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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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119 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Dimensionless parameter scaling of intrinsic torque in C-Mod enhanced confinement plasmas

A dimensionless parameter dependence study of intrinsic torque has been performed on a database of H- and I-mode plasmas from the Alcator C-Mod tokamak. Here, the torque was determined by comparing intrinsic angular momentum density profiles just before and just after L–H and L–I transitions. The intrinsic torque has been found to scale as ${\beta }_{N}^{ 1.5}\enspace {\rho }_{{\ast}}^{-1.0}\enspace {\nu }_{{\ast}}^{0.1}$, with the parameter ranges 0.3 $\leqslant$ β N $\leqslant$ 1.5, 0.004 $\leqslant$ ρ* $\leqslant$ 0.011 and 0.04 $\leqslant$ ν * $\leqslant$ 0.9. Comparison with results from JET and DIII-D suggests that the intrinsic torque should be normalized by some measure of the device size. Depending upon this normalization, the estimated total intrinsic torques for ITER, SPARC and ARC are ~20, ~4 and ~8 Nm, respectively.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Thermal and structural analysis of a new WEST reciprocating probe head with Titanium-Zirconium-Molybdenum armour material

The Tore Supra tokamak was transformed into an X-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support of the ITER tungsten divertor strategy. WEST began operation in 2017. A new reciprocating probe head, equipped with a Titanium-Zirconium-Molybdenum (TZM) heat shield, was recently built and operated on WEST during the C9 Campaign in early 2024. TZM was chosen as armour material for its very good mechanical properties between 700 °C and 1400 °C. This material was already extensively used for probe heads and other plasma facing components on the Alcator C-Mod tokamak. Here, this paper presents the design of the new WEST probe head. It is equipped with tungsten filaments that can be heated to emit electrons and provide a direct measurement of the plasma potential in the scrape-off layer, and it is also equipped with a pair of Langmuir probes. The probe plunges vertically into the plasma, reaches its maximum dive depth, then returns to its protected resting position in about 0.25 s. Thermal loads are calculated using as input the thermal properties of TZM and radial profiles of the heat flux measured in the past by other reciprocating probes, and scaled upwards to the expected heat loads in the WEST scrape-off layer. An analytical estimation of Electromagnetic (EM) loads will also be presented, as well as a thermal and structural analysis of the probe for 15 cycles at maximum plunge depth. The probe is observed by an infrared camera. Comparisons between the heat flux deduced by the Langmuir probe and the infrared radiance of the probe housing during plasma exposure are finally presented.

Diagnostic↗
DOE OSTI2012

Foreign Travel Trip Report for LLNL travel with DOE FES funding, May 19th-30th, 2012

I attended the 20th biannual International Conference on Plasma Surface Interaction (PSI) in Fusion Devices in Aachen, Germany, hosted this year by the Forschungszentrum Julich (FZJ) research center. The PSI conference is one of the main international forums for the presentation and discussion of results on plasma surface interactions and edge plasma physics relevant to magnetic confinement fusion devices. I disseminated the recent results of FESP/LLNL tokamak research by presenting three posters on: (i) understanding reconnection and controlling edge localized modes (ELMs) using the BOUT++ code, (ii) simulation of resistive ballooning mode turbulence, and (iii) innovative design of Snowflake divertors. I learned of many new and recent results from international tokamak facilities and had the opportunity for discussion of these topics with other scientists at the poster sessions, conference lunches/receptions, etc. Some of the major highlights of the PSI conference topics were: (1) Review of the progress in using metallic tungsten and beryllium (ITER-like) walls at international tokamak facilities: JET (Culham, UK), TEXTOR (FZJ, Germany) and Alcator CMOD (MIT, USA). Results included: effect of small and large-area melting on plasma impurity content and recovery, expected reduction in retention of hydrogenic species, increased heat load during disruptions and need for mitigation with massive gas injection. (2) A review of ELM control in general (T. Evans, GA) and recent results of ELM control using n=2 external magnetic perturbations on ASDEX-Upgrade (MPI-Garching, Germany). (3) General agreement among the international tokamak database that, along the outer midplane of a low collisionality tokamak, the SOL power width in current experiments varies inversely with respect to plasma current (Ip), roughly as 1/Ip, with little dependence on other plasma parameters. This would imply roughly a factor of 1/4 of the width that was assumed for the design of the ITER tokamak. The first studies of the implications for ITER (A. Kukushkin, ITER) have shown a great reduction in operational parameter space that, at present, can only be lifted by increasing target plate heat flux limits. During my visit to the CRPP at the EPFL, I delivered an invited talk in order to disseminate new results of the recent publication [1] on using non-axisymmetric perturbations of the SOL to control the edge plasma. I was given a tour of both the TCV tokamak and the TORPEX simple magnetized plasma device/divertor simulator. TORPEX is an excellent laboratory for exploring the physics of simple magnetized plasmas that are relevant to the scrape-off layer of a tokamak. Properly designed experiments on TORPEX can potentially be used to test the theory of controlling the edge plasma using non-axisymmetric potentials and currents in the SOL developed by LLNL described in [1].

