COMPASS Tokamak Research Articles

Correlation between non-ambipolar currents and divertor heat loads in the COMPASS tokamak

Abstract Electric current flowing onto the divertor of the COMPASS tokamak influences its heat loading. The current measured when a Langmuir probe is grounded to the divertor gives a local measurement of this heat loading, which, according to the classical theory of the Debye sheath, should be enhanced with respect to the case of locally ambipolar currents. The comparison of the calculated heat flux by probes with infrared thermography, when the influence of non-ambipolar currents is not considered, is grossly wrong; when the theoretical effect of non-ambipolar currents is included, however, the agreement is very good.

Quantification of locked mode instability triggered by a change in confinement

Abstract This work presents the first analysis of the disruptive locked mode (LM) triggered by the dynamics of a confinement change. It shows that, under certain conditions, the LM threshold during the transient is significantly lower than expected from steady states. We investigate the sensitivity to a controlled n = 1 error field (EF) activated prior to the L-H transition in the COMPASS tokamak, at q 95 ∼ 3, β N ∼ 1, and using EF coils on the high-field side of the vessel. A threshold for EF penetration subsequent to the L-H transition is identified, which shows no significant trend with density or applied torque, and is an apparent consequence of the reduced intrinsic rotation of the 2/1 mode during this transient phase. This finding challenges the assumption made in theoretical and empirical works that natural mode rotation can be predicted by global plasma parameters and urges against using any parametric EF penetration scaling derived from steady-state experiments to define the EF correction strategy in the entire discharge. Furthermore, even at EFs below the identified penetration threshold, disruptive locking of sawtooth-seeded 2/1 tearing modes is observed after about 30% of L-H transitions without external torque.

Mutual interactions between plasma filaments in a tokamak evidenced by fast imaging and machine learning.

Magnetically confined fusion plasmas are subject to various instabilities that cause turbulent transport of particles and heat across the magnetic field. In the edge plasma region, this transport takes the form of long filaments stretched along the magnetic field lines. Understanding the dynamics of these filaments, referred to as blobs, is crucial for predicting and controlling their impact on reactor performance. To achieve this, highly resolved passive fast camera measurements have been conducted on the COMPASS tokamak. These measurements are analyzed using both conventional tracking methods and a custom-developed machine-learning approach designed to characterize more particularly the mutual interactions between filaments. Our findings demonstrate that up to 18% of blobs exhibit mutual interactions in the investigated area close to the separatrix, at the border between confined and nonconfined plasma. Notably, we present direct observations and radial dependence of blob coalescence and splitting, rapid reversals in the propagation direction of the blob, as well as their dependence on the radial position. The comparison between observations realized with passive and gas puff imaging does not evidence any significant bias due to the use of the latter technique.

Experimental evidence of very short power decay lengths in H-mode discharges in the COMPASS tokamak

Analysis of heat fluxes in scrape-off layer (SOL) plasma is important for the prediction of divertor tile heat loads in future reactor-sized tokamak machines. Typically, the radial heat flux profile can be accurately characterized by the decay length parameter λ q . The predictions are then based on the dependence of λ q on plasma and machine parameters. In recent years, several empirical scaling models were derived from a collective database of decay lengths observed in several large tokamaks as well as two spherical tokamaks. Most recently, a report from the TCV tokamak showed a deviation from several of the proposed models (Maurizio et al (TCV team) 2021 Nucl. Fusion 61 024003). In this work, edge plasma profiles of ELMy H-mode discharges were collected from the COMPASS tokamak database for heat flux analysis. The COMPASS tokamak has a similar machine size and plasma parameters to the TCV tokamak, offering a valuable comparison. The λ q was measured in both downstream and upstream SOL utilizing a divertor probe array validated by an IR camera and Thomson scattering diagnostics. A comparison with the predictions of several scaling models indicated an occurrence of anomalously short (by a factor of 2–5) inter-ELM power decay lengths in both upstream and downstream, as well as a difference in λ q between these two regions. Possible causes related to the accuracy of magnetic reconstruction were eliminated and physics-based sources of apparent compression effects were hypothesized as a motivation for future research.

