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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Demonstration of Tokamak Discharge Shutdown with Shell Pellet Payload Impurity Dispersal

The first rapid tokamak discharge shutdown using dispersive core payload deposition with shell pellets has been achieved in the DIII-D tokamak. Shell pellets are being investigated as a possible new path toward achieving tokamak disruption mitigation with both low conducted wall heat loads and slow current quench. Conventional disruption mitigation injects radiating impurities into the outer edge of the tokamak plasma, which tends to result in poor impurity assimilation and creates a strong edge cooling and outward heat flow, thus requiring undesirable high-$Z$ impurities to achieve low conducted heat loads. The shell pellet technique aims to produce a hollow temperature profile by using a thin, low-ablation shell surrounding a dispersive payload, giving a greatly increased impurity ablation (and radiation) rate when the payload is released in the plasma core. This principle was demonstrated successfully using 3.6 mm outer diameter, 40 μ m thickness diamond shells holding boron powder. The pellets caused rapid (<10 ms) discharge shutdown with low conducted divertor heat fluence (~0.1 MJ/m 2 ). Confirmation of massive release of the boron powder payload into the plasma core was obtained spectroscopically. Some evidence for the formation of a hollow temperature profile during the shutdown was observed. Furthermore, these first results open a new avenue for disruption mitigation research, hopefully enabling development of highly effective methods of avoiding disruption wall damage in future reactor-scale tokamaks.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Not one more US tokamak

The tokamak has been the primary magnetic fusion confinement platform in the USA since the Model C Stellarator was converted to the Symmetric Tokamak in 1970. The tokamak concept holds records for the best fusion performance parameters of any MFE concept, to-date. Presently, the only mid-scale operating US DOE magnetic fusion confinement facility, is the DIII-D tokamak at General Atomics in San Diego (at least until NSTX-U repairs are completed). But it has been more than thirty years since DIII-D was first put into operation. There is not a single new next-generation DOE-funded mid-scale (or larger) magnetic fusion confinement facility planned in the USA for the next 10 years. When and if there is a next generation new US domestic confinement machine funded by DOE, it should not be a tokamak.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Homoclinic tangles in the DIII-D divertor tokamak

A new approach was developed for the calculation of homoclinic tangles of the primary unperturbed separatrix in divertor tokamaks. This new approach is different than calculating tangles by following single trajectories in phase space [M. V. Berry, Topics in Nonlinear Dynamics (Ed. S. Jorna), Am. Inst. Phys. Conf. Proc. 46, 16-120 (1978)] or calculating the trajectories asymptotically in canonical time [R. K. W. Roeder, B. I. Rapoport, and T. E. Evans, Phys. Plasmas 10, 3796 (2003)]. The approach is to consider an equilibrium manifold as a whole and to calculate how the entire manifold evolves in phase space in time under the influence of perturbations. When the forward moving and the backward moving manifolds meet in a poloidal plane and intersect, they form tangles to preserve the topological invariance of the Hamiltonian system. This approach was applied to the DIII-D tokamak. The low mn, the high mn, the peeling-ballooning modes, and topological noise and field error perturbations were considered. The results of the peeling-ballooning modes were analyzed. It was found that the perturbations create very complicated structures in phase space [A. Punjabi and A. Boozer, Phys. Lett A 378, 2410-2416 (2014); A. Punjabi, Phys. Plasmas 21, 122516 (2014)]. The structure of tangles gives a possible cause for generation of radial electric fields in divertor tokamaks, and the possible consequent drifts, currents, shears, and curvature effects from the resulting complicated electric fields can possibly indicate a path to L to H mode transition [F. Wagner F et al, Phys. Rev. Lett. 49, 1408 (1982)] in divertor tokamaks. During the present period, four (4) refereed paper were published; six (6) contributed papers were presented at the EPS Conference on Plasma Physics and published in the EPS conference proceedings; six (6) contributed poster papers, co-authored with HU undergraduate students, were presented at APS/DPP Annual Meetings; three contributed papers were presented at HU School of Science Annual Research Symposia; four (4) invited talks were given; one talk, co-authored by Dr. Allen Boozer and the PI, was given by Dr. Boozer at the 20th International Stellarator-Heliotron Workshop 2015, in Greifswald, Germany; and four (4) graduate students and nine (9) undergraduate students were mentored and provided research experience in fusion science. For the renewal period, it is planned to work on four (4) new problems. These are: Homoclinic tangles of the separatrices of the DIII-D and the simple map, magnetic topology and homoclinic tangles of separatrices in divertor tokamaks, resonant stellarator divertors, and nonresonant stellarator divertors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Tokamak Disruption Simulation (Final Technical Report)

