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DOE OSTI2021

Prediction of electron density and pressure profile shapes on NSTX-U using neural networks

A new model for prediction of electron density and pressure profile shapes on NSTX and NSTX-U has been developed using neural networks. The model has been trained and tested on measured profiles from experimental discharges during the first operational campaign of NSTX-U. By projecting profiles onto empirically derived basis functions, the model is able to efficiently and accurately reproduce profile shapes. In order to project the performance of the model to upcoming NSTX-U operations, a large database of profiles from the operation of NSTX is used to test performance as a function of available data. The rapid execution time of the model is well suited to the planned applications, including optimization during scenario development activities, and real-time plasma control. A potential application of the model to real-time profile estimation is demonstrated.

Boyer, Mark↗

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 OSTI2013

Power Supply Changes for NSTX Resistive Wall Mode Coils

The National Spherical Torus Experiment (NSTX) has been designed and installed in the existing facilities at Princeton Plasma Physics Laboratory (PPPL). Most of the hardware, plant facilities, auxiliary sub-systems, and power systems originally used for the Tokamak Fusion Test Reactor (TFTR) have been used with suitable modifications to reflect NSTX needs. Prior to 2004, the NSTX power system was feeding twelve (12) circuits in the machine. In 2004 the Resistive Wall Mode (RWM) Coils were installed on the machine to correct error fields. There are six of these coils installed around the machine in the mid-plane. Since these coils need fast and accurate controls, a Switching Power Amplifier (SPA) with three sub-units was procured, installed and commissioned along with other power loop components. Two RWM Coils were connected in series and fed from one SPA sub-unit. After the initial RWM campaign, operational requirements evolved such that each of the RWM coils now requires separate power and control. Hence a second SPA with three sub-units has been procured and installed. The second unit is of improved design and has the controls and power components completely isolated. The existing thyristor rectifier is used as DC Link to both of the Switching Power Amplifiers. The controls for the RWM are integrated into the overall computer control of the DC Power systems for NSTX. This paper describes the design changes in the RWM Power system for NSTX.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2013

Hardwired Control Changes For NSTX DC Power Feeds

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2026

Investigation of Drift Effects in UEDGE Simulations of NSTX-U Edge Plasma With Lithium Divertors

Lithium is a low-Z material, and lithium-based plasma-facing components (PFCs) are planned for the National Spherical Torus Experiment Upgrade (NSTX-U) to explore potential benefits for divertor power exhaust and core plasma management. NSTX-U is a medium-sized spherical tokamak with up to 12 MW of auxiliary heating, capable of generating reactor-relevant plasma conditions. This work presents boundary plasma simulations for NSTX-U with lithium PFCs using the UEDGE code, incorporating full magnetic and 𝐄 ×𝐁 drift physics. The simulations show that drifts strongly influence heat and particle transport: they enhance convective transport, broaden the scrape-off layer heat-flux width 𝜆 𝑞 , and reduce the anomalous heat diffusivity 𝜒 required to reproduce predicted SOL heat-flux width. 𝐄 ×𝐁 drifts provide poloidal transport, while ∇𝐵 (which includes both gradB and curvature) drifts provide radial heat and particle transport. Lithium transport is also affected by drifts, with lithium ions migrating from the outer divertor to the inner divertor through the private flux region (PFR) following the 𝐄 ×𝐁 drifts flow, lowering upstream impurity lithium densities. UEDGE is self-consistently coupled with the Wall-Li model to study plasma lithium PFC interactions depending on the local lithium sourcing based on local plasma conditions and lithium surface temperature. In these simulations, lithium evaporation shows a vapor-shielding effect that reduces divertor heat flux and increases radiative losses once surface temperatures exceed 450°C. This research work provides a first step toward self-consistent modeling of lithium PFCs in NSTX-U, demonstrating the impact of drift-driven plasma transport in SOL and divertor regions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2014