Subdiffusive transport of runaway electrons in presence of small amplitude MHD perturbations in COMPASS

Radial transport of runaway electrons (REs) in tokamaks is affected by the presence of magnetic perturbations, either caused by internal magnetohydrodynamic instabilities or induced by external coils. The magnetic field configuration inside the plasma volume consists in general of intact magnetic surfaces alternated with magnetic islands and stochastic layers, which make the usual diffusive approach, based on the Rechester–Rosenbluth formula, inadequate to the study of transport. Here the fractional diffusion approach is employed to model RE transport in presence of intrinsic magnetic perturbations (magnetic islands) in the flat-top phase of RE-dedicated discharges on COMPASS tokamak. The character of RE transport is found to be subdiffusive. The degree of subdiffusion is evaluated by running simulations with the ORBIT code and a time-fractional diffusion equation is applied to calculate the time evolution of RE particle number. The results are compared with the observed RE losses, estimated from the time integrated neutron signal.

Depth profile CF LIBS analysis of the wall deposited layer in the COMPASS tokamak after LiSn testing campaign

Plasma-facing materials in tokamak nuclear reactors are subject to intense heat and radiation. The design of liquid metal (LM) divertor walls has several advantages over the solid counterparts. It is important to understand the migration and re-deposition of LMs such as lithium (Li), lead (Pb) or even tin (Sn). This work presents CF LIBS depth profile analysis of the deposited layer on Ni–Cr based screws situated at different places in the COMPASS tokamak vessel during the Capillary Porous System (CPS) LM (LiSn) divertor testing campaign. The experiment was carried out at low Ar pressure (100 mbar) using ns Nd:YAG laser operating at 532 nm. The spectra were recorded using echelle spectrometer (M5000, Andor Tech.) equipped with an iCCD camera (iStar DH743, Andor Tech.). The quantification was based on observed C I, B I, Li I and (weak) Sn I lines, in addition to trace elements (Ca, Na, K) and main matrix elements (Cr, Ni, Fe). The evaluation of electron temperature and density was based on Cr I-II, Ni-II and Fe I-II lines. The LM based content is estimated to be Li (0.3-0.81 at.%) and Sn (0.1-0.23 at.%) with Li/Sn at.% ratio ranging from 3 to 3.5 (wt.% ratio of 0.17-0.20). In practise, the limit of detection of Sn by CF LIBS for this type of Cr-Ni-Fe based matrix is observed to be ∼0.1 at.%.

ELM temperature in JET and COMPASS tokamak divertors

Analysis of the divertor edge localized mode (ELM) electron temperature at a uniquely high temporal resolution (10−5 s) was reported at the JET tokamak (Guillemaut et al 2018 Nucl. Fusion 58 066006). By collecting divertor probe data obtained during many dozens of ELMs, the conditional-average (CAV) technique yields surprisingly low peak electron temperatures, far below the pedestal ones (70%–99% reduction!) which we, however, question. This result was interpreted through the collisional free-streaming kinetic model of ELMs, by a transfer of most of the electron energy to ions, implying a high tungsten sputtering for unmitigated ELMs in future fusion devices like ITER. Recently, direct microsecond temperature measurements on the COMPASS tokamak, however, showed that the electron temperature peak of ELM filaments measured in the divertor is reduced by less than a third with respect to the pedestal one. This was further confirmed by a dedicated 1D particle-in-cell (PIC) simulation and tends to prove that the pedestal electrons can transfer only their parallel energy to ions (due to low collisionality), thus less than a third, as is predicted by the collisionless free-streaming model. This finding strongly contradicts the JET observations. We have therefore compared the CAV to the direct (microsecond) ball-pen and Langmuir probes measurements in COMPASS and found very good agreement between them. Revisiting the aforementioned JET CAV analysis indeed shows that the electron temperatures are much higher than previously reported, close to those predicted by the PIC simulation, and thus the ion energy seems to not significantly increase in the scrape-off layer.

Characterisation of the auxiliary heating power delivered by the 40 keV neutral beam injectors on the COMPASS tokamak

Two identical neutral beam injectors (NBIs) have been installed on the COMPASS tokamak. Each of them is capable of producing a deuterium (or hydrogen, helium) neutral beam with nominal power of up to 400 kW at 40 keV. Since the COMPASS tokamak is a compact machine, the design of the neutral beams was carefully optimized to fit the neutral beam through the narrow ports of the tokamak. However, once the neutral beams are operated out of the scope of their optimal parameters, additional power losses occur due to neutral beam scraping. This article investigates the amount of power loss due to neutral beam scraping. The losses are calculated using the neutral beam models and measured neutral beam parameters. The calculated results are compared with the measured power delivered into the plasma. Two different neutral beam models are employed. One describes the neutral beam as a sum of small beamlets, which are Gaussian-like. The second one describes the neutral beam as a single Gaussian beam. The amount of the power lost to scraping is determined by both models and compared. Neutral beam parameters, the input parameters for both models, are derived from the Doppler-shifted spectra emitted by fast neutral atoms. It is shown that when operating NBIs at nominal parameters, the scraping losses are negligible (5% ). If the NBIs are operated too far from the nominal parameters, the scraping losses can reach up to ∼60% of the NBI output power. As the highest fractional losses occur with low power NBI, the beam duct is capable of withstanding the corresponding power loads. The expected auxiliary heating power , the power entering the tokamak chamber carried by fast neutrals, is benchmarked against the additional plasma input power observed in the measured plasma parameters . The additional plasma input power is derived from the plasma power balance. The power balance terms are independent of the neutral beam model used for the auxiliary heating power calculation . The results of the power balance evaluation reveal a difference between two NBIs installed on COMPASS. One of the injectors systematically delivers 17% less power than the other into the plasma given the same beam parameters. It is concluded that the additional losses occur before the beam enters the tokamak chamber.

