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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.

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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Plasma diagnostics in spherical tokamaks with silicon charged-particle detectors

Detection of charged fusion products, such as protons and tritons resulting from D(d, p) t reactions, can be used to determine the position and time dependent fusion reaction rate profile in spherical tokamak plasmas with neutral beam heating. We have developed a prototype instrument consisting of 6 ion-implanted-silicon surface barrier detectors combined with collimators in such a way that each detector can accept 3 MeV protons and 1 MeV tritons and thus provides a curved view across the plasma cross section. The combination of the results from all six detectors will provide information on the spatial distribution of the fusion reaction rate. The expected time resolution of about 1 ms makes it possible to study changes in the reaction rate due to slow variations in the neutral beam density profile, as well as rapid changes resulting from MHD instabilities. Details of the new instrument, its data acquisition system, simulation results, and electrical noise testing results are discussed in this paper. First experimental data are expected to be taken during the current experimental campaign at NSTX-U.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Design and characterization of a prototype divertor viewing infrared video bolometer for NSTX-U

The InfraRed Video Bolometer (IRVB) is a powerful tool to measure radiated power in magnetically confined plasmas due to its ability to obtain 2D images of plasma emission using a technique that is compatible with the fusion nuclear environment. A prototype IRVB has been developed and installed on NSTX-U to view the lower divertor. The IRVB is a pinhole camera which images radiation from the plasma onto a 2.5 μm thick, 9 × 7 cm{sup 2} Pt foil and monitors the resulting spatio-temporal temperature evolution using an IR camera. The power flux incident on the foil is calculated by solving the 2D+time heat diffusion equation, using the foil’s calibrated thermal properties. An optimized, high frame rate IRVB, is quantitatively compared to results from a resistive bolometer on the bench using a modulated 405 nm laser beam with variable power density and square wave modulation from 0.2 Hz to 250 Hz. The design of the NSTX-U system and benchtop characterization are presented where signal-to-noise ratios are assessed using three different IR cameras: FLIR A655sc, FLIR A6751sc, and SBF-161. The sensitivity of the IRVB equipped with the SBF-161 camera is found to be high enough to measure radiation features in the NSTX-U lower divertor as estimated using SOLPS modeling. The optimized IRVB has a frame rate up to 50 Hz, high enough to distinguish radiation during edge-localized-modes (ELMs) from that between ELMs.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Compact and multi-view solid state neutral particle analyzer arrays on National Spherical Torus Experiment-Upgrade

A compact and multi-view solid state neutral particle analyzer (SSNPA) diagnostic based on silicon photodiode arrays has been successfully tested on the National Spherical Torus Experiment-Upgrade. The SSNPA diagnostic provides spatially, temporally, and pitch-angle resolved measurements of fast-ion distribution by detecting fast neutral flux resulting from the charge exchange (CX) reactions. The system consists of three 16-channel subsystems: t-SSNPA viewing the plasma mid-radius and neutral beam (NB) line #2 tangentially, r-SSNPA viewing the plasma core and NB line #1 radially, and p-SSNPA with no intersection with any NB lines. Due to the setup geometry, the active CX signals of t-SSNPA and r-SSNPA are mainly sensitive to passing and trapped particles, respectively. In addition, both t-SSNPA and r-SSNPA utilize three vertically stacked arrays with different filter thicknesses to obtain coarse energy information. The experimental data show that all channels are operational. The signal to noise ratio is typically larger than 10, and the main noise is x-ray induced signal. The active and passive CX signals are clearly observed on t-SSNPA and r-SSNPA during NB modulation. The SSNPA data also indicate significant losses of passing particles during sawteeth, while trapped particles are weakly affected. Fluctuations up to 120 kHz have been observed on SSNPA, and they are strongly correlated with magnetohydrodynamics instabilities.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Far-infrared tangential interferometer/polarimeter design and installation for NSTX-U

The Far-infrared Tangential Interferometer/Polarimeter (FIReTIP) system has been refurbished and is being reinstalled on the National Spherical Torus Experiment—Upgrade (NSTX-U) to supply real-time line-integrated core electron density measurements for use in the NSTX-U plasma control system (PCS) to facilitate real-time density feedback control of the NSTX-U plasma. Inclusion of a visible light heterodyne interferometer in the FIReTIP system allows for real-time vibration compensation due to movement of an internally mounted retroreflector and the FIReTIP front-end optics. Real-time signal correction is achieved through use of a National Instruments CompactRIO field-programmable gate array.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Using LGI experiments to achieve better understanding of pedestal-edge coupling in NSTX-U