The Tokamak Disruption Simulation collaboration was a SciDAC (Scientific Discovery through Advanced Computing) program led by Xianzhu Tang of the Los Alamos National Laboratory but with separate grants to each institution. I was the PI for Columbia University and the only person at Columbia directly funded by this SciDAC. Disruptions are a sudden loss of confinement in tokamaks, and a reliable method of preventing disruptions must be developed before tokamak fusion power plants become feasible. A related plasma confinement concept, the stellarator, is the only way known way of preventing disruptions with the required reliability. The largest devices, both existing and under construction, for confining plasmas for fusion using magnetic fields are tokamaks, so it is important to solve their disruption issues. The sudden loss of confinement in tokamaks can be from a bit of material falling into the plasma, which causes the plasma energy to be quickly lost by radiation, or by an instability causing a breakup of the confining surfaces formed by the magnetic field. This breakup can occur more than a million times faster than one would naively think possible. The sudden breakup of surfaces is called a fast magnetic reconnection and also results in a sudden cooling of the plasma.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

UCLA Tokamak Program Close Out Report

The results of UCLA experimental fusion program are summarized. Starting with smaller devices like Microtor, Macrotor, CCT and ending the research on the large (5 m) Electric Tokamak. CCT was the most diagnosed device for H-mode like physics and the effects of rotation induced radial fields. ICRF heating was also studied but plasma heating of University Type Tokamaks did not produce useful results due to plasma edge disturbances of the antennae. The Electric Tokamak produced better confinement in the seconds range. However, it presented very good particle confinement due to an "electric particle pinch". This effect prevented us from reaching a quasi steady state. This particle accumulation effect was numerically explained by Shaing's enhanced neoclassical theory. The PI believes that ITER will have a good energy confinement time but deleteriously large particle confinement time and it will disrupt on particle pinching at nominal average densities. The US fusion research program did not study particle transport effects due to its undue focus on the physics of energy confinement time. Energy confinement time is not an issue for energy producing tokamaks. Controlling the ash flow will be very expensive.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

TRANSP integrated modeling code for interpretive and predictive analysis of tokamak plasmas

This paper provides a comprehensive review of the TRANSP code, a sophisticated tool for interpretive and predictive analysis of tokamak plasmas, detailing its major capabilities and features. It describes the equations for particle, power, and momentum balance analysis, as well as the poloidal field diffusion equations. The paper outlines the spatial and time grids used in TRANSP and details the equilibrium assumptions and solvers. Various models for heating and current drive and radiation, including updates to the NUBEAM model, are discussed. The handling of large-scale events such as sawtooth crashes and pellet injections is examined, along with the predictive capabilities for advancing plasma profiles. The integration of TRANSP with the ITER Integrated Modeling and Analysis Suite (IMAS) is highlighted, demonstrating enhanced data access and analysis capabilities. Additionally, the paper discusses best practices and continuous integration techniques to enhance TRANSP's robustness. The suite of TRANSP tools, designed for efficient data analysis and simulation, further supports the optimization of tokamak operations and coupling with other tokamak codes. Continuous development and support ensure that TRANSP remains a major code for the analysis of experimental data for controlled thermonuclear fusion, maintaining its critical role in supporting the optimization of tokamak operations and advancing fusion research.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

Comparing linear ion-temperature-gradient-driven mode stability of the National Compact Stellarator Experiment and a shaped tokamak