Current quench and vessel currents characterisation at the COMPASS tokamak

The characterisation of plasma current quench and understanding of its underlying physical processes play a crucial role when designing large fusion devices such as ITER. For the first time, an extensive analysis of the COMPASS tokamak disruption database is presented. A unique set of magnetic diagnostics allows the investigation of local toroidal and poloidal vessel currents, including currents flowing along the open magnetic field lines from the plasma to the vacuum vessel (VV) (i.e., halo currents). Area-normalised current quench times are in agreement with the ITPA 1.67 ms m−2 lower limit. Extremely fast I p quench rates of up to 0.6 MA ms−1 are observed during runaway electron campaigns at COMPASS, which are under the ITPA lower limit. Vertical movement of the plasma column is accelerated by the I p quench during major disruptions. Toroidal vessel currents of around 2% – 4% of the predisruptive plasma current are observed during I p quench. Net poloidal eddy currents are obtained by Mirnov coils and diamagnetic loop, reaching 3% of . Using the Mirnov coils it is shown that the halo current magnitude grows and its poloidal profile broadens with increasing plasma current I p . Geometric features of the VV structure and in-vessel component positions on the poloidal vessel current measurements are discussed.

Semi-supervised deep networks for plasma state identification

Correct and timely detection of plasma confinement regimes and edge localized modes (ELMs) is important for improving the operation of tokamaks. Existing machine learning approaches detect these regimes as a form of post-processing of experimental data. Moreover, they are typically trained on a large dataset of tens of labeled discharges, which may be costly to build. We investigate the ability of current machine learning approaches to detect the confinement regime and ELMs with the smallest possible delay after the latest measurement. We also demonstrate that including unlabeled data into the training process can improve the results in a situation where only a limited set of reliable labels is available. All training and validation is performed on data from the COMPASS tokamak. The InceptionTime architecture trained using a semi-supervised approach was found to be the most accurate method based on the set of tested variants. It is able to achieve good overall accuracy of the regime classification at the time instant of 100 µs delayed behind the latest data record. We also evaluate the capability of the model to correctly predict class transitions. While ELM occurrence can be detected with a tolerance smaller than 50 µs, detection of the confinement regime transition is more demanding and it was successful with 2 ms tolerance. Sensitivity studies to different values of model parameters are provided. We believe that the achieved accuracy is acceptable in practice and the method could be used in real-time operation.

Numerical modelling for beam duct heat loads calculations and application to the new 1 MW neutral beam injector in the COMPASS tokamak

We introduce detailed numerical modelling of the fast neutral particles inside the duct of a new neutral beam injector (NBI) recently installed at the COMPASS tokamak (major radius , vessel midplane minor radius , toroidal field = 0.9–2.1 ). This new NBI system is able to deliver power to the plasma at nominal injection energy of . Collisions with the background neutrals inside the beam duct give birth to fast ions according to the density of the gas and tabulated cross-sections. The ion trajectories are then computed in the complete 3D magnetic field, showing the importance of the stray magnetic field and the magnitude of the field within the gap in between toroidal coils.During the experimental campaign dedicated to the new 1 MW NBI, the beam duct heating was measured by a row of thermocouples located on the top-half of the beam duct. The fast ions collisions with the duct wall cause a local temperature increase with a characteristic pattern. In COMPASS, the location of the fast ions power deposition measured experimentally is in qualitative agreement with modelling of ion losses when following them after the re-ionization process. We trace back the details of the orbits corresponding to deposition at the hot-spot inside the beam duct. Quantitative comparison between the experiment and the simulation shows that a larger than expected amount of neutral gas was inside the duct and this study will guide the design of the future NBI duct in COMPASS Upgrade.