NSTX-U Control System Upgrades

The National Spherical Tokamak Experiment (NSTX) is undergoing a wealth of upgrades (NSTX-U). These upgrades, especially including an elongated pulse length, require broad changes to the control system that has served NSTX well. A new fiber serial Front Panel Data Port input and output (I/O) stream will supersede the aging copper parallel version. Driver support for the new I/O and cyber security concerns require updating the operating system from Redhat Enterprise Linux (RHEL) v4 to RedHawk (based on RHEL) v6. While the basic control system continues to use the General Atomics Plasma Control System (GA PCS), the effort to forward port the entire software package to run under 64-bit Linux instead of 32-bit Linux included PCS modifications subsequently shared with GA and other PCS users. Software updates focused on three key areas: (1) code modernization through coding standards (C99/C11), (2) code portability and maintainability through use of the GA PCS code generator, and (3) support of 64-bit platforms. Central to the control system upgrade is the use of a complete real time (RT) Linux platform provided by Concurrent Computer Corporation, consisting of a computer (iHawk), an operating system and drivers (RedHawk), and RT tools (NightStar). Strong vendor support coupled with an extensive RT toolset influenced this decision. The new real-time Linux platform, I/O, and software engineering will foster enhanced capability and performance for NSTX-U plasma control.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2025

Self-consistent equilibrium and transport simulations for NSTX-U plasmas enhanced via machine learning surrogate models

The Control-Oriented Transport SIMulator (COTSIM) is an advanced equilibrium and transport code designed for simulating tokamak discharges at computational speeds suitable for control applications. COTSIM’s modular framework enables users to select models that balance accuracy with speed according to specific needs, allowing the code to operate from fast to faster-than-real-time performance levels. This work presents recent enhancements to COTSIM’s predictive accuracy for NSTX-U scenarios, achieved by integrating neural-network-based surrogate models and self-consistent equilibrium calculations. To improve source deposition predictions, a surrogate model for NUBEAM has been incorporated. Additionally, a surrogate model for the Multi-Mode Module (MMM) now supports predictions of anomalous thermal, momentum, and particle diffusivities—key factors for modeling the evolution of temperature and rotation. Each surrogate model was specifically trained for the NSTX-U operational regime to enhance COTSIM’s accuracy while maintaining computational efficiency. Moreover, COTSIM now couples fixed-boundary equilibrium solvers with its transport solvers, enabling self-consistent predictions of plasma profiles and equilibrium evolution over the discharge. Simulation results demonstrate strong agreement between COTSIM and TRANSP predictions for NSTX-U discharges. These substantial advancements expand COTSIM’s utility in model-based control applications for NSTX-U. Potential applications include simultaneous optimization of equilibrium and transport scenarios, integration into digital twins, real-time profile estimation (e.g., temperature and rotation) from limited or noisy measurements, and advanced feedback-based scenario control.

Equilibrium and transport modeling↗

DOE OSTI2025

Effects beyond ideal MHD on stability of wide and enhanced pedestal regimes in NSTX

Stability of edge-localized modes (ELMs) in spherical tokamaks is explored using the extended MHD model. Linear NIMROD simulations have been performed for three NSTX discharges 132543, 132588, and 141133, to investigate the role of resistivity, diffusivity, and shear flows on the onset of ELMs. The first discharge represents the wide pedestal regime and the later two discharges represent ELM-free enhanced pedestal H-mode. We first present the effect of toroidal rotation shear and find a flow shear destabilizing effect in these NSTX discharges. Simulations are also extended to include the two-fluid and ion gyroviscosity effects. Simulations show that the flow shear can shift the mode spectrum and alter the critical condition of ELM onset. We also uncover that ELM onset prediction in spherical tokamaks requires effects beyond MHD, in particular gyroviscosity and diamagnetic terms could stabilize Peeling-Ballooning modes consistent with the experimental observation of ELM-free regimes in NSTX. The findings give new insight into the nature of the interplay between resistivity, flows, and diamagnetic stabilization in ELM suppression and have potential applications to ELM control schemes in NSTX-U and next-generation spherical tokamaks. This study identifies the essential physical effects that must be included in future predictive and validation simulations.