One metric for comparing confinement properties of different magnetic fusion energy configurations is the linear critical gradient of drift wave modes. The critical gradient scale length determines the ratio of the core to pedestal temperature when a plasma is limited to marginal stability in the plasma core. The gyrokinetic turbulence code GS2 was used to calculate critical temperature gradients for the linear, collisionless ion temperature gradient (ITG) mode in the National Compact Stellarator Experiment (NCSX) and a prototypical shaped tokamak, based on the profiles of a JET H-mode shot and the stronger shaping of ARIES-AT. While a concern was that the narrow cross section of NCSX at some toroidal locations would result in steep gradients that drive instabilities more easily, it is found that other stabilizing effects of the stellarator configuration offset this so that the normalized critical gradients for NCSX are competitive with or even better than for the tokamak. For the adiabatic ITG mode, NCSX and the tokamak had similar adiabatic ITG mode critical gradients, although beyond marginal stability, NCSX had larger growth rates. However, for the kinetic ITG mode, NCSX had a higher critical gradient and lower growth rates until a/L{sub T} Almost-Equal-To 1.5 a/L{sub T,crit}, when it surpassed the tokamak's. A discussion of the results presented with respect to a/L{sub T} vs. R/L{sub T} is included.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Ion and electron cyclotron wall conditioning in stellarator and tokamak magnetic field configuration on WEGA

Discharge wall conditioning is an effective tool to improve plasma performance in tokamaks and stellarators. RF Discharge Conditioning (RFDC) techniques are envisaged for use during operational campaigns on superconducting devices like the ITER tokamak and W7-X stellarator, as alternative to DC Glow Discharge Conditioning which is inefficient in presence of magnetic fields. This contribution investigates RFDC in both the ion and electron cyclotron range of frequencies (ICRF and ECRF) on the WEGA device (Max-Planck-Institute for Plasma Physics, Greifswald, Germany) as preparation for W7-X operation. ECRF discharges produced by localised absorption of RF power at EC resonance layers suffer from poor radial discharge homogeneity in the tokamak vacuum magnetic field configuration, severely limiting the plasma wetted wall areas and consequently the conditioning efficiency. The non-localised production of ICRF discharges by collisional RF power absorption features much improved discharge homogeneity making Ion Cyclotron Wall Conditioning (ICWC) the favoured RFDC technique for superconducting tokamaks. RFDC with the stellarator vacuum magnetic field needs to aim at sufficient plasma densities at and outside the last closed flux surface (LCFS), maximising the convective plasma flux along the open field lines to the wall. Whereas for ICRF discharges this condition is easily fulfilled, on WEGA for He-ECRF discharges this could be achieved as well by off axis heating close to the LCFS. In stellarator magnetic field configuration it is found that He-ICWC for wall desaturation is at least one order of magnitude more efficient than He-ECWC. Novel ECWC methods are proposed that can decrease this efficiency gap with ICWC to a factor 2-3. The efficiency difference is less pronounced in case of H{sub 2}-ICWC and ECWC for isotopic exchange.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2014

The residual zonal flow in tokamak plasmas toroidally rotating at arbitrary velocity

Zonal flows, initially driven by ion-temperature-gradient turbulence, may evolve due to the neoclassic polarization in a collisionless tokamak plasma. In our previous work [D. Zhou, Nucl. Fusion 54, 042002 (2014)], the residual zonal flow in a tokamak plasma rotating toroidally at sonic speed is found to have the same form as that of a static plasma. In the present work, the form of the residual zonal flow is presented for tokamak plasmas rotating toroidally at arbitrary velocity. The gyro-kinetic equation is analytically solved for low speed rotation to give the expression of residual zonal flows, and the expression is then generalized for cases with arbitrary rotating velocity through interpolation. The zonal flow level decreases as the rotating velocity increases. The numerical evaluation is in good agreement with the former simulation result for high aspect ratio tokamaks.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Global gyrokinetic stability of collisionless microtearing modes in large aspect ratio tokamaks

