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At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Initial Development of Fusion Magnet Simulation Capabilities for Performance and Safety Evaluation Using the MOOSE Framework

Fusion energy holds the promise of being a transformative technology as a carbon-neutral, sustainable source of energy. Whole device modeling and the development of fusion digital twins will be increasingly important for emerging fusion device concepts at both national laboratories and within the commercial fusion industry. However, meeting the challenge of whole device modeling of fusion energy devices requires robust, multiphysics, multiscale modeling and simulation technologies capable of running on large-scale supercomputers. Detailed analysis of individual systems at-scale is also required to ensure safe and efficient operation as well as provide the safety basis for future device designs and licensing activities. In a tokamak, toroidal and poloidal magnets confine and shape the fusion plasma to promote the fusion reaction. High plasma temperatures and high magnetic field requirements in modern design concepts (leading to high amounts of energy stored within each magnet) impose electrical, thermal, and mechanical loads on the magnet components, which in turn impacts the safety considerations of the magnet and their supporting systems. Idaho National Laboratory (INL) has a history of working in this space, including development and benchmarking of the Magnetic System Circuitry Analysis Program (MSCAP) and Magnet Arcing (MAGARC) codes to study magnet quench events; notably, MAGARC was used to study quenching during the ITER Engineering Design Activity. However, these legacy codes and capabilities are not parallel and scalable, and new tools are required for future advances in this area, which leads to the INL-developed Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. Developed originally for fission reactor systems under United States Department of Energy, Office of Nuclear Energy modeling and simulation programs, the MOOSE framework has been well-suited to multiscale, multiphysics modeling and simulation needs for nuclear systems. The framework is open-source, well-tested, under continuous development and deployment, and developed to a Nuclear Quality Assurance, Level 1 software quality standard. MOOSE has also been used in the fusion space previously in several projects: INL’s Tritium Migration Analysis Program, Version 8 (TMAP8) for tritium migration, UK Atomic Energy Authority’s A Unified Resource for OpenMC (fusion) Reactor Applications (AURORA) code for fusion thermo-mechanical and neutronics analysis, and Argonne National Laboratory’s Cardinal for high-fidelity computational fluid dynamics and neutronics. However, to model superconducting magnets, several MOOSE enhancements are required: additions to the current MOOSE electromagnetic capabilities, new material libraries for superconductors of interest (such as YBCO), as well as fusion-specific models for thermo-mechanics. This talk will discuss initial development activities to build these capabilities in MOOSE, focusing on initial validation and benchmarking activities. Proposed coupling workflows and future work to support the simulation of fusion magnets and magnet structural assemblies for performance and safety evaluation in MOOSE will also be discussed.

70 - PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Recent DIII-D progress toward validating models of tungsten erosion, re-deposition, and migration for application to next-step fusion devices

Abstract Fundamental mechanisms governing the erosion and prompt re-deposition of tungsten impurities in tokamak divertors are identified and analyzed to inform the lifetime of tungsten plasma-facing components in ITER and other future devices. Various experiments conducted at DIII-D to benchmark predictive models are presented, leveraging the DiMES removable sample exposure probe capability and the Metal Rings Campaign, in which toroidally symmetric rows of tungsten-coated tiles were installed in the DIII-D divertor. In tokamak divertors, the width of the electric sheath is of the order of the main ion Larmor radius, and a vast majority of sputtered tungsten impurities are typically ionized within the sheath. Therefore, W prompt redeposition is mainly governed by the ratio of the characteristic ionization mean-free path of neutral tungsten to the width of the sheath. In-situ monitoring of the prompt redeposition of tungsten impurities in divertors is demonstrated via the use of WII/WI line ratios and the ionizations/photon (S/XB) method in L-mode discharges. Even with this relatively limited set of emission measurements, net erosion measurements were found to be a consistent upper bound to an analytic scaling based on the ratio of the W ionization length, λ iz , and the width of the magnetic sheath rather than the ratio of λ iz and the W + gyro-radius. In the far-scrape-off layer (SOL) of the ITER divertor, however, it is calculated that the measurement of photon emissions associated with the ionization of tungsten impurities up to W 5 + may be required. Finally, W deposition patterns on DiMES collector probes, interpreted via DIVIMP-WallDYN modelling, reveal the key roles of progressive W erosion/re-deposition staps and E × B drifts in regulating long-range high-Z material migration.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Comprehensive analysis of disruption mitigation methods using gas and pellet-like injections in ITER-like Tokamaks

