Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “NSTX DEVICE”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Four-Dimensional Finite-Orbit-Width Fokker-Planck Code with Sources, for Neoclassical/Anomalous Transport Simulation of Ion and Electron (Final Technical Report)

Within the US Department of Energy/Office of Fusion Energy magnetic fusion research program, there is an important whole-plasma-modeling need for a radio-frequency/neutral-beam-injection (RF/NBI) transport-oriented finite-difference Fokker-Planck (FP) code with combined capabilities for 4D (2R2V) geometry near the fusion plasma periphery, and computationally less demanding 3D (1R2V) bounce-averaged capabilities for plasma in the core of fusion devices. Demonstration of proof-of-principle achievement of this goal has been carried out in research carried out under Phase I of the SBIR award. Two DOE-sponsored codes, the CQL3D bounce-average Fokker-Planck code in which CompX has specialized, and the COGENT 4D, plasma edge-oriented Fokker-Planck code which has been constructed by Lawrence Livermore National Laboratory and Lawrence Berkeley Laboratory scientists, where coupled. Coupling was achieved by using CQL3D calculated velocity distributions including an energetic tail resulting from NBI, as boundary conditions for the COGENT code over the two-dimensional velocity space on a spatial interface (flux) surface at a given radius near the plasma periphery. The finite-orbit-width fast ions from the CQL3D distributions penetrated into the peripheral plasma modeled by the COGENT code. This combined code demonstrates the feasibility of the proposed 3D/4D code. By combining these codes, the greatest computational efficiency is achieved subject to present modeling needs in toroidally symmetric magnetic fusion devices. The more efficient 3D code can be used in its regions of applicability, coupled to the more computationally demanding 4D code in higher collisionality edge plasma regions where that extended capability is necessary for accurate representation of the plasma. More efficient code leads to greater use and utility of the model. An ancillary aim of the project is to make the combined 3D/4D code user friendly. Achievement of full-coupling of these two Fokker-Planck codes will advance computational modeling of plasma devices important to the USDOE magnetic fusion energy program, in particular the DIII-D tokamak at General Atomics, San Diego, the NSTX spherical tokamak at Princeton, New Jersey, and the MST reversed-field-pinch Madison, Wisconsin. The validation studies of the code against the experiments will improve understanding of physics important for magnetic fusion, and will increase our design capabilities for achieving the goals of the International Tokamak Experimental Reactor (ITER) project in which the US is a participant and which seeks to demonstrate at least a factor of five in fusion power production divided by input power.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

High-resolution Tangential AXUV Arrays for Radiated Power Density Measurements on NSTX-U

Precise measurements of the local radiated power density and total radiated power are a matter of the uttermost importance for understanding the onset of impurity-induced instabilities and the study of particle and heat transport. Accounting of power balance is also needed for the understanding the physics of various divertor con gurations for present and future high-power fusion devices. Poloidal asymmetries in the impurity density can result from high Mach numbers and can impact the assessment of their flux-surface-average and hence vary the estimates of P[sub]rad (r, t) and (Z[sub]eff); the latter is used in the calculation of the neoclassical conductivity and the interpretation of non-inductive and inductive current fractions. To this end, the bolometric diagnostic in NSTX-U will be upgraded, enhancing the midplane coverage and radial resolution with two tangential views, and adding a new set of poloidally-viewing arrays to measure the 2D radiation distribution. These systems are designed to contribute to the near- and long-term highest priority research goals for NSTX-U which will integrate non-inductive operation at reduced collisionality, with high-pressure, long energy-confinement-times and a divertor solution with metal walls.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

Center for Momentum Transport and Flow Organization (CMTFO) (Final Technical Report)