Linear full radius gyrokinetic calculations show the existence of unstable microtearing modes (MTMs) in purely collisionless, high temperature, large aspect ratio tokamak plasmas. The present study takes into account fully gyrokinetic highly passing ions and electrons. The global 2-D structures of the collisionless mode with full radius coupling of the poloidal modes is obtained and compared with another electromagnetic mode, namely, the Alfvén Ion Temperature Gradient (AITG) mode (or Kinetic Ballooning Mode, KBM) for the same equilibrium profile. Several important characteristics of the modes are brought out and compared, such as a clear signature in the symmetry properties of the two modes, the plasma–β dependence, and radial and poloidal length scales of the electrostatic and magnetic vector potential fluctuations. Extensive parameter scans for this collisionless microtearing mode reveal the scaling of the growth rate with β and the electron temperature gradient η{sub e}. Scans at different β values show an inverse relationship between the η{sub e} threshold and β, leading to a stability diagram, and implying that the mode might exist at moderate to strong temperature gradients for finite β plasmas in large aspect ratio tokamaks. In contrast to small aspect ratio tokamaks where the trapped electron magnetic drift resonance is found to be important, in large aspect ratio tokamaks, a strong destabilization due to the magnetic drift resonance of passing electrons is observed and is identified as a possible collisionless drive mechanism for the collisionless MTM.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Sensitivity of magnetic field-line pitch angle measurements to sawtooth events in tokamaks

The sensitivity of the pitch angle profiles measured by the motional Stark effect (MSE) diagnostic to the evolution of the safety factor, q, profiles during the tokamak sawtooth events has been investigated for Korea Superconducting Tokamak Advanced Research (KSTAR). An analytic relation between the tokamak pitch angle, γ, and q estimates that Δγ ∼ 0.1° is required for detecting Δq ∼ 0.05 near the magnetic axis (not at the magnetic axis, though). The pitch angle becomes less sensitive to the same Δq for the middle and outer regions of the plasma (Δγ ∼ 0.5°). At the magnetic axis, it is not straightforward to directly relate the γ sensitivity to Δq since the gradient of γ(R), where R is the major radius of the tokamak, is involved. Many of the MSE data obtained from the 2015 KSTAR campaign, when calibrated carefully, can meet these requirements with the time integration down to 10 ms. The analysis with the measured data shows that the pitch angle profiles and their gradients near the magnetic axis can resolve the change of the q profiles including the central safety factor, q{sub 0}, during the sawtooth events.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

X-point effects on the ideal MHD modes in tokamaks in the description of dual-poloidal-region safety factor

The flux coordinates with dual-region safety factor (q) in the poloidal direction are developed in this work. The X-point effects on the ideal MHD modes in tokamaks are then analyzed using this coordinate system. Since the X-point effects mainly affect the edge region, the modes localized at the tokamak edge are particularly examined. Two types of modes are studied. The first is related to the conventional peeling or peeling-ballooning modes. The mode existence aligned with the local magnetic field in the poloidally core region as observed experimentally is confirmed. The X points are shown to contribute to a stabilizing effect for the conventionally treated modes with the surface-averaged q and with the tokamak edge portion truncated. The other is the axisymmetric modes localized in the vicinity of X points, which can affect the cross-field-line transport near the X points. The existence of axisymmetric modes points to the possibility of applying a toroidally axisymmetric resonant magnetic perturbation (RMP) in the X-point area for mitigating the edge localized modes, which can be an alternative to the current RMP design. The dual q description also has important implications for the existing non-axisymmetric RMP concept. It helps to understand why the RMP suppression of edge localized modes is difficult to achieve in the double-null tokamak configurations and points to the possibility of further improving the current RMP concept by considering the alignment to the local q.

