Search NASA⌕ Search

Engineering topics

Ono, M.

Publications and source records attributed to Ono, M..

At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Electromagnetic particle injector for fast time response disruption mitigation in tokamaks

A novel, rapid time-response, disruption mitigation system referred to as the electromagnetic particle injector (EPI) is described. This method can accurately deliver the radiative payload to the plasma center on a <10 ms time scale, much faster, and deeper, than what can be achieved using conventional methods. The EPI system accelerates a sabot electromagnetically. The sabot is a metallic capsule that can be accelerated to desired velocities by an electromagnetic impeller. At the end of its acceleration, within 2 ms, the sabot will release a radiative payload, which is composed of low-z granules, or a shell pellet containing smaller pellets. The primary advantage of the EPI concept over gas propelled systems is its potential to meet short warning time scales, while accurately delivering the required particle size and materials at the velocities needed for achieving the required penetration depth in high power ITER-scale discharges for thermal and runaway current disruption mitigation. In conclusion, the present experimental tests from a prototype system have demonstrated the acceleration of a 3.2 g sabot to over 150 m s –1 within 1.5 ms, consistent with the calculations, giving some degree of confidence that larger ITER-scale injector can be developed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Hysteresis Relation between Turbulence and Temperature Modulation during the Heat Pulse Propagation into a Magnetic Island in DIII-D

Here, the hysteresis relation between turbulence and temperature modulation during the heat pulse propagation into a magnetic island is studied for the first time in toroidal plasmas. Lissajous curves of the density fluctuation ($\sim\atop{n}$/n) and the electron temperature (T e ) modulation show that the ($\sim\atop{n}$/n) propagation is faster than the heat pulse propagation near the O-point of the magnetic island. This faster $\sim\atop{n}$/n propagation is an experimental evidence of the turbulence spreading from X-point to the O-point of the magnetic island.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

NSTX-U In-Vessel Control Coils’ Design Concept

In this paper, a successful conceptual design was completed to develop in-vessel control coils [non-axisymmetric control coils (NCCs)]. The NCCs are a series of saddle coils that are intended to satisfy a number of physics criteria including magnetic breaking, error field control, fast resistive wall mode (RWM) control, and edge-localized-mode stabilization. Customized mineral-insulated cable (MIC) was selected for the conductor material. The MIC was made from oxygen-free copper conductor, high purity magnesium oxide powder insulation, and 304 stainless steel sleeve cover. Sample MIC was purchased from an outside supplier, and various tests were conducted for design and performance verification including high-voltage testing, which necessitated the development of special test terminations. A concept design was also developed for terminating the MIC ends using the nonconductive and vacuum-sealed technique. Two alternative designs were proposed for joints inside the vacuum vessel (VV). The NCCs are designed to be mounted in front of the primary passive plates and underneath the plasma-facing component (PFC) tiles. The passive plates will be modified to accommodate the coils. A new PFC tile design concept was developed using high-Z materials. New designs for the penetrations through the VV wall were developed. One port per coil was planned. Finally, a new power patch panel will be required to provide the ability for various combinations of connections between the NCC, the existing RWM coils, and the existing switching power amplifier power suppliers.

42 ENGINEERING↗
DOE OSTI2018

Scenario development during commissioning operations on the National Spherical Torus Experiment Upgrade

The National Spherical Torus Experiment Upgrade (NSTX-U) will advance the physics basis required for achieving steady-state, high-beta, and high-confinement conditions in a tokamak by accessing high toroidal field (1 T) and plasma current (1.0 - 2.0 MA) in a low aspect ratio geometry (A = 1.6 - 1.8) with flexible auxiliary heating systems (12 MW NBI, 6 MW HHFW). This paper describes progress in the development of L- and Hmode discharge scenarios and the commissioning of operational tools in the first ten weeks of operation that enable the scientific mission of NSTX-U. Vacuum field calculations completed prior to operations supported the rapid development and optimization of inductive breakdown at different values of ohmic solenoid current. The toroidal magnetic field (BT0 = 0.65 T) exceeded the maximum values achieved on NSTX and novel long-pulse L-mode discharges with regular sawtooth activity exceeded the longest pulses produced on NSTX (tpulse > 1.8s). The increased flux of the central solenoid facilitated the development of stationary L-mode discharges over a range of density and plasma current (Ip). H-mode discharges achieved similar levels of stored energy, confinement (H98y,2 > 1) and stability (βN/βN-nowall > 1) compared to NSTX discharges for Ip ≤ 1 MA. High-performance H-mode scenarios require an L-H transition early in the Ip ramp-up phase in order to obtain low internal inductance (li) throughout the discharge, which is conducive to maintaining vertical stability at high elongation (κ > 2.2) and achieving long periods of MHD quiescent operations. The rapid progress in developing L- and H-mode scenarios in support of the scientific program was enabled by advances in real-time plasma control, efficient error field identification and correction, effective conditioning of the graphite wall and excellent diagnostic availability.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Overview of NSTX Upgrade initial results and modelling highlights