Abstract Inert-gas shielding could be an effective mechanism for protection of plasma facing surfaces (PFS) against plasma particles impact and photon radiation heat loads during transient events in fusion devices. Neutral gas injection is one promising way to mitigate erosion of tokamak components and contamination. The objective of this work is to study and optimize mitigation methods using neutral gas and pellet-like injections to decrease the heat load to the divertor surfaces and to prevent vaporization of the various internal surfaces due to transient events in ITER-like devices. The integrated self-consistent models implemented in the HEIGHTS package was used for detailed analysis of the potential secondary plasma generation from the injected inert gas, its radiative characteristics, and shielding effectiveness. We varied the density, size, and location of an argon gas cloud to minimize the disruption energy deposited into the divertor components. We also investigated innovative ways for minor changes in ITER-like internal design to mitigate disruptions. We found the optimum parameters to fully protect ITER tokamak surfaces from erosion and vaporization during plasma instabilities. This preliminary analysis showed that using Ar gas injection methods could lead to enhancement in components lifetime in ITER-like and future DEMO devices with minor design changes.

Science & Technology - Other Topics↗
DOE OSTI2023

Long plasma duration operation analyses with an international multi-machine (tokamaks and stellarators) database

Combined high-fusion performance and long-pulse operation is one of the key integration challenges for fusion energy development in magnetic devices. Addressing these challenges requires an integrated vision of physics and engineering aspects with the purpose of simultaneously increasing time duration and fusion performance. Significant progress has been made in tokamaks and stellarators, including very recent achievement in duration and/or performance. This progress is reviewed by analyzing the experimental data (109 plasma pulses with a total of 3200 data points, i.e. on average 29 data per pulse) provided by ten tokamaks (in alphabetical order: Axially Symmetric Divertor Experiment Upgrade, DIII-D, Experimental Advanced Superconducting Tokamak, Joint European Torus, JT-60 Upgrade, Korea Superconducting Tokamak Advanced Research, tokamak à configuration variable, Tokamak Fusion Test Reactor, Tore Supra, W Environment in Steady-State Tokamak) and two stellarators (Large Helical Device and Wendelstein 7-X) expanding the pioneering work of Kikuchi (Kikuchi M. and Azumi M. 2015 Frontiers in Fusion Research II: Introduction to Modern Tokamak Physics (Springer)). Data have been gathered up to January 2022 and coordination has been provided by the recently created International Energy Agency-International Atomic Energy Agency international Coordination on International Challenges on Long duration OPeration group. By exploiting the multi-machine international database, recent progress in terms of injected energies (e.g. 1730 MJ in L-mode, 425 MJ in H-mode), durations (1056 s in L-mode, 101 s in H-mode), injected powers, and sustained performance will be reviewed. Progress has been made to sustain long-pulse operation in tokamaks and stellarators with superconducting coils, actively cooled components, and/or with metallic walls. The graph of the fusion triple products as a function of duration shows a dramatic reduction of, at least two orders of magnitude when increasing the plasma duration from less than 1 s to 100 s. Indeed, long-pulse operation is usually reached in dominant electron-heating modes at reduced density (current drive optimization) but with low ion temperatures ranging from 1 to 3 keV for discharges above 100 s. Difficulties in extending the duration may arise from coupling high heating powers over long durations and the evolving plasma-wall interaction towards an unstable operational domain. Possible causes limiting the duration and critical issues to be addressed prior to ITER operation and DEMO design are reported and analyzed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Particle resonances in stellarators