PowerPoint presentation. Latest advances in granule or dust injection technologies, fast and high-resolution imaging, together with micro-/nano-structured material fabrication, provide new opportunities to examine plasma-material interaction (PMI) in magnetic fusion environment. Some of our previous work in these areas is summarized. The upcoming LGI experiments in NSTX-U will shed new light on granular matter transport in the pedestal-edge region. In addition to particle control, these results can also be used for code validation and achieving better understanding of pedestal-edge coupling in fusion plasmas in both NSTX-U and others.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

Lithium pellet production (LiPP): A device for the production of small spheres of lithium

With lithium as a fusion material gaining popularity, a method for producing lithium pellets relatively quickly has been developed for NSTX. The Lithium Pellet Production device is based on an injector with a sub-millimeter diameter orifice and relies on a jet of liquid lithium breaking apart into small spheres via the Plateau-Rayleigh instability. A prototype device is presented in this paper and for a pressure difference of {Delta}P= 5 Torr, spheres with diameters between 0.91 < D < 1.37 mm have been produced with an average diameter of D= 1.14 mm, which agrees with the developed theory. Successive tests performed at Princeton Plasma Physics Laboratory with Wood's metal have confirmed the dependence of sphere diameter on pressure difference as predicted.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Design of microwave broadband CMOS transmitter and receiver circuits for MIR and ECEI plasma diagnostics

To efficiently determine the plasma electron density fluctuations using the MIR diagnostic technique, a 55–75 GHz 65 nm-CMOS transmitter has been developed where four separate intermediate frequency (IF) signals are up-converted, amplified, and then combined to generate an 8-tone RF output; a broadband 90 nm-CMOS receiver has also been constructed, which consists of an RF-low noise amplifier (LNA), mixer, and IF amplifier. The circuits and their corresponding modules will soon be deployed on the DIII-D and NSTX-U fusion devices. A 110–140 GHz 65 nm-CMOS receiver has also been designed, which is suitable for measuring the deep-core temperature fluctuations in the DIII-D tokamak using the electron cyclotron emission imaging diagnostic system. In addition to the RF-LNA/balun, mixer, and IF amplifier, an LO balun/tripler and driving amplifier are now included in this highly integrated circuit chip. By adopting the microwave and millimeterwave system-on-chip concept in the front-end system design, this paper demonstrates that compact transmitter and receiver modules can be easily built, which, in turn, facilitates array implementation and maintenance.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Comment on “Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake” [Phys. Plasmas 20, 102507 (2013)]

In the recently published paper “Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake” [Phys. Plasmas 20, 102507 (2013)], the authors raise interesting and important issues concerning divertor physics and design. However, the paper contains significant errors: (a) The conceptual framework used in it for the evaluation of divertor “quality” is reduced to the assessment of the magnetic field structure in the outer Scrape-Off Layer. This framework is incorrect because processes affecting the pedestal, the private flux region and all of the divertor legs (four, in the case of a snowflake) are an inseparable part of divertor operation. (b) The concept of the divertor index focuses on only one feature of the magnetic field structure and can be quite misleading when applied to divertor design. (c) The suggestion to rename the divertor configurations experimentally realized on NSTX (National Spherical Torus Experiment) and DIII-D (Doublet III-D) from snowflakes to X-divertors is not justified: it is not based on comparison of these configurations with the prototypical X-divertor, and it ignores the fact that the NSTX and DIII-D poloidal magnetic field geometries fit very well into the snowflake “two-null” prescription.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Modeling of carbon pellets disruption mitigation in an NSTX-U plasma

Single carbon pellet disruption mitigation simulations using M3D-C1 were conducted for an NSTX-U-like plasma to support the electromagnetic pellet injection concept (EPI). A carbon ablation model has been implemented in M3D-C1 and tested with available data. 2D simulations were conducted in order to estimate the amount of carbon needed to quench the plasma, finding that the content in a 1 mm radius vitreous carbon pellet (~3.2x10E20 atoms) would be enough if it is entirely ablated. 3D simulations were performed, scanning over pellet velocity and parallel thermal conductivity, as well as different injection directions and pellet concepts (solid pellets and shell pellets). The sensitivity of the thermal quench and other related quantities to these parameters has been evaluated. A 1 mm radius solid pellet only partially ablates at velocities of 300 m/s or higher, thus being unable to fully quench the plasma. To further enhance the ablation, approximations to an array of pellets and the shell pellet concept were also explored. 3D field line stochastization plays an important role in both quenching the center of the plasma and in heat flux losses, thus lowering the amount of carbon needed to mitigate the plasma when compared to the 2D case. This study constitutes an important step forward in `predict-first' simulations for disruption mitigation in NSTX-U and other devices, such as ITER.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Study of MHD Stability, Active Mode Control, and Disruption Avoidance in NSTX-U