The Center for Momentum Transport and Flow Organization (CMTFO) is a DOE Plasma Science Center formed in late 2009 to focus on the general principles underlying momentum transport in magnetic fusion and astrophysical systems. It is composed of funded researchers from UCSD, UW Madison, U. Colorado, PPPL. As of 2011, UCSD supported postdocs are collaborating at MIT/Columbia and UC Santa Cruz and beginning in 2012, will also be based at PPPL. In the initial startup period, the Center supported the construction of two basic experiments at PPPL and UW Madison to focus on accretion disk hydrodynamic instabilities and solar physics issues. We now have computational efforts underway focused on understanding recent experimental tests of dynamos, solar tachocline physics, intrinsic rotation in tokamak plasmas and L-H transition physics in tokamak devices. In addition, we have the basic experiments discussed above complemented by work on a basic linear plasma device at UCSD and a collaboration at the LAPD located at UCLA. We are also performing experiments on intrinsic rotation and L-H transition physics in the DIII-D, NSTX, C-Mod, HBT EP, HL-2A, and EAST tokamaks in the US and China, and expect to begin collaborations on K-STAR in the coming year. Center funds provide support to over 10 postdocs and graduate students each year, who work with 8 senior faculty and researchers at their respective institutions. The Center has sponsored a mini-conference at the APS DPP 2010 meeting, and co-sponsored the recent Festival de Theorie (2011) with the CEA in Cadarache, and will co-sponsor a Winter School in January 2012 in collaboration with the CMSO-UW Madison. Center researchers have published over 50 papers in the peer reviewed literature, and given over 10 talks at major international meetings. In addition, the Center co-PI, Professor Patrick Diamond, shared the 2011 Alfven Prize at the EPS meeting. Key scientific results from this startup period include initial simulations of the effects of boundary conditions on turbulent dynamo experiments; simulations of intrinsic rotation showing the strong link between toroidal rotation and temperature gradients and elucidation of the turbulence symmetry breaking mechanisms that lead to this macroscopic behavior; first experiments in a large tokamak testing the roll of turbulent momentum transport in driving intrinsic rotation; experiments in tokamaks showing strong evidence that zonal flows, together with the more widely recognized mean sheared ExB flow, act to trigger the L-H transition in tokamak devices and the first experimental measurement of collisional viscosity in an unmagnetized plasma. In the coming three year period, we will continue these efforts by a combination of basic hydrodynamic, liquid metal and plasma experiments combined with experiments on numerous tokamak devices around the world. In addition, we will use MHD, gyrofluid and gyrokinetic codes combined with theory to address the problems of interest to the Center.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

Center for Momentum Transport and Flow Organization (CMTFO) (Final Technical Report)

The Center for Momentum Transport and Flow Organization (CMTFO) is a DOE Plasma Science Center formed in late 2009 to focus on the general principles underlying momentum transport in magnetic fusion and astrophysical systems. It is composed of funded researchers from UCSD, UW Madison, U. Colorado, PPPL. As of 2011, UCSD supported postdocs are collaborating at MIT/Columbia and UC Santa Cruz and beginning in 2012, will also be based at PPPL. In the initial startup period, the Center supported the construction of two basic experiments at PPPL and UW Madison to focus on accretion disk hydrodynamic instabilities and solar physics issues. We now have computational efforts underway focused on understanding recent experimental tests of dynamos, solar tacholine physics, intrinsic rotation in tokamak plasmas and L-H transition physics in tokamak devices. In addition, we have the basic experiments discussed above complemented by work on a basic linear plasma device at UCSD and a collaboration at the LAPD located at UCLA. We are also performing experiments on intrinsic rotation and L-H transition physics in the DIII-D, NSTX, C-Mod, HBT EP, HL-2A, and EAST tokamaks in the US and China, and expect to begin collaborations on K-STAR in the coming year. Center funds provide support to over 10 postdocs and graduate students each year, who work with 8 senior faculty and researchers at their respective institutions. The Center has sponsored a mini-conference at the APS DPP 2010 meeting, and co-sponsored the recent Festival de Theorie (2011) with the CEA in Cadarache, and will co-sponsor a Winter School in January 2012 in collaboration with the CMSO-UW Madison. Center researchers have published over 50 papers in the peer reviewed literature, and given over 10 talks at major international meetings. In addition, the Center co-PI, Professor Patrick Diamond, shared the 2011 Alfven Prize at the EPS meeting. Key scientific results from this startup period include initial simulations of the effects of boundary conditions on turbulent dynamo experiments; simulations of intrinsic rotation showing the strong link between toroidal rotation and temperature gradients and elucidation of the turbulence symmetry breaking mechanisms that lead to this macroscopic behavior; first experiments in a large tokamak testing the roll of turbulent momentum transport in driving intrinsic rotation; experiments in tokamaks showing strong evidence that zonal flows, together with the more widely recognized mean sheared ExB flow, act to trigger the L-H transition in tokamak devices and the first experimental measurement of collisional viscosity in an unmagnetized plasma. In the coming three year period, we will continue these efforts by a combination of basic hydrodynamic, liquid metal and plasma experiments combined with experiments on numerous tokamak devices around the world. In addition, we will use MHD, gyrofluid and gyrokinetic codes combined with theory to address the problems of interest to the Center.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Regarding the optimization of O1-mode ECRH and the feasibility of EBW startup on NSTX-U