Resonances of high energy particles in magnetic confinement devices due to electromagnetic instabilities can strongly modify the particle distribution, leading to a reduction in fusion power and even discharge termination and particle loss to the device walls through an avalanche. The existence of a mode particle resonance depends on the properties of the equilibrium and particle parameters, and their number, location, and density can vary with device design. Recently, the advent of more powerful computing capabilities and advanced theoretical understanding has led to the design of non-axisymmetric devices or stellarators, which could prove to be more advantageous than tokamaks. Stellarators have the advantage of being immune to major disruptions because of the very low plasma current. One of the problems shared by both types of devices is the existence of resonances in particle orbits, which can lead to large amplitude high frequency instabilities and subsequent induced particle loss. We examine the number of resonances, their location, and dependence on particle energy for some stellarator designs.

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

Projected global stability of high beta MAST-U spherical tokamak plasmas

Assessment of the limits of stability of tokamak plasmas is essential to operation in high fusion performance ranges without disruption of the plasma current. Projected equilibria have been generated for the MAST-U spherical tokamak experiment, an upgrade of the previous MAST device, in order to prepare for operation. These equilibria are scanned in pressure and current profiles, and assessed with the DCON and MARS-F stability codes to find the so-called 'no-wall' beta limit, above which resistive wall mode instabilities can be expected in the absence of other stabilising effects. The no-wall limit was generally found to increase as plasma internal inductance increased. The equilibria are also assessed for the 'with-wall' limit, theoretically the highest achievable performance point, again with the DCON and MARS-F codes, including different approximate axisymmetric walls, and with the VALEN code which includes a 3D model of the surrounding conducting structure. Similar limits were found, despite the difference between the 2D and 3D codes in the treatment of the wall. Conducting passive stabilisation plates, which were newly installed in MAST-U, are in a region of significant mode perturbation when the plasma β N is sufficiently high and eddy currents are driven in these structures. Due to the increased stabilising effect of the wall in MAST-U vs. MAST, a significant gap exists between the approximate no-wall limits of β N / l i = 6.71 and 7.13, found from DCON and MARS-F respectively, and the with-wall limits of β N / l i = 8.23 and 8.53 for the equilibrium profiles analysed in this study. This opens a region of high beta operating space in MAST-U for potentially stable operation if non-ideal effects or active control can stabilise the resistive wall mode.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

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 m 3 . 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.

EAST↗
DOE OSTI2023

Operational space and performance limiting events in the first physics campaign of MAST-U

Abstract The MAST-U fusion plasma research device, an upgrade to the Mega Amp Spherical Tokamak, has recently completed its first campaign of physics operation. MAST-U operated with Ohmic, or one or two neutral beams for heating, at 400–800 kA plasma current, in conventional or ‘SuperX’ divertor configurations. Equilibrium reconstructions provide key plasma physics parameters vs. time for each discharge, and diagrams are produced which show where the prevalence of operation occurs as well as the limits in various operational spaces. When compared to stability limits, the operation of MAST-U so far has generally stayed out of the low q , low density instability region, and below the high density Greenwald limit, high beta global stability limit, and high elongation vertical stability limit. MAST-U still has the potential to reach higher elongation, which could benefit the plasma performance. Despite the majority of operations happening below established stability limits, disruptions do occur in the flat-top phase of MAST-U plasmas. The reasons for these disruptions are highlighted, and possible strategies to avoid them and to extend the operational space of MAST-U in future campaigns are discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