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Overview of density pedestal structure: Role fueling versus transport

How much of the density pedestal structure profile is determined by fueling versus transport is still a valid question. In most current tokamaks, the Scrape-Off Layer is not opaque enough to screen the neutrals and limit their penetration past the separatrix. As such recycled neutrals as well as additional gas puffing can directly affect the pedestal structure through ionization. Yet, in ITER and other future burning plasma magnetic confinement devices, this ionization source will be pushed further out into the Scrape-Off Layer. This poses the question, how much of the pedestal density structure is governed by this edge ionization source versus plasma transport effects. Theoretically, several turbulent modes have been identified which could provide a 'pinch' like up-gradient transport mechanism. Up-gradient transport is necessary in a source-free region to obtain a peaked density profile, which is often observed in the core of tokamaks. In an opaque Scrape-Off Layer, without the existence of such a pinch, the pedestal density structure would eventually disappear. Not only would this limit our ability to fuel a fusion reactor, it would also affect the pedestal stability. In this paper we will give an overview of the state of the art on what sets the pedestal density structure based on experimental observations as well as theoretical models. We will complement these studies with new results from Alcator C-Mod and DIII-D where the opaqueness is increased to reach values similar to those in ITER.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Invited Review Article: Gas puff imaging diagnostics of edge plasma turbulence in magnetic fusion devices

Gas puff imaging (GPI) is a diagnostic of plasma turbulence which uses a puff of neutral gas at the plasma edge to increase the local visible light emission for improved space-time resolution of plasma fluctuations. This paper reviews gas puff imaging diagnostics of edge plasma turbulence in magnetic fusion research, with a focus on the instrumentation, diagnostic cross-checks, and interpretation issues. The gas puff imaging hardware, optics, and detectors are described for about 10 GPI systems implemented over the past similar to 15 years. Comparison of GPI results with other edge turbulence diagnostic results is described, and many common features are observed. Here, several issues in the interpretation of GPI measurements are discussed, and potential improvements in hardware and modeling are suggested.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Results and lessons learned from accelerating radio frequency modeling using machine learning [slides]

The “advanced tokamak” reactor concept is a leading candidate for a steady state fusion pilot plant. An advanced tokamak (AT) sustains a majority of the required plasma current with effects resulting from maintenance of the peaked pressure at the device center. This current is augmented by auxiliary current drive sources. These auxiliary actuators may consist of neutral particle beams and/or radio frequency (RF) systems such as lower hybrid current drive (LHCD) and high harmonic fast wave (HHFW) current drive using radio and microwaves from antennas. The primary focus of this work is to develop models of RF current profile control suitable for use in integrated modeling frameworks and for real-time control in experiments. Direct physics models of RF current drive can be computationally intensive. In order to achieve predictive times appropriate for the thousands of calls needed in real-time control of experiments and for use in integrated models, we will apply modern machine learning (ML) techniques to accelerate these models and interpolate their results. To generate the fast and accurate models for use in control level algorithms and integrated modeling we need to replace present models with high dimensional interpolation of their results. We will perform additional simulations across a broader parameter range for EAST and other tokamaks in different physics regimes (Alcator C-Mod, DIII-D, WEST, CFETR, ARC, ITER) and combine them into a larger database for training and testing of the ML models. Further testing of the control level models with experimental current profile data from EAST and C-Mod tokamaks will provide additional confirmation of the control level model before integration in a tokamak control system or integrated modeling suite. ML will be used to optimize the selection of training data consisting of RF current driven at different values of density profile, temperature profile, plasma current, and wavenumber. ML will also be used to facilitate classification of current drive from these input data. The output of this effort will be a validated classifier capable of determining the current drive profiles for HHFW CD and LHCD on a mille-second timescale. This will provide a breakthrough capability enabling real-time control of RF driven current profiles in experiments including ITER ICRF and use integrated modeling frameworks requiring thousands of current profile calculations in discharge simulations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Final Technical Report -- Bridging the PSI Knowledge Gap: A Multiscale Approach

The Plasma Surface Interactions (PSI) Science Center formed by the grant undertook a multidisciplinary set of studies on the complex interface between the plasma and solid states of matter. The strategy of the center was to combine and integrate the experimental, diagnostic and modeling toolkits from multiple institutions towards specific PSI problems. In this way the Center could tackle integrated science issues which were not addressable by single institutions, as well as evolve the underlying science of the PSI in a more general way than just for fusion applications. The overall strategy proved very successful. The research result and highlights of the MIT portion of the Center are primarily described. A particular highlight is the study of tungsten nano-tendril growth in the presence of helium plasmas. The Center research provided valuable new insights to the mechanisms controlling the nano-tendrils by developing coupled modeling and in situ diagnostic methods which could be directly compared. For example, the role of helium accumulation in tungsten distortion in the surface was followed with unique in situ helium concentration diagnostics developed. These depth-profiled, time-resolved helium concentration measurements continue to challenge the numerical models of nano-tendrils. The Center team also combined its expertise on tungsten nano-tendrils to demonstrate for the first time the growth of the tendrils in a fusion environment on the Alcator C-Mod fusion experiment, thus having significant impact on the broader fusion research effort. A new form of isolated nano-tendril “columns” were identified which are now being used to understand the underlying mechanisms controlling the tendril growth. The Center also advanced PSI science on a broader front with a particular emphasis on developing a wide range of in situ PSI diagnostic tools at the DIONISOS facility at MIT. For example the strong suppression of sputtering by the certain combination of light-species plasmas and metals was experimentally studied with independent measurement methods across the Center. This surprising result challenges the universal use of the binary-collision approximation in sputtering predictions and continues to be the subject of study. In order to address this issue MIT developed a new in situ erosion measurement technique based on ion beam analysis which can be used at elevated material temperatures. This exciting new technique is now being used to study material erosion in high performance plasma thrusters for space exploration and is being adopted to fusion experimental devices. This is an indicator of the positive synergies that arise from such a Center, with the research having impact beyond the initial area of study. The Center also served successfully as an organizing force for communication to the science community. The MIT members of the Center provided many high-profile overview presentations at prestigious international conferences and national workshops. The research resulted in three student theses and 24 peer-reviewed publications. PSI research continues to be identified as a critical area for fusion energy.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Physics and performance of the I-mode regime over an expanded operating space on Alcator C-Mod

New results on the I-mode regime of operation on the Alcator C-Mod tokamak are reported. This ELM-free regime features high energy confinement and