The National Spherical Torus Experiment (NSTX) has undergone a major upgrade, and the NSTX Upgrade (NSTX-U) Project was completed in the summer of 2015. NSTX-U first plasma was subsequently achieved, diagnostic and control systems have been commissioned, the H-mode accessed, magnetic error fields identified and mitigated, and the first physics research campaign carried out. During ten run weeks of operation, NSTX-U surpassed NSTX record pulse-durations and toroidal fields (TF), and high-performance ~1 MA H-mode plasmas comparable to the best of NSTX have been sustained near and slightly above the n = 1 no-wall stability limit and with H-mode confinement multiplier H 98y,2 above 1. Transport and turbulence studies in L-mode plasmas have identified the coexistence of at least two ion-gyro-scale turbulent micro-instabilities near the same radial location but propagating in opposite (i.e. ion and electron diamagnetic) directions. These modes have the characteristics of ion-temperature gradient and micro-tearing modes, respectively, and the role of these modes in contributing to thermal transport is under active investigation. The new second more tangential neutral beam injection was observed to significantly modify the stability of two types of Alfven eigenmodes. Improvements in offline disruption forecasting were made in the areas of identification of rotating MHD modes and other macroscopic instabilities using the disruption event characterization and forecasting code. Finally, the materials analysis and particle probe was utilized on NSTX-U for the first time and enabled assessments of the correlation between boronized wall conditions and plasma performance. These and other highlights from the first run campaign of NSTX-U are described.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Analysis of higher harmonics on bidirectional heat pulse propagation experiment in helical and tokamak plasmas

In this contribution we analyze modulation electron cyclotron resonance heating (MECH) experiment and discuss higher harmonic frequency dependence of transport coefficients. We use the bidirectional heat pulse propagation method, in which both inward propagating heat pulse and outward propagating heat pulse are analyzed at a radial range, in order to distinguish frequency dependence of transport coefficients due to hysteresis from that due to other reasons, such as radially dependent transport coefficients, a finite damping term, or boundary effects. The method is applied to MECH experiments performed in various helical and tokamak devices, i.e. Large Helical Device (LHD), TJ-II, Korea Superconducting Tokamak Advanced Research (KSTAR), and Doublet III-D (DIII-D) with different plasma conditions. Here, the frequency dependence of transport coefficients are clearly observed, showing a possibility of existence of transport hysteresis in flux-gradient relation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Overview of NSTX Upgrade Initial Results and Modelling Highlights

The National Spherical Torus Experiment (NSTX) has undergone a major upgrade, and the NSTX Upgrade (NSTX-U) Project was completed in the summer of 2015. NSTX-U first plasma was subsequently achieved, diagnostic and control systems have been commissioned, H-Mode accessed, magnetic error fields identified and mitigated, and the first physics research campaign carried out. During 10 run weeks of operation, NSTX-U surpassed NSTX-record pulse-durations and toroidal fields, and high-performance ~1MA H-mode plasmas comparable to the best of NSTX have been sustained near and slightly above the n=1 no-wall stability limit and with H-mode confinement multiplier H98y2 above 1. Transport and turbulence studies in L-mode plasmas have identified the coexistence of at least two ion-gyro-scale turbulent micro-instabilities near the same radial location but propagating in opposite (i.e. ion and electron diamagnetic) directions. These modes have the characteristics of ion-temperature gradient and micro-tearing modes, respectively, and the role of these modes in contributing to thermal transport is under active investigation. The new second more tangential neutral beam injection was observed to significantly modify the stability of two types of Alfven Eigenmodes. Improvements in offline disruption forecasting were made in the areas of identification of rotating MHD modes and other macroscopic instabilities using the Disruption Event Characterization and Forecasting (DECAF) code. Lastly, the Materials Analysis and Particle Probe (MAPP) was utilized on NSTX-U for the first time and enabled assessments of the correlation between boronized wall conditions and plasma performance. These and other highlights from the first run campaign of NSTX-U are described.