MST resistive wall tearing mode simulations

Here, the Madison Symmetric Torus (MST) is a toroidal device that, when operated as a stan dard tokamak, does not have major disruptions. Unlike most tokamaks, the MST plasma is surrounded by a close fitting highly conducting wall, with a resistive wall penetration time two orders of magnitude longer than in JET or DIII-D, and three times longer than in ITER. The MST can operate with edge q a ⩽ 2, unlike standard tokamaks. Simulations presented here indicate that the MST is unstable to resistive wall tearing modes (RWTMs) and resistive wall modes (RWMs). They could in principle cause disruptions, but the predicted thermal quench time is much longer than the experimental pulse time. If the MST thermal quench time were comparable to measurements in JET and DIII-D, theory and simulations predict that disruptions would have been observed in MST. This is consistent with the modeling herein, predicting that disruptions are caused by RWTMs and RWMs. In the low q a ~ 2 regime of MST, the RWTM asymptotically satisfies the RWM dispersion relation. The transition from RWTM to RWM occurs smoothly at q a = m/n, where m, n are poloidal and toroidal mode numbers.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Modeling the complete prevention of disruption-generated runaway electron beam formation with a passive 3D coil in SPARC

Here, the potential formation of multi-mega-ampere beams of relativistic 'runaway' electrons (REs) during sudden terminations of tokamak plasmas poses a significant challenge to the tokamak's development as a fusion energy source. Here, we use state-of-the-art modeling of disruption magnetohydrodynamics coupled with a self-consistent evolution of RE generation and transport to show that a non-axisymmetric in-vessel coil will passively prevent RE beam formation during disruptions in the SPARC tokamak, a compact, high-field, high-current device capable of achieving a fusion gain Q > 2 in deuterium–tritium plasmas.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Fusion Pilot Plant performance and the role of a Sustained High Power Density tokamak

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 fusion pilot plant. 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 a FPP is described.

core-edge integration↗
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

Radiofrequency sheath rectification on WEST: application of the sheath-equivalent dielectric layer technique in tokamak geometry *

Radiofrequency sheath rectification is a phenomenon relevant to the operation of Ion Cyclotron Range of Frequencies (ICRFs) actuators in tokamaks. Techniques to model the sheath rectification on 3D ICRF antenna geometries have only recently become available (Shiraiw et al 2023 Nucl. Fusion 63 026024; Beers et al 2021 Phys. Plasmas 28 093503). In this work, we apply the 'sheath-equivalent dielectric layer' technique, used previously only on linear devices (Beers et al 2021 Phys. Plasmas 28 103508), in tokamak geometry, computing rectified sheath potentials on the WEST ICRF antenna. Advancing the state of the art in sheath rectification modeling, we compute the sheath potentials not just on the limiters, but also on the Faraday Screen bars. The calculations show a peak rectified DC potential of 300 V on the limiters and 500 V on the Faraday screen. Assuming a typical sputtering yield curve, the RF sheath rectification increases the sputtering yield from the limiters by a factor of 2.6 w.r.t. the sputtering due to the non-rectified thermal sheath.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Observation of pedestal mode spreading into SOL and broadening of divertor particle flux width on EAST tokamak

Pedestal turbulence spreading into a crape-off layer (SOL) can be used to explain the experimentally observed strong pedestal-SOL coupling and is expected to be important for the broadening of divertor deposition profiles in future devices (Xu et al 2019 Nucl. Fusion 59 126039). In the EAST tokamak, it is found that an electromagnetic (EM) mode in the pedestal region can spread into the SOL and broaden the divertor particle flux width. Multi-channel fluctuation reflectometry is used to measure the density fluctuations at the plasma edge. The EM mode rotates in the electron diamagnetic drift direction in the lab frame with a frequency range of [40–90] kHz, toroidal mode number n= 12–13 and poloidal wavenumber k θ = 0.41 cm −1 . The mode amplitude peaks around the maximum of the pedestal density gradient. As the mode amplitude increases, the reflectometry channel in the SOL can clearly capture the mode. This result suggests that the EM mode is excited in the pedestal gradient region and spreads into the SOL. It is further found that the particle flux deposition profile in the divertor is broadened as the EM mode appears.