Fourier analysis↗
DOE OSTI2019

MCNP Models for Tokamak Cooling Water System Equipment Evaluations

Heat generated in the ITER fusion reactor is deposited in the tokamak vacuum vessel, in-vessel components, and in the components of the neutral beam injector during plasma operations and during subsequent decay of activation products. This heat is managed by the tokamak cooling water system (TCWS). The stainless steel material in the integrated loop of blanket edge-localized mode vertical stabilization coils and divertor (IBED) components (e.g., piping, heat exchangers (HXs), and pumps) contains activation sources because of its exposure primarily to neutron radiation from the decay of 17 N, which is a short-lived radionuclide produced by neutron capture reactions with oxygen nuclei in the IBED primary heat transfer system (PHTS) cooling water during plasma operations. A detailed geometry model of the IBED stainless steel components and neutron radiation sources is required for an accurate assessment of the gamma activation sources on level 3 of the tokamak building. In the baseline design, each of the eight IBED PHTS cooling trains has two shell-and-tube heat exchangers (HXs) connected in series. Because these HXs are very large and contain a large amount of radioactive water, the possibility of using compact HXs of the welded shell-and-plate type is under investigation. This paper presents two Monte Carlo N-Particle (MCNP) TCWS geometry models, one model for each HX type, along with the associated piping. These models were obtained by automatic geometry conversion from TCWS computer-aided design models. In conclusion, the TCWS geometry models and neutron source definitions were incorporated into a baseline MCNP model of the Tokamak Complex.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Energetic particles in spherical tokamak plasmas

Spherical tokamaks (STs) typically have lower magnetic fields than conventional tokamaks, but similar mass densities. Suprathermal ions with relatively modest energies, in particular beam-injected ions, consequently have speeds close to or exceeding the Alfvén velocity, and can therefore excite a range of Alfvénic instabilities which could be driven by (and affect the behaviour of) fusion α-particles in a burning plasma. STs heated with neutral beams, including the small tight aspect ratio tokamak (START), the mega amp spherical tokamak (MAST), the national spherical torus experiment (NSTX) and Globus-M, have thus provided an opportunity to study toroidal Alfvén eigenmodes (TAEs), together with higher frequency global Alfvén eigenmodes (GAEs) and compressional Alfvén eigenmodes (CAEs), which could affect beam current drive and channel fast ion energy into bulk ions in future devices. In NSTX GAEs were correlated with a degradation of core electron energy confinement. In MAST pulses with reduced magnetic field, CAEs were excited across a wide range of frequencies, extending to the ion cyclotron range, but were suppressed when hydrogen was introduced to the deuterium plasma, apparently due to mode conversion at ion–ion hybrid resonances. At lower frequencies fishbone instabilities caused fast particle redistribution in some MAST and NSTX pulses, but this could be avoided by moving the neutral beam line away from the magnetic axis or by operating the plasma at either high density or elevated safety factor. Fast ion redistribution has been observed during GAE avalanches on NSTX, while in both NSTX and MAST fast ions were transported by saturated kink modes, sawtooth crashes, resonant magnetic perturbations and TAEs. The energy dependence of fast ion redistribution due to both sawteeth and TAEs has been studied in Globus-M. High energy charged fusion products are unconfined in present-day STs, but have been shown in MAST to provide a useful diagnostic of beam ion behaviour, supplementing the information provided by neutron detectors. In MAST electrons were accelerated to highly suprathermal energies as a result of edge localised modes, while in both MAST and NSTX ions were accelerated due to internal reconnection events. Lastly, ion acceleration has also been observed during merging-compression start-up in MAST.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Density limits as disruption forecasters for spherical tokamaks

Fusion power output from spherical tokamaks would benefit from increased confined plasma density, but there exists a limit on the density before confinement is lost and the plasma current is disrupted. This density limit has long been characterized by a simple, global Greenwald limit proportional to the plasma current and inversely proportional to the cross sectional area of the plasma. It is shown that in the database of discharges from the National Spherical Tokamak Experiment (NSTX) and Mega Ampere Spherical Tokamak (MAST) spherical tokamaks, the likelihood of disruption does increase above the Greenwald limit, and especially in the plasma current rampdown phase. The physics of the density limit has been recently theoretically explored through local criteria. Several of these are tested using the disruption event characterization and forecasting (DECAF TM ) code for their potential effectiveness as disruption warning signals. For a limited set of NSTX discharges, a local island power balance criteria was found to be less reliable, presently, than the Greenwald limit. For this work, an empirical critical edge line density and a boundary turbulent transport limit were both tested for MAST-U, which has an electron density profile measurement with high spatial resolution in the outer part of the plasma. Both were found to have similar dependencies on key plasma parameters. In a limited set of MAST-U discharges that appear to disrupt due to rising density at values under the Greenwald limit, crossing of the boundary turbulent transport limit occurred close to the time of disruption. Finally, these limits were evaluated for their potential use in real-time, and it was found that with the necessary real-time inputs and with refinement through further testing, these limits could be implemented in a real-time disruption forecasting system.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Fusion pilot plant performance and the role of a sustained high power density tokamak

