the TFC casing and, in particular, on the dissipations due to the eddy currents induced on the casing by time-varying magnetic field during a Fast Plasma Disruption (FPD). A finite element model whose geometry has partly been imported into Ansys 2021/R2 SpaceClaim (TFC casing), and partly made in Ansys 2021/R2 DesignModeler (plasma, Vacuum Vessel (VV) and Stabilizing Plates (SP)), has been implemented in Ansys 2021/R2 Maxwell 3D to perform a transient analysis of the FPD event. The current induced on the TFC casing, as well as the total power losses, are the output of our computation to be used for the cooling system dimensioning. In order to estimate the heat rate to be removed during normal operation, the breakdown and Single Null (SN) scenarios have also been studied.

DECAF perks up plasma forecasting

Physics-based approach accurately predicts plasma disruption events in tokamaks producing fusion energy