Recently published scenarios for fully non-inductive startup and operation on the National SphericalTorus eXperiment Upgrade (NSTX-U) (Menard et al 2012 Nucl. Fusion 52 083015) show ElectronCyclotron Resonance Heating (ECRH) as an important component in preparing a target plasma forefficient High Harmonic Fast Wave and Neutral Beam heating. The modeling of the propagation andabsorption of EC waves in the evolving plasma is required to define the most effective window ofoperation, and to optimize the launcher geometry for maximal heating and current drive during thiswindow. Here, we extend a previous optimization of O1-mode ECRH on NSTX-U to account for thefull time-dependent performance of the ECRH using simulations performed with TRANSP. We findthat the evolution of the density profile has a prominent role in the optimization by defining the timewindow of operation, which in certain cases may be a more important metric to compare launcherperformance than the average power absorption. This feature cannot be captured by analysis on staticprofiles, and should be accounted for when optimizing ECRH on any device that operates near thecutoff density. Additionally, the utility of the electron Bernstein wave (EBW) in driving current andgenerating closed flux surfaces in the early startup phase has been demonstrated on a number ofdevices. Using standalone GENRAY simulations, we find that efficient EBW current drive ispossible on NSTX-U if the injection angle is shifted below the midplane and aimed towards the tophalf of the vacuum vessel. However, collisional damping of the EBW is projected to be significant, insome cases accounting for up to 97% of the absorbed EBW power

Electron Bernstein wave↗
DOE OSTI2017

Simultaneous feedback control of plasma rotation and stored energy on NSTX-U using neoclassical toroidal viscosity and neutral beam injection

In this study, a model-based feedback system is presented enabling the simultaneous control of the stored energy through β n and the toroidal rotation profile of the plasma in National Spherical Torus eXperiment Upgrade device. Actuation is obtained using the momentum from six injected neutral beams and the neoclassical toroidal viscosity generated by applying three-dimensional magnetic fields. Based on a model of the momentum diffusion and torque balance, a feedback controller is designed and tested in closed-loop simulations using TRANSP, a time dependent transport analysis code, in predictive mode. Promising results for the ongoing experimental implementation of controllers are obtained.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Developing a predictive capability for compressional and global Alfvén eigenmodes in NSTX-U and advancing understanding of the associated electron thermal transport. Final report, 04/15/2014 - 04/14/2019

The supported research was motivated by a need to understand a robustly observed degradation of core electron thermal confinement in NSTX that is believed to be caused by beam- driven, high frequency, core-localized compressional (CAE) and global (GAE) Alfvén eigenmodes. The supported research sought to understand the role of CAEs and GAEs in the confinement degradation and to experimentally investigate leading hypotheses for the underlying mechanism. The supported research also sought to advance a predictive capability for CAEs and GAEs, which is potentially necessary for accurate prediction of Te profiles in future ST devices. Under the supported research, significant progress was made toward both goals. Multiple opportunities were realized to validate codes for predicting the characteristics of CAEs and GAEs, including the leading model for CAE and GAE stability, implemented in the Hybrid kinetic code HYM, and the CAE3B Hall MHD eigenmode code. Measurements of CAE and GAE mode internal structure and amplitude were obtained in NSTX and DIII-D (as part of the NSTX-U/DIII- D Collaborative Research National Campaign in 2017), contributing to this validation effort. In many cases, these measurements were firsts of their kind. These measurements also contributed to tests of the leading hypotheses for the role of CAEs and GAEs in core thermal confinement degradation in NSTX. The results of the tests cast doubt on these hypotheses, although questions remain. Activities under the supported research also included other firsts. For instance, experimental evidence was obtained for CAE-KAW mode coupling in NSTX (one of the hypothesized energy transport mechanisms) and a beam density threshold was observed for a CAE in DIII-D.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Self-consistent modeling of tokamak edge plasma transport with lithium sources