Here, recent U.S. fusion development strategy reports all recommend that the U.S. should pursue innovative science and technology to enable construction of a fusion pilot plant (FPP) that produces net electricity from fusion at low capital cost. Compact tokamaks have been proposed as a means of potentially reducing the capital cost of a FPP. However, compact steady-state tokamak FPPs face the challenge of integrating a high fraction of self-driven current with high core confinement, plasma pressure, and high divertor parallel heat flux. This integration is sufficiently challenging that a dedicated sustained-high-power-density (SHPD) tokamak facility is proposed by the U.S. community as the optimal way to close this integration gap. Performance projections for the steady-state tokamak FPP regime are presented and a preliminary SHPD device with substantial flexibility in lower aspect ratio (A = 2–2.5), shaping, and divertor configuration to narrow gaps to an FPP is described.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Cross-machine comparison of runaway electron generation during tokamak start-up for extrapolation to ITER

A cross-machine comparison of global parameters that determine the runaway electron (RE) generation and loss process during tokamak start-up was carried out with the aim to extrapolate these to ITER. The study found that all considered discharges, also those that do not show signs of RE, are non-thermal at the start, i.e. have a streaming parameter larger than 0.1. During the current ramp-up the electric field, E, remains above the critical value, E c , that allows RE in the plasma. The distinction to be made is not if RE can form but, if sufficient RE can form fast enough such that they are detected or start to dominate the dynamics of the tokamak discharge. The dynamics of the value of E, density and temperature during tokamak are key to the formation of RE. It was found that larger devices operate with E closer to E c , due to their higher temperatures, hence the RE generation is relatively slower. The slower time scales for the formation of RE, estimated to be of the order of 100s of ms in ITER simplifies the development of avoidance schemes. The RE confinement time is also an important determinant of the entire process and is found to increase with the device size. The study also revealed that drift orbit losses, a mechanism often attributed as the main RE loss mechanism during the early tokamak discharge, are actually more difficult to achieve. RE losses might be more likely attributed to RE diffusion due to magnetic turbulence.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Tokamak divertor plasma emulation with machine learning

Abstract Future tokamak devices that aim to create conditions relevant to power plant operations must consider strategies for mitigating damage to plasma facing components in the divertor. One of the goals of MAST-U tokamak operations is to inform these considerations by researching advanced divertor configurations that aid stable plasma detachment. Machine design, scenario planning and detachment control would all greatly benefit from tools that enable rapid calculation of scenario-relevant quantities given some input parameters. This paper presents a method for generating large, simulated scrape-off layer data sets, which was applied to generate a data set of steady-state Hermes-3 simulations of the MAST-U tokamak. A machine learning model was constructed using a Bayesian approach to hyperparameter optimisation to predict diagnosable output quantities given control-relevant input features. The resulting best-performing model, which is based on a feedforward neural network, achieves high accuracy when predicting electron temperature at the divertor target and carbon impurity radiation front position and runs in around 1 ms in inference mode. Techniques for interpreting the predictions made by the model were applied, and a high-resolution parameter scan of upstream conditions was performed to demonstrate the utility of rapidly generating accurate predictions using the emulator. This work represents a step forward in the design of machine learning-driven emulators of tokamak exhaust simulation codes in operational modes relevant to divertor detachment control and plasma scenario design.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