NSTX↗

DOE OSTI2017

Recent progress in understanding electron thermal transport in NSTX

The anomalous level of electron thermal transport inferred in magnetically confined configurations is one of the most challenging problems for the ultimate realization of fusion power using toroidal devices: tokamaks, spherical tori and stellarators. It is generally believed that plasma instabilities driven by the abundant free energy in fusion plasmas are responsible for the electron thermal transport. The National Spherical Torus eXperiment (NSTX) (Ono et al 2000 Nucl. Fusion 40 557) provides a unique laboratory for studying plasma instabilities and their relation to electron thermal transport due to its low toroidal field, high plasma beta, low aspect ratio and large ExB flow shear. Recent findings on NSTX have shown that multiple instabilities are required to explain observed electron thermal transport, given the wide range of equilibrium parameters due to different operational scenarios and radial regions in fusion plasmas. Here we review the recent progresses in understanding anomalous electron thermal transport in NSTX and focus on mechanisms that could drive electron thermal transport in the core region. The synergy between experiment and theoretical/ numerical modeling is essential to achieving these progresses. The plans for newly commissioned NSTX-Upgrade will also be discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2026

Purpose, Design, and Analysis of the NSTX-U Centerstack Casing Lateral Support Shims

The National Spherical Torus Experiment Upgrade (NSTX-U) centerstack casing surrounds the central magnet core of NSTXU. It provides the inner vacuum boundary of the plasma chamber and supports the radially inboard plasma facing components lining the plasma chamber. The casing is supported like a cantilever, fixed at the lower flange by a connection to the TF Bundle through a G-10 ring. It is flexibly supported at the top by bellows, and insulated shims intended to react the largest part of the inventory of the disruption loads at the upper support, thereby protecting the upper and lower connections from excessive loads. A large part of the qualification of the casing and tiles requires addressing and quantifying disruption loads. These have been estimated based on extensions of the behavior of NSTX during its run period roughly 15 years ago. Structural qualification of the Upgrade used these estimates of the loading as input to calculations to demonstrate adequacy for a “most probable” disruption. The estimates are included in a requirements document to guide conservative design of all components loaded by disruptions. However, the changes made for the upgrade made the extrapolation of disruption behavior uncertain. Disruptions include axisymmetric loading from axisymmetric currents induced in the vessel shells. In addition, more complex currents called Halo currents are a result of plasma poloidal currents impinging on the casing. The halo currents add another layer of uncertain behavior and net lateral loading to the casing. As a result, NSTX-U is equipped with a number of instruments intended to assess the nature and severity of disruptions. The casing shims are required for lateral support of the casing, but these have also been converted to load cells to provide another measure of the magnitude of symmetric and asymmetric loading. Some of the logic in the development of the requirements document are presented. Electromagnetic analysis of axisymmetric and halo loading is described. Global simulations of the casing transient dynamic response, and local analysis and testing of the shim/load cell are included. The shim load cells and other instrumentation will be an important guide in planning NSTX-U experiments beyond its next run period expected in late 2026 or early 2027.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2013

Upward-facing Lithium Flash Evaporator for NSTX-U

NSTX plasma performance has been significantly enhanced by lithium conditioning [1]. To date, the lower divertor and passive plates have been conditioned by downward facing lithium evaporators (LITER) as appropriate for lower null plasmas. The higher power operation expected from NSTX-U requires double null plasma operation in order to distribute the heat flux between the upper and lower divertors making it desirable to coat the upper divertor region with Li as well. An upward aiming LITER (U-LITER) is presently under development and will be inserted into NSTX-U using a horizontal probe drive located in a 6" upper midplane port. In the retracted position the evaporator will be loaded with up to 300 mg of Li granules utilizing one of the calibrated NSTX Li powder droppers[2]. The evaporator will then be inserted into the vessel in a location within the shadow of the RF limiters and will remain in the vessel during the discharge. About 10 seconds before a discharge, it will be rapidly heated and the lithium completely evaporated onto the upper divertor, thus avoiding the complication of a shutter that prevents evaporation during the shot when the diagnostic shutters are open. The minimal time interval between the evaporation and the start of the discharge will avoid the passivation of the lithium by residual gases and enable the study of the conditioning effects of un-passivated Li surfaces [3]. Two methods are being investigated to accomplish the rapid (few second) heating of the lithium. A resistive method relies on passing a large current through a Li filled crucible. A second method requires using a 3 kW e-beam gun to heat the Li. In this paper the evaporator systems will be described and the pros and cons of each heating method will be discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2015