Magnetic confinement fusion devices require effective heat and particle exhaust solutions on the divertor plates to operate sustainably, especially under reactor-relevant conditions. Liquid lithium divertors have been proposed to address two major challenges: control of excessive heat flux to plasma-facing components through vapor shielding and minimization of core plasma contamination from impurities. The National Spherical Torus Experiment-Upgrade (NSTX-U) will explore lithium as a divertor material due to its potential to meet both objectives. We present a self-consistent coupling framework between the plasma boundary transport code UEDGE and the lithium wall transport code Wall–Li to evaluate the feasibility and operational limits of lithium-based divertors. The model aims to optimize lithium sourcing levels to prevent core plasma contamination via fuel dilution while ensuring divertor protection through vapor shielding. This integrated framework, applicable to any tokamak with lithium sources, dynamically adjusts lithium sourcing based on local plasma conditions and surface temperature. The coupled model is tested using NSTX-like geometry and plasma conditions to assess its performance and reliability. Wall–Li calculates lithium fluxes from plasma-facing components, incorporating physical sputtering, thermally enhanced sputtering, and evaporation driven by surface temperature and ion flux. These fluxes are reintroduced into UEDGE as neutral lithium atoms, enabling simulation of their transport and distribution within the plasma. UEDGE computes plasma and neutral transport, surface heat flux, and iteratively feeds this information back to Wall–Li. A small time step is employed to ensure numerical stability and convergence, enabling accurate simulations over typical tokamak discharge durations. This integrated modeling approach provides a robust tool for identifying operational regimes that balance effective lithium sourcing with minimal core plasma contamination, offering critical insights for optimizing lithium-based divertor systems in current and future fusion devices.

Magnetic confinement fusion↗
DOE OSTI2020

Suppression of edge localized modes with real-time boron injection using the tungsten divertor in EAST

We report an observation of robust suppression of edge-localized modes (ELMs) in the Experimental Advanced Superconducting Tokamak (EAST), enabled by continuous boron (B) powder injection. Edge harmonic oscillations appear during B powder injection, providing sufficient particle transport to maintain constant density and avoid impurity accumulation in ELM-stable plasmas. Here, quasi-steady ELM suppression discharges are demonstrated with modest energy confinement improvement and over a wide range of conditions: heating power and technique variation, electron density range over a factor ~3.5, deuterium or helium ion species, and with either direction of the toroidal magnetic field. ELM suppression is observed above a threshold edge B intensity and ceases within 0.5 s of termination of the B injection. In contrast to ELM suppression accompanied by recycling reduction during Li powder injection in NSTX and EAST (Maingi et al 2018 Nucl. Fusion 58 024003), reduced recycling due to hydrogenic species retention is unnecessary for the ELM suppression with B powder injection, paving the way for its consideration as an ELM control tool for future fusion devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Continuum electromagnetic gyrokinetic simulations of turbulence in the tokamak scrape-off layer and laboratory devices

Here, we present algorithms and results from Gkeyll, a full-f continuum, electromagnetic gyrokinetic code, designed to study turbulence in the edge region of fusion devices. The edge is computationally very challenging, requiring robust algorithms that can handle large-amplitude fluctuations and stable interactions with plasma sheaths. We present an energy-conserving high-order discontinuous Galerkin scheme that solves gyrokinetic equations in Hamiltonian form. Efficiency is improved by a careful choice of basis functions and automatically generated computation kernels. Previous verification tests were performed in the straight-field-line large plasma device [Shi et al., J. Plasma Phys. 83, 905830304 (2017)] and the Texas Helimak, a simple magnetized torus [Bernard et al., Phys. Plasmas 26, 042301 (2019)], including the effect of end-plate biasing on turbulence. Results for the scrape-off layer for NSTX parameters with a model helical magnetic geometry with bad curvature have been obtained [Shi et al., Phys. Plasmas 26, 012307 (2019)]. In this paper, we present algorithms for the two formulations of electromagnetic gyrokinetics: the Hamiltonian and the symplectic. We describe each formulation and show results of benchmark tests. Although our scheme works for the Hamiltonian formulation, the presence of spurious numerical modes for high-β and large $k^2_⊥ρ^2_s$ regimes shows that the symplectic formulation is more robust. We then review our recent algorithm for the symplectic formulation [Mandell et al., J. Plasma Phys. 86, 905860109 (2020)], along with example application of this new capability. Maintaining positivity of the distribution function can be challenging, and we describe a new and novel exponential recovery based algorithm to address this.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Core-edge integrated predictive studies of ST40 and NSTX plasmas with the scrape-off layer box model