Alfven Eigenmodes↗
DOE OSTI2016

First upper limits on the radar cross section of cosmic-ray induced extensive air showers

TARA (Telescope Array Radar) is a cosmic ray radar detection experiment colocated with Telescope Array, the conventional surface scintillation detector (SD) and fluorescence telescope detector (FD) near Delta, Utah, U.S.A. Furthermore, the TARA detector combines a 40 kW, 54.1 MHz VHF transmitter and high-gain transmitting antenna which broadcasts the radar carrier over the SD array and within the FD field of view, towards a 250 MS/s DAQ receiver. TARA has been collecting data since 2013 with the primary goal of observing the radar signatures of extensive air showers (EAS). Simulations indicate that echoes are expected to be short in duration (~10 µs) and exhibit rapidly changing frequency, with rates on the order 1 MHz/µs. The EAS radar cross-section (RCS) is currently unknown although it is the subject of over 70 years of speculation. One novel signal search technique is described in which the expected radar echo of a particular air shower is used as a matched filter template and compared to waveforms obtained by triggering the radar DAQ using the Telescope Array fluorescence detector. No evidence for the scattering of radio frequency radiation by EAS is obtained to date. Finally, we report the first quantitative RCS upper limits using EAS that triggered the Telescope Array Fluorescence Detector.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2016

Power Balance Estimation in Long Duration Discharges on QUEST

Fully non-inductive plasma start-up was successfully achieved by using a well-controlled microwave source on the spherical tokamak, QUEST. Non-inductive plasmas were maintained for approximately 3–5 min, during which time power balance estimates could be achieved by monitoring wall and cooling-water temperatures. Approximately 70%–90% of the injected power could be accounted for by calorimetric measurements and approximately half of the injected power was found to be deposited on the vessel wall, which is slightly dependent on the magnetic configuration. Lastly, the power distribution to water-cooled limiters, which are expected to be exposed to local heat loads, depends significantly on the magnetic configuration, however some of the deposited power is due to energetic electrons, which have large poloidal orbits and are likely to be deposited on the plasma facing components.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Fusion Nuclear Science Facilities and Pilot Plants Based on the Spherical Tokamak

A Fusion Nuclear Science Facility (FNSF) could play an important role in the development of fusion energy by providing the nuclear environment needed to develop fusion materials and components. The spherical torus/tokamak (ST) is a leading candidate for an FNSF due to its potentially high neutron wall loading and modular configuration. A key consideration for the choice of FNSF configuration is the range of achievable missions as a function of device size. Possible missions include: providing high neutron wall loading and fluence, demonstrating tritium self-sufficiency, and demonstrating electrical self-sufficiency. All of these missions must also be compatible with a viable divertor, first-wall, and blanket solution. ST-FNSF configurations have been developed simultaneously incorporating for the first time: (1) a blanket system capable of tritium breeding ratio TBR approximately 1, (2) a poloidal field coil set supporting high elongation and triangularity for a range of internal inductance and normalized beta values consistent with NSTX/NSTX-U previous/planned operation, (3) a long-legged divertor analogous to the MAST-U divertor which substantially reduces projected peak divertor heat-flux and has all outboard poloidal field coils outside the vacuum chamber and superconducting to reduce power consumption, and (4) a vertical maintenance scheme in which blanket structures and the centerstack can be removed independently. Progress in these ST-FNSF missions vs. configuration studies including dependence on plasma major radius R0 for a range 1m to 2.2m are described. In particular, it is found the threshold major radius for TBR = 1 is R0 greater than or equal to 1.7m, and a smaller R0=1m ST device has TBR approximately 0.9 which is below unity but substantially reduces T consumption relative to not breeding. Calculations of neutral beam heating and current drive for non-inductive ramp-up and sustainment are described. An A=2, R0=3m device incorporating high-temperature superconductor toroidal field coil magnets capable of high neutron fluence and both tritium and electrical self-sufficiency is also presented following systematic aspect ratio studies.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Fusion nuclear science facilities and pilot plants based on the spherical tokamak