On the applicability of three and four parameter fits for analysis of swept embedded Langmuir probes in magnetised plasma

The problem of power exhaust is one of the grand challenges of nuclear fusion research today. In order to understand the physics phenomena occurring in the scrape-off layer and the divertor regions of tokamaks, it is essential to correctly determine the divertor plasma parameters, which are often measured by swept Langmuir probes (LPs). While the construction and operation of this diagnostic can be straightforward, the data analysis using three- or four-parameter fits presents a challenge and can potentially lead to erroneous values of electron temperature and ion saturation current. In this work, we present modelling and experiments aimed at determination of conditions for proper analysis of swept LPs using these two fitting models. Particle-in-cell modelling was employed to evaluate the sheath-expansion effects for particular probe geometry and plasma conditions, yielding a semi-empirical rule capable of predicting its magnitude. Experiments with unusually wide range of swept voltage in the divertor of the COMPASS tokamak explored the magnitude of voltage range required for successful analysis with either three or four-parameter fitting. With the use of our new semi-empirical rule, it is possible to improve the four-parameter fit reliability in situations where the available voltage range is limited. In addition, we introduce the tangent method—an independent and fast method of electron temperature estimation, which allows to reliably determine the available voltage range and as such assist more complex methods of probe analysis.

Current density limitation during disruptions due to plasma-sheaths

The presented experimental study realized in the COMPASS tokamak demonstrates, for the first time, that the current density that flows from the plasma into the vacuum vessel during disruptions is limited by the ion particle flux. Such a limitation shows that, at least in COMPASS, the sheath that forms between the plasma and the first wall dominates the halo current flow. This observation is achieved by measuring simultaneously the ion saturation current with negatively biased Langmuir probes and the halo current with grounded probes to the vacuum vessel. These comparative measurements, which were never performed during disruptions in other machines, directly confirm that the halo current density remains below the ion particle flux in COMPASS. The study also shows, using Mirnov coils measurement, that the total electric current entering the wall grows with the plasma current while the current density obtained by Langmuir probes remains unaffected. This, together with the current density limitation, leads to a novel finding that the halo current width increases with the pre-disruptive plasma current, which limits the local forces. The new findings reported here could also provide potential constraints on the modeling of disruption-induced loads on future reactor scale tokamaks and motivation for further experiments on existing devices.

3D deposition patterns of deuterium retention and impurities in the COMPASS divertor: a data-driven root cause analysis and prediction approach

Divertor tiles from the COMPASS tokamak have been examined with ion beam analysis for the determination of deuterium retention and co-deposits. A novel approach was used to draw 2D deuterium retention maps at different depths and correlated to the concentrations of other elements: boron, carbon, oxygen and metals. This approach, which employs all three spatial dimensions, revealed that the deuterium retention pattern is not toroidally symmetric at different depths while the overall deuterium pattern is toroidally symmetric. Analysis of visible-light camera records disclosed that parts of the divertor were not wetted by plasma. The radial profile of deuterium in both wetted and non-wetted (“shadowed”) divertor regions was compared to strike point positions, angle of incidence, electron temperature Te, parallel heat flux Qpar and ion saturation current Isat in order to find the influence of these parameters on deuterium retention. There are certain series of consecutive shots for which correlations between deuterium retention and separate plasma parameters (Te, Isat, Qpar) are high. Combining the three parameters in a linear model or a power law model enabled reconstruction of the radial deuterium retention pattern. After fitting the model to one third of the data points, the deuterium retention for shadowed tiles was reproduced at the other two thirds of the data points within 20% deviation. This study shows the need for detailed tile analysis in three dimensions, the great benefits of broad statistical analysis in solving challenges in fusion, and motivates the investigation of COMPASS tiles after dismantling.

Statistical properties of ion and electron temperature fluctuations in the edge of the COMPASS tokamak

This paper presents measurements of the plasma turbulence with an emphasis on the ion temperature fluctuations in comparison to the electron temperature and the ion saturation current fluctuations in the edge of the COMPASS tokamak. The ion temperature values are obtained by using a fast swept ball-pen probe, with temporal resolution up to 10 µs, accompanied by independent 1 µs measurements of the electron temperature and ion saturation current. The experimental results obtained for a wide range of line-averaged densities show that statistical properties of the fluctuations are in good agreement with a form of beta or gamma distribution function. In addition, the dependence of the averaged ion and electron temperatures on the line-averaged density is studied in the near and main scrape off layer region.