Learning how to achieve sustained operation at high plasma pressure is an important goal of the National Spherical Torus Upgrade (NSTX-U) program and is critically important to any future fusion power plant. This research addresses high priority stability and control issues outlined by the NSTX-U Program. Reduced models of the stabilization physics learned continue to be created and tested against actual plasma operation in NSTX, NSTX-U, and other tokamak devices. This research is forming physics-based models for disruption forecasting that aims to unify results across different fusion machines for more confident extrapolation of the techniques studied to future devices including ITER. The Disruption Event Characterization and Forecasting (DECAF) code was expanded to analyze data from the TCV tokamak, including real-time MHD analysis signals. The presence of a mode and likelihood of its m/n numbers can be tracked, as well as the mode frequency, although a limitation at low frequency does not yet allow tracking modes to the locking point in TCV. Initial analyses of NSTX-U operation were conducted using the DECAF code. The timing of sawtooth oscillations was investigated taking advantage of the more extensive L-mode operation of the device in its first run campaign. An initial investigation of thermal collapses in NSTX-U was also conducted with specific comparison to the Greenwald density limit. A local island power balance model to explain the density limit in tokamaks has been tested with NSTX data and is being evaluated as a potential disruption forecaster in the DECAF code, presently showing similarities to the global Greenwald density evolution and limit. A module was written for the DECAF code that generalizes magnetic spectrogram analysis of MHD modes. This new portable code, written in Python, discriminates the toroidal mode number of instabilities and can track each mode as a DECAF event, monitoring the evolution of the mode through bifurcation and locking points. This is a critical new capability to allow disruption event characterization for general tokamak databases.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Advances in boronization on NSTX-Upgrade

Boronization has been effective in reducing plasma impurities and enabling access to higher density, higher confinement plasmas in many magnetic fusion devices. The National Spherical Torus eXperiment, NSTX, has recently undergone a major upgrade to NSTX-U in order to develop the physics basis for a ST-based Fusion Nuclear Science Facility (FNSF) with capability for double the toroidal field, plasma current, and NBI heating power and increased pulse duration from 1–1.5 s to 5–8 s. A new deuterated tri-methyl boron conditioning system was implemented together with a novel surface analysis diagnostic. We report on the spatial distribution of the boron deposition versus discharge pressure, gas injection and electrode location. The oxygen concentration of the plasma facing surface was measured by in-vacuo XPS and increased both with plasma exposure and with exposure to trace residual gases. Furthermore, this increase correlated with the rise of oxygen emission from the plasma.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Advances in boronization on NSTX-Upgrade

Boronization has been effective in reducing plasma impurities and enabling access to higher density, higher confinement plasmas in many magnetic fusion devices. The National Spherical Torus eXperiment, NSTX, has recently undergone a major upgrade to NSTX-U in order to develop the physics basis for a ST-based Fusion Nuclear Science Facility (FNSF) with capability for double the toroidal field, plasma current, and NBI heating power and increased pulse duration from 1–1.5 s to 5–8 s. A new deuterated tri-methyl boron conditioning system was implemented together with a novel surface analysis diagnostic. We report on the spatial distribution of the boron deposition versus discharge pressure, gas injection and electrode location. The oxygen concentration of the plasma facing surface was measured by in-vacuo XPS and increased both with plasma exposure and with exposure to trace residual gases. This increase correlated with the rise of oxygen emission from the plasma.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Edge transport studies in the edge and scrape-off layer of the National Spherical Torus Experiment with Langmuir probes