Here, a fusion nuclear science facility (FNSF) could play an important role in the development of fusion energy by providing the nuclear environment needed to develop fusion materials and components. The spherical torus/tokamak (ST) is a leading candidate for an FNSF due to its potentially high neutron wall loading and modular configuration. A key consideration for the choice of FNSF configuration is the range of achievable missions as a function of device size. Possible missions include: providing high neutron wall loading and fluence, demonstrating tritium self-sufficiency, and demonstrating electrical self-sufficiency. All of these missions must also be compatible with a viable divertor, first-wall, and blanket solution. ST-FNSF configurations have been developed simultaneously incorporating for the first time: (1) a blanket system capable of tritium breeding ratio TBR ≈ 1, (2) a poloidal field coil set supporting high elongation and triangularity for a range of internal inductance and normalized beta values consistent with NSTX/NSTX-U previous/planned operation, (3) a long-legged divertor analogous to the MAST-U divertor which substantially reduces projected peak divertor heat-flux and has all outboard poloidal field coils outside the vacuum chamber and superconducting to reduce power consumption, and (4) a vertical maintenance scheme in which blanket structures and the centerstack can be removed independently. Progress in these ST-FNSF missions versus configuration studies including dependence on plasma major radius R 0 for a range 1 m–2.2 m are described. In particular, it is found the threshold major radius for TBR = 1 is ${{R}_{0}}\geqslant 1.7$ m, and a smaller R 0 = 1 m ST device has TBR ≈ 0.9 which is below unity but substantially reduces T consumption relative to not breeding. Calculations of neutral beam heating and current drive for non-inductive ramp-up and sustainment are described. An A = 2, R = 3 m device incorporating high-temperature superconductor toroidal field coil magnets capable of high neutron fluence and both tritium and electrical self-sufficiency is also presented following systematic aspect ratio studies.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Upgrades toward high-heat flux, liquid lithium plasma-facing components in the NSTX-U

Liquid metal plasma-facing components (PFCs) provide numerous potential advantages over solid-material components. One critique of the approach is the relatively less developed technologies associated with deploying these components in a fusion plasma-experiment. Exploration of the temperature limits of liquid lithium PFCs in a tokamak divertor and the corresponding consequences on core operation are a high priority informing the possibilities for future liquid lithium PFCs. An all-metal NSTX-U is envisioned to make direct comparison between all high-Z wall operation and liquid lithium PFCs in a single device. By executing the all-metal upgrades incrementally, scientific productivity will be maintained while enabling physics and engineering-science studies to further develop the solid- and liquid-metal components. Six major elements of a flowing liquid-metal divertor system are described and a three-step program for implementing this system is laid out. The upgrade steps involve the first high-Z divertor target upgrade in NSTX-U, pre-filled liquid metal targets and finally, an integrated, flowing liquid metal divertor target. As a result, two example issues are described where the engineering and physics experiments are shown to be closely related in examining the prospects for future liquid metal PFCs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2016

Liquid lithium applications for solving challenging fusion reactor issues and NSTX-U contributions

Steady-state fusion reactor operation presents major divertor technology challenges, including high divertor heat flux both steady-state and transients. In addition to those issues, there are unresolved issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiative liquid lithium divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues while potentially improving the reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept and its variant, the active liquid lithium divertor concept (ARLLD), taking advantage of the enhanced Li radiation in relatively poorly confined divertor plasmas. It was estimated that only a few moles/sec of lithium injection would be needed to significantly reduce the divertor heat flux in a tokamak fusion power plant. By operating at lower temperatures ≤ 500°C than the first wall ~ 600 – 700°C, the LL-covered divertor chamber wall surfaces can serve as an effective particle pump, as impurities generally migrate toward lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity of ~ 1 liter/second (l/sec) is envisioned to sustain the steady-state operation of a 1 GW-electric class fusion power plant. By running the Li loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to outside where the dust / impurities are removed by relatively simple filter and cold/hot trap systems. Using a cold trap system, it can recover in tritium (T) in real time from LL at a rate of ~ 0.5 g / sec needed to sustain the fusion reaction while minimizing the T inventory issue. With an expected T fraction of ≤ 0.7 %, an acceptable level of T inventory can be achieved. In NSTX-U, preparations are now underway to elucidate the physics of Li plasma interactions with a number of Li application tools and Li radiation spectroscopic instruments. The NSTX-U Li evaporator which provides Li coating over the lower divertor plate, can offer important information on the RLLD concept, and the Li granule injector will test some of the key physics issue on the ARLLD concept. A LL-loop is also being prepared off line for prototyping future use on NSTX-U.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