The ability to model the interplay between the core and edge of tokamak plasmas is crucial to designing both the plasma operating scenario of a fusion pilot plant and the design of the tokamak itself. Scrape-off-layer (SOL) models that are tailored to integrated scenario modeling need to have fast turn-around time and minimal computational burden to enable wide parameter-space coverage for design scoping. The SOL 0-D Box model is a reduced SOL model based on global power and particle balance that captures the essential physics of SOL transport with little computational cost. The usage of the 0-D Box model in core-edge coupled simulations has been demonstrated in both interpretive and predictive modes on a variety of devices. This paper presents a sensitivity study of the 0-D Box model to the input SOL heat-flux width for an ST40 plasma. This study demonstrates that accurate prediction of this width is crucial to predicting global performance parameters of a plasma scenario, such as energy confinement time and flux consumption. We also present an extension of the Box model to 1-D to allow for parallel variation of plasma parameters along the magnetic field lines. The 1-D Box model is then compared with SOLPS-ITER simulations of an NSTX plasma. Advantages and limitations of the Box model are discussed, and future directions are outlined.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

DOE FES FY2017 Joint Research Target Fourth Quarter Milestone Report for theNational Spherical Torus Experiment Upgrade.

A successful high-performance plasma operation with a radiative divertor has been demonstrated on many tokamak devices, however, significant uncertainty remains in accurately modeling detachment thresholds, and in how detachment depends on divertor geometry. Whereas it was originally planned to perform dedicated divertor experiments on the National Spherical Tokamak Upgrade to address critical detachment and divertor geometry questions for this milestone, the experiments were deferred due to technical difficulties. Instead, existing NSTX divertor data was summarized and re-analyzed where applicable, and additional simulations were performed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Computation of Alfvèn eigenmode stability and saturation through a reduced fast ion transport model in the TRANSP tokamak transport code

Alfvénic instabilities (AEs) are well known as a potential cause of enhanced fast ion transport in fusion devices. Given a specific plasma scenario, quantitative predictions of (i) expected unstable AE spectrum and (ii) resulting fast ion transport are required to prevent or mitigate the AE- induced degradation in fusion performance. Reduced models are becoming an attractive tool to analyze existing scenarios as well as for scenario prediction in time-dependent simulations. In this work, a neutral beam heated NSTX discharge is used as reference to illustrate the potential of a reduced fast ion transport model, known as kick model, that has been recently implemented for interpretive and predictive analysis within the framework of the time-dependent tokamak transport code TRANSP. Predictive capabilities for AE stability and saturation amplitude are first assessed, based on given thermal plasma profiles only. Predictions are then compared to experimental results, and the interpretive capabilities of the model further discussed. Overall, the reduced model captures the main properties of the instabilities and associated effects on the fast ion population. Additional information from the actual experiment enables further tuning of the model’s parameters to achieve a close match with measurements.

Alfven instabilities↗
DOE OSTI2017

Computation of Alfvèn eigenmode stability and saturation through a reduced fast ion transport model in the TRANSP tokamak transport code

Alfvénic instabilities (AEs) are well known as a potential cause of enhanced fast ion transport in fusion devices. Given a specific plasma scenario, quantitative predictions of (i) expected unstable AE spectrum and (ii) resulting fast ion transport are required to prevent or mitigate the AE-induced degradation in fusion performance. Reduced models are becoming an attractive tool to analyze existing scenarios as well as for scenario prediction in time-dependent simulations. Here, in this work, a neutral beam heated NSTX discharge is used as reference to illustrate the potential of a reduced fast ion transport model, known as kick model, that has been recently implemented for interpretive and predictive analysis within the framework of the time-dependent tokamak transport code TRANSP. Predictive capabilities for AE stability and saturation amplitude are first assessed, based on given thermal plasma profiles only. Predictions are then compared to experimental results, and the interpretive capabilities of the model further discussed. Overall, the reduced model captures the main properties of the instabilities and associated effects on the fast ion population. Finally, additional information from the actual experiment enables further tuning of the model's parameters to achieve a close match with measurements.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Investigation of radiated-power for low aspect ratio fusion plasmas