Energy↗
DOE OSTI2020

Multi-device study of temporal characteristics of magnetohydrodynamic modes initiating disruptions

Disruptions in tokamaks are often preceded by magnetohydrodynamic (MHD) instabilities that can rotate or become locked to the wall. Measurements from magnetic diagnostics in the presence of MHD mode precursors to disruptions can yield potentially valuable input to the plasma control system, with a view to disruption avoidance, prediction and mitigation. This report presents an exploratory analysis of the growth of MHD modes and corresponding time scales on the basis of magnetic measurements in multiple tokamaks. To this end, a database was compiled using disruptive discharges from COMPASS, ASDEX Upgrade, DIII-D and JET, manually classified according to disruption root cause, and characterized by a great diversity of operational conditions and mode dynamics. The typical time during which a mode can be detected using saddle coils and the duration of the locked mode phase in the database both extend over several orders of magnitude, but generally the time scales increase with plasma size. Several additional factors are discussed that can influence these durations, including the disruption root cause. A scaling law for the locked phase duration was estimated, yielding predictions toward ITER of the order of hundreds of milliseconds or even seconds. In addition, a scaling law for the mode amplitude at the disruption onset, proposed earlier by de Vries et al. (2016), is applied to the database, and its predictive capabilities are assessed. Despite significant uncertainty on the predictions from both scaling laws, encouraging trends are observed of the fraction of disruptions that may be detected with sufficient warning time to allow mitigation or even avoidance, based solely on observations of MHD mode dynamics. When combined with similar analysis of measurements from diagnostics that are sensitive to other disruption precursors, our analysis methods and results may contribute to the reliability, robustness and generalization of disruption warning schemes for ITER.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Models of resistive wall tearing mode disruptions

Disruptions are a serious issue in tokamaks. In a disruption, the thermal energy is lost by means of an instability which could be a resistive wall tearing mode (RWTM). During precursors to a disruption, the plasma edge region cools, causing the current to contract. Model sequences of contracted current equilibria are given, and their stability is calculated. A linear stability study shows that there is a maximum value of edge q a ≈ 3 for RWTMs to occur. Further, this also implies a minimum rational surface radius normalized to plasma radius from RWTMs to be unstable. Nonlinear simulations are performed using a similar model sequence derived from an equilibrium reconstruction. There is a striking difference in the results, depending on whether the wall is ideal or resistive. With an ideal wall, the perturbations saturate at moderate amplitude, causing a minor disruption without a thermal quench. With a resistive wall, there is a major disruption with a thermal quench, if the edge q a ≤ 3. There is a sharp transition in nonlinear behavior at q a = 3. This is consistent with the linear model and with experiments. If disruptions are caused by RWTMs, then devices with highly conducting walls, such as the International Tokamak Experimental Reactor will experience much milder, tolerable, disruptions than presently predicted.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Full-wave simulation of ion cyclotron range of frequency heating in a mirror device

We are adapting the all-orders spectral algorithm (AORSA) [Jaeger Phys. Plasmas 8 (2001)] from tokamak geometry to axisymmetric magnetic mirror geometry. We will use the modified AORSA to study ion cyclotron range of frequency (ICRF) heating in the Wisconsin HTS Axisymmetric Mirror (WHAM) magnetic mirror device [D. Endrizzi J. Plasma Phys. 5 (2023)]. ICRF power will be used in WHAM to accelerate high-energy neutral beam injected deuterium ions with ion cyclotron absorption at the second to fourth harmonics of deuterium. At these harmonics, for the spatial scales present in WHAM, full-wave simulations most accurately capture wave propagation, including transmission, reflection, and absorption at the cyclotron layer. The antenna is modeled as a single strap with m = 0 excitation, where m is the azimuthal mode number. We present some preliminary results for wave coupling and heating.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