Fast Time Response Electromagnetic Disruption Mitigation Concept

An important and urgent issue for ITER is predicting and controlling disruptions. Tokamaks and spherical tokamaks have the potential to disrupt. Methods to rapidly quench the discharge after an impending disruption is detected are essential to protect the vessel and internal components. The warning time for the onset of some disruptions in tokamaks could be <10 ms, which poses stringent requirements on the disruption mitigation system for reactor systems. In this proposed method, a cylindrical boron nitride projectile containing a radiative payload composed of boron, boron nitride, or beryllium particulate matter and weighing similar to 15 g is accelerated to velocities on the order of 1 to 2 km/s in <2 ms in a linear rail gun accelerator. A partially fragmented capsule is then injected into the tokamak discharge in the 3- to 6-ms timescale, where the radiative payload is dispersed. The device referred to as an electromagnetic particle injector has the potential to meet the short warning timescales for which a reactor disruption mitigation system must be built. The system is fully electromagnetic, with no mechanical moving parts, which ensures high reliability after a period of long standby.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2015

An overview of recent physics results from NSTX

Currently, the National Spherical Torus Experiment (NSTX) is being upgraded to operate at twice the toroidal field and plasma current (up to 1 T and 2 MA), with a second, more tangentially aimed neutral beam (NB) for current and rotation control, allowing for pulse lengths up to 5 s. Recent NSTX physics analyses have addressed topics that will allow NSTX-Upgrade to achieve the research goals critical to a Fusion Nuclear Science Facility. These include producing stable, 100% non-inductive operation in high-performance plasmas, assessing plasma-material interface (PMI) solutions to handle the high heat loads expected in the next-step devices and exploring the unique spherical torus (ST) parameter regimes to advance predictive capability. Non-inductive operation and current profile control in NSTX-U will be facilitated by co-axial helicity injection (CHI) as well as radio frequency (RF) and NB heating. CHI studies using NIMROD indicate that the reconnection process is consistent with the 2D Sweet-Parker theory. Full-wave AORSA simulations show that RF power losses in the scrape-off layer (SOL) increase significantly for both NSTX and NSTX-U when the launched waves propagate in the SOL. Moreover, Toroidal Alfven eigenmode avalanches and higher frequency Alfven eigenmodes can affect NB-driven current through energy loss and redistribution of fast ions. The inclusion of rotation and kinetic resonances, which depend on collisionality, is necessary for predicting experimental stability thresholds of fast growing ideal wall and resistive wall modes. Neutral beams and neoclassical toroidal viscosity generated from applied 3D fields can be used as actuators to produce rotation profiles optimized for global stability. DEGAS-2 has been used to study the dependence of gas penetration on SOL temperatures and densities for the MGI system being implemented on the Upgrade for disruption mitigation. PMI studies have focused on the effect of ELMs and 3D fields on plasma detachment and heat flux handling. Simulations indicate that snowflake and impurity seeded radiative divertors are candidates for heat flux mitigation in NSTX-U. Studies of lithium evaporation on graphite surfaces indicate that lithium increases oxygen surface concentrations on graphite, and deuterium-oxygen affinity, which increases deuterium pumping and reduces recycling. In situ and test-stand experiments of lithiated graphite and molybdenum indicate temperature-enhanced sputtering, although that test-stand studies also show the potential for heat flux reduction through lithium vapour shielding. Non-linear gyro kinetic simulations have indicated that ion transport can be enhanced by a shear-flow instability, and that non-local effects are necessary to explain the observed rapid changes in plasma turbulence. Predictive simulations have shown agreement between a microtearing-based reduced transport model and the measured electron temperatures in a microtearing unstable regime. Finally, two Alfven eigenmode-driven fast ion transport models have been developed and successfully benchmarked against NSTX data. Upgrade construction is moving on schedule with initial physics research operation of NSTX-U planned for mid-2015.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