Investigating radiated power in fusion plasmas is of utmost importance to understand the effect of undesired impurities, such as metals, in present devices, or also desired impurities, such as noble gasses, to purposefully radiate a large fraction of power in future devices. These studies are especially important for high Z impurities which will play a crucial role in future generation fusion pilot plants. In this work, we have developed a power radiation analysis module, which is used to investigate 2D distributions of impurity densities and radiated power asymmetries caused due to both plasma rotation and the cooling rate dependence on temperature profile for the cases of experimental NSTX plasmas and designed scenarios for the spherical tokamak advanced reactor (STAR), both being low aspect ratio tokamaks. Two different atomic databases have been tested during this work to study their impact on the radiated power distribution, especially for high Z impurities. Also in PRAM, self-consistent calculations of two-dimensional electron, main ion, and impurity ion densities are carried out using one-dimensional input density, temperature, and rotation profiles. In the case of NSTX, discharges with high rotation of ∼ 170 km s, measured with charge exchange recombination spectroscopy, have been investigated. Rotation-induced charge separation, leading to an electrostatic potential, is calculated iteratively to a self-consistent solution while testing high Z impurities to observe any 2D asymmetry in the core radiated power due to centrifugal forces. The STAR design, being much larger (R = 4 m), is projected to have a much lower rotation, and is shown to have low rotation-induced asymmetries, on the order of ten percent or less, between the low field and high field sides. However, another effect not due to rotation but to the dependence of impurity cooling rates on temperature can lead to radiation peaking off-axis, near the plasma edge. This effect is noticeable for argon in NSTX, for example, but can also be enhanced for certain impurities at much higher temperatures projected for STAR (T e0 ~ 32 keV), for example for undesired tungsten or possibly desired xenon.

NSTX↗
DOE OSTI2024

Vertical instability forecasting and controllability assessment of multi-device tokamak plasmas in DECAF with data-driven optimization

Abstract Reliable vertical position control will be an essential element of any future tokamak-based fusion power plant in order to reduce disruptions and maximize performance. We investigate methods to improve vertical controllability boundary determination in plasma operational space and demonstrate a data-driven approach based on direct pseudoinversion of operational space data that is rigorously quantitative, applicable in real-time plasma control systems, and physically intuitive to interpret. Applied to historical shot data from entire run campaigns on the MAST-U, KSTAR, and NSTX tokamaks, this approach, implemented in DECAF, improves vertical displacement event identification accuracy to 98.9%–100%. Further, we explore the application of a physics-based vertical stability metric as an early warning forecaster for vertical displacement events. The development of a linear surrogate model for the plasma current density profile, with a coefficient of determination of 0.992 on the training dataset, enables potential employment of this forecaster in real-time. The application of this approach on historical data from the MAST-U MU02 campaign yields a forecaster with 62.6% accuracy, indicating promise for this method when further refined and potentially coupled with other stability metrics.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Application of transient CHI plasma startup to future ST and AT devices

Employment of non-inductive plasma start-up techniques would considerably simplify the design of a spherical tokamak fusion reactor. Transient coaxial helicity injection (CHI) is a promising method, expected to scale favorably to next-step reactors. Yet, the implications of reactor-relevant parameters on the initial breakdown phase for CHI have not yet been considered. Here, we evaluate CHI breakdown in reactor-like configurations using an extension of the Townsend avalanche theory. We find that a CHI electrode concept in which the outer vessel wall is biased to achieve breakdown, while previously successful on NSTX and HIT-II, may exhibit a severe weakness when scaled up to a reactor. On the other hand, concepts which employ localized biasing electrodes such as those used in QUEST would avoid this issue. Assuming that breakdown can be successfully attained, we then apply scaling relationships to predict plasma parameters attainable in the transient CHI discharge. Assuming the use of 1 Wb of injector flux, we discover that plasma currents of 1 MA should be achievable. Moreover, these plasmas are expected to Ohmically self-heat with more than 1 MW of power as they decay, facilitating efficient hand-off to steady-state heating sources. These optimistic scalings are confirmed by Tokamak Simulation Code simulations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