Detection of runaway electrons at the COMPASS tokamak using a Timepix3-based semiconductor detector

Runaway electrons are considered dangerous for the integrity of tokamak vacuum vessels. To secure the success of the future tokamak-based machines, reliable diagnostics and mitigation strategies are necessary. The COMPASS tokamak supported the research of runaway electron physics via regular experimental campaigns. During the last two experimental campaigns dedicated to runaway electrons, a semiconductor detector with a Timepix3 readout chip, Si sensor, and the SPIDR readout system was tested. Time evolution signals, energy measurements, and sensor snapshots collected with the Timepix3-based detector are presented.

Progress in HXR diagnostics at GOLEM and COMPASS tokamaks

Scintillation detectors are widely used for hard X-ray spectroscopy and allow us to investigate the dynamics of runaway electrons in tokamaks. This diagnostic tool proved to be able to provide information about the energy or the number of runaway electrons. Presently it has been used for runaway studies at the GOLEM and the COMPASS tokamaks. The set of scintillation detectors used at both tokamaks was significantly extended and improved. Besides NaI(Tl) (2 × 2 inch) scintillation detectors, YAP(Ce) and CeBr3 were employed. The data acquisition system was accordingly improved and the data from scintillation detectors is collected with appropriate sampling rate (≈300 MHz) and sufficient bandwidth (≈100 MHz) to allow a pulse analysis. Up to five detectors can currently simultaneously monitor hard X-ray radiation at the GOLEM. The same scintillation detectors were also installed during the runaway electron campaign at the COMPASS tokamak. The aim of this contribution is to report progress in diagnostics of HXR radiation induced by runaway electrons at the GOLEM and the COMPASS tokamaks. The data collected during the 12th runaway electron campaign (2020) at COMPASS shows that count rates during typical low-density runaway electron discharges are in a range of hundreds of kHz and detected photon energies go up to 10 MeV (measured outside the tokamak hall). Acquired data from experimental campaigns from both machines will be discussed.

Calibration-free laser-based spectroscopic study of Sn-based alloys

The elemental quantification of liquid metal divertor (LMD) surface is important for understanding the material erosion, migration, re-deposition, and fuel retention in Plasma-Facing Components (PFCs). Currently, LMD are attractive candidates for the short- and long-term operation of fusion devices like DEMO. Liquid metals can provide self-cooling, self-replenishing plasma-facing surfaces requiring very little upkeep. In a previous work, we studied Li and LiSn layers deposited on attachment screws in the COMPASS tokamak by means of Calibration-Free Laser-Induced Breakdown Spectroscopy (CF-LIBS). Several problems were encountered related to the detection of Sn in LiSn. Thus, in the present work, we are optimizing the experimental conditions for the detection of Sn I-II and Pb I-II in Pb-containing Sn-based alloys, performing the quantification of Pb in traces and in bulk quantities using CF-LIBS approach.

Predictive modelling of liquid metal divertor: from COMPASS tokamak towards Upgrade

Following ELMy H-mode experiments with liquid metal divertor target on the COMPASS tokamak, we predict the behavior of a similar target on COMPASS Upgrade, where it will be exposed to surface heat fluxes even higher than those expected in the future EU DEMO attached divertor. We simulate the heat conduction, sputtering, evaporation, excitation and radiation of lithium and tin in the divertor area. Measured high-resolution data from COMPASS tokamak were rescaled towards the Upgrade based on many established scalings. Our simulation then yields the amount of released metal which ranges from 4 mg s−1 upto 12 g s−1 depending mainly on the geometry and Li/Sn choice, quite independently from active cooling or strike point sweeping.

Conceptual design of Thomson scattering diagnostics for the COMPASS-U tokamak.

The Thomson scattering (TS) diagnostic, one of the key diagnostics used on the tokamaks around the world, is planned for the COMPASS-U tokamak, which is recently under design and construction in the Institute of Plasma Physics in Prague, Czech Republic. This tokamak is supposed to be a world-unique, high magnetic field device with hot walls, allowing for the study of the plasma exhaust in advanced operational scenarios and testing cutting-edge technologies relevant to future fusion reactors, e.g., use of liquid metals. The core and edge TS systems are planned to be designed and operational, with a limited performance, already in the early stage of the tokamak operation. In this contribution, requirements and the most important constraints defining the TS system design are presented. The impact of both the possible collection lens location and spatial resolution on the plasma pedestal observation is simulated. Design considerations also take into account the high-resolution TS core and edge systems available from the COMPASS tokamak, which will be reused. The collection lenses will be newly built. Extension of the detection system will complete the plasma radius coverage in the future. The divertor TS is considered for later periods.