Impact of ELM filaments on divertor heat flux dynamics in NSTX

In this study, the ELM induced change in wetted area (A wet ) and peak heat flux (q peak ) of divertor heat flux is investigated as a function of the number of striations, which represent ELM filaments, observed in the heat flux profile in NSTX. More striations are found to lead to larger A wet and lower q peak . The typical number of striations observed in NSTX is 0–9, while 10–15 striations are normally observed in other machines such as JET, and the ELM contracts heat flux profile when the number of striations is less than 3–4 but broadens it with more of them. The smaller number of striations in NSTX is attributed to the fact that NSTX ELMs are against kink/peeling boundary with lower toroidal mode number (n = 1–5), while typical peeling–ballooning ELMs have higher mode number of n = 10–20. For ELMs with smaller number of striations, relative A wet change is rather constant and qpeak change rapidly increases with increasing ELM size, while A wet change slightly increases leading to a weaker increase of q peak change for ELMs with larger number of striations, both of which are unfavourable trend for the material integrity of divertor tiles.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2021

Microtearing instabilities and electron thermal transport in low and high collisionality NSTX discharges

Microtearing mode (MTM) real frequency, growth rate, magnetic fluctuation amplitude and resulting electron thermal transport are studied in systematic NSTX scans of relevant plasma parameters. Here, the dependency of the MTM real frequency and growth rate on plasma parameters, suitable for low and high collision NSTX discharges, is obtained by using the reduced MTM transport model. The plasma parameter dependencies are compared and found to be consistent with the results obtained from MTM using the Gyrokinetic GYRO code. The scaling trend of collision frequency and plasma beta is found to be consistent with the global energy confinement trend observed in the NSTX experiment. The strength of the magnetic fluctuation is found to be consistent with the gyrokinetic estimate. In earlier studies, it was found that the version of the Multi-Mode (MM) anomalous transport model, which did not contain the effect of MTMs, provided an appropriate description of the electron temperature profiles in standard tokamak discharges and not in spherical tokamaks. When the MM model, which involves transport associated with MTMs, is incorporated in the TRANSP code and is used in the study of electron thermal transport in NSTX discharges, it is observed that the agreement with the experimental electron temperature profile is substantially improved.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2023

On the frequency bifurcations of the MHD startup modes in NSTX

The observed bifurcations of the low frequency (<50 kHz) and low toroidal periodicity (n < 5) magnetohydrodynamic (MHD) activity often present in the initial part of the National Spherical Tokamak Experiment (NSTX) discharges can be explained by the evolution of the radial profile of the safety factor (q=rB φ /RB θ ) crossing multiple rational surfaces in the core. Important performance limiting instability mechanisms in the NSTX spherical tokamak are often linked to low frequency and low-n MHD activity. They are quite common in long-pulse NSTX plasmas. They can be present at the beginning of the plasma current flat-top, at the end of the discharge or during the whole duration, and they have been observed to deleteriously impact performance over a wide range of q 95 . An interesting feature observed in some NSTX discharges is the presence of a bifurcation in the frequency of the low n modes, as low as n = 1, that have frequencies comparable to the plasma core rotation divided by n. Equilibrium reconstructions constrained by magnetic diagnostics data and motional stark effect pitch angle radial profiles suggest that the observed bifurcations are linked to a fast evolving minimum value of q. Finally, 3D non-linear resistive MHD simulations show that these modes are ideal and exist as non-resonant before the correspondent rational surface enters the plasma.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2025

Analysis of the impact of parallel magnetic fluctuations on linear gyrokinetic stability in NSTX-U and verification of gyro-fluid models

In this work, we use the CGYRO gyrokinetic code to analyze two L- and one H-mode discharges from the National Spherical Torus Experiment (NSTX) and NSTX-Upgrade (NSTX-U) selected due to their different mix of ion-scale driftwaves, ion temperature gradient (ITG) mode and trapped electron mode (TEM), and electromagnetic instabilities, kinetic ballooning mode (KBM), and micro-tearing mode (MTM) in the plasma core. It is found that the effect of parallel magnetic fluctuations is strongly destabilizing to the unstable KBMs compared to calculations with only perpendicular magnetic fluctuations. Two discharges have a mix of ITG/TEM and MTMs that are predicted to be dominant instability across the plasma radius. The parallel magnetic fluctuations are found to have little effect on the MTM stability but are destabilizing to ITG/TEM modes. To test the validity of the gyro-fluid linear stability codes TGLF and GFS at low aspect ratio, a database of linear growth rates has been created using the CGYRO gyrokinetic code. The database is comprised of various parameter scans around a standardized set of NSTX-U core parameters. It contains a group of electrostatic cases and an electromagnetic group that includes the effects of perpendicular and parallel magnetic fluctuations. Comparing the results from the GFS and TGLF models, we find that GFS exhibits the best agreement with the database of CGYRO linear growth rates. Comparing the model results for the electromagnetic scans shows that GFS captures the effects of parallel magnetic fluctuations accurately, while the TGLF model does not, as it lacks sufficient perpendicular energy resolution.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2017