Transport and turbulence profiles were directly evaluated using probes for the first time in the edge and scrape-off layer (SOL) of NSTX [Ono et al., Nucl. Fusion 40, 557 (2000)] in low (L) and high (H) confinement, low power (P{sub in}∼ 1.3 MW), beam-heated, lower single-null discharges. Radial turbulent particle fluxes peak near the last closed flux surface (LCFS) at ≈4×10{sup 21} s{sup −1} in L-mode and are suppressed to ≈0.2×10{sup 21} s{sup −1} in H mode (80%–90% lower) mostly due to a reduction in density fluctuation amplitude and of the phase between density and radial velocity fluctuations. The radial particle fluxes are consistent with particle inventory based on SOLPS fluid modeling. A strong intermittent component is identified. Hot, dense plasma filaments 4–10 cm in diameter, appear first ∼2 cm inside the LCFS at a rate of ∼1×10{sup 21} s{sup −1} and leave that region with radial speeds of ∼3–5 km/s, decaying as they travel through the SOL, while voids travel inward toward the core. Profiles of normalized fluctuations feature levels of 10% inside LCFS to ∼150% at the LCFS and SOL. Once properly normalized, the intermittency in NSTX falls in similar electrostatic instability regimes as seen in other devices. The L-H transition causes a drop in the intermittent filaments velocity, amplitude and number in the SOL, resulting in reduced outward transport away from the edge and a less dense SOL.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

A Star on Earth

At the Energy Department's Princeton Plasma Physics Lab, scientists are trying to accomplish what was once considered the realm of science fiction: create a star on Earth. The National Spherical Torus Experiment (NSTX) is a magnetic fusion device that is used to study the physics principles of spherically shaped plasmas -- hot ionized gases in which, under the right conditions, nuclear fusion will occur. Fusion is the energy source of the sun and all of the stars. Not just limited to theoretical work, the NSTX is enabling cutting-edge research to develop fusion as a future energy source.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Resistive wall mode active control physics design for KSTAR

As KSTAR H-mode operation approaches the region where the resistive wall mode (RWM) can be unstable, an important issue for future long pulse, high beta plasma operation is to evaluate RWM active feedback control performance using a planned active/passive RWM stabilization system on the device. In particular, an optimal design of feedback sensors allows mode stabilization up to the highest achievable β{sub N} close to the ideal with-wall limit, β{sub N}{sup wall}, with reduced control power requirements. The computed ideal n = 1 mode structure from the DCON code has been input to the VALEN-3D code to calculate the projected performance of an active RWM control system in the KSTAR three-dimensional conducting structure device geometry. Control performance with the midplane locked mode detection sensors, off-midplane saddle loops, and magnetic pickup coils is examined. The midplane sensors measuring the radial component of the mode perturbation is found to be strongly affected by the wall eddy current. The off-axis saddle loops with proper compensation of the prompt applied field are computed to provide stabilization at β{sub N} up to 86% of β{sub N}{sup wall} but the low RWM amplitude computed in the off-axis regions near the sensors can produce a low signal-to-noise ratio. The required control power and bandwidth are also estimated with varied noise levels in the feedback sensors. Further improvements have been explored by examining a new RWM sensor design motivated by the off-midplane poloidal magnetic field sensors in NSTX. The new sensors mounted off of the copper passive stabilizer plates near the device midplane show a clear advantage in control performance corresponding to achieving 99% of β{sub N}{sup wall} without the need of compensation of the prompt field. The result shows a significant improvement of RWM feedback stabilization using the new sensor set which motivates a future feedback sensor upgrade.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Numerical simulations of global Alfvén eigenmodes excitation and stabilization in NSTX-U

Three-dimensional nonlinear simulations of Alfvén eigenmodes in the subcyclotron frequency range demonstrate a robust physical stabilizing mechanism via modest off-axis beam injection, in agreement with experimental observations from the National Spherical Torus Experiment (NSTX-U). Experimental results from NSTX-U have demonstrated that neutral beam injection from the new beam sources with large tangency radii deposits beam ions with large pitch, which can very effectively stabilize all unstable Global Alfvén Eigenmodes (GAEs). Beam-driven GAEs have been linked to enhanced electron transport in NSTX, and the ability to control these modes will have significant implications for NSTX-U, ITER, and other fusion devices where super-Alfvénic fast ions might be present. Nonlinear simulations using the HYM code have been performed to study the excitation and stabilization of GAEs in the NSTX-U right before and shortly after the additional off-axis beam injection. The simulations reproduce the experimental finding, namely, it is shown that off-axis neutral beam injection reliably and strongly suppresses all unstable GAEs. Before additional beam injection, the simulations show unstable counter-rotating GAEs with toroidal mode numbers and frequencies that match the experimentally determined modes. Additional off-axis beam injection has been modeled by adding beam ions with large pitch and varying density. The complete stabilization occurs at less than 7% of the total beam ion inventory. New analytical theory of GAE (de)stabilization has also been derived, indicating a different interpretation for the GAE stabilization mechanism compared to previous publications.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