Edge loss of high-harmonic fast-wave heating power in NSTX: a cylindrical model

Efficient high-harmonic fast-wave (HHFW) heating in the National Spherical Torus Experiment Upgrade (NSTX-U) would facilitate experiments in turbulence, transport, fast-ion studies, and more. However, previous HHFW operation in NSTX exhibited a large loss of fast-wave power to the divertor along the scrape-off layer field lines for edge densities above the fast-wave cutoff. It was postulated that the wave amplitude is enhanced in the scrapeoff layer due to cavity-like modes, and that these enhanced fields drive sheath losses through RF rectification. As part of ongoing work to confirm this hypothesis, we have developed a cylindrical cold-plasma model to identify and understand scenarios where a substantial fraction of wave power is confined to the plasma periphery. We previously identified a peculiar class of modes, named annulus resonances, that conduct approximately half of their wave power in the periphery and can also account for a significant fraction of the total wave power. Here, we study the influence of annulus resonances on wave field reconstructions and find instances where annulus-resonant modes dominate the spectrum and trap over half of the total wave power at the edge. The work is part of an ongoing effort to determine the mechanism underlying these scrape-off layer losses in NSTX, identify optimal conditions for operation in NSTX-U, and predict whether similar losses occur for the ion-cyclotron minority heating scheme for both current experiments and future devices such as ITER.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2025

Reversed magnetic shear scenario development in NSTX-U using TRANSP

Abstract Understanding and control of electron thermal transport is a critical point of research in magnetic fusion experiments. Previous experiments have shown that operation with reversed magnetic shear (RMS) can suppress electron thermal transport, resulting in the generation of internal transport barriers (ITBs), with the location of the ITB correlated with the location of minimum magnetic shear. The recent upgrades to NSTX—increased magnetic field up to 1 T, increased plasma current up to 2 MA, 2nd neutral beam—present an increased operating space in which to explore electron thermal transport in RMS plasmas. Utilizing TRANSP, we have developed operating scenarios by which to generate RMS in NSTX-U. The results suggest that RMS in NSTX-U can be generated through fast current ramp and early beam injection into a large plasma volume. This is very similar to the procedure that was followed in both TFTR and NSTX to generate RMS. Sustainment of RMS, disregarding non-( q min = 1) MHD events, requires maintaining a large plasma volume, and increasing the core T e , either via increased plasma current and/or adding heating power. Using this procedure, RMS was sustained for ∼1 s, with q min > 1 for that period.

Galante, M. E. (ORCID:0009000098149425)↗

DOE OSTI2024

Real-Time Ethernet Interface for NSTX-U’s Thomson Scattering Diagnostic (2023)

Here, the multipoint Thomson scattering (MPTS) diagnostic system at the National Spherical Torus Experiment Upgrade (NSTX-U) facility is undergoing an upgrade to operate in real-time and interface with the plasma control system (PCS) for NSTX-U. Previous prototyping efforts have shown that spectral analysis and rapid calculations of electron temperature and density are possible on a real-time Linux machine when using up to a 100-Hz laser pulse repetition rate. A remaining challenge was transferring the real-time data to NSTX-U’s PCS, which utilizes the front panel data port (FPDP) protocol. The original proposed method was to convert the real-time data into analog values, but a new solution was developed to keep the output format digital by using an Ethernet controller with a field-programmable gate array (FPGA). This article focuses on a new input module that has been developed to accept incoming user datagram protocol (UDP) packets sent over Ethernet, convert into FPDP format, and integrate into the existing data stream under NSTX-U’s real-time framework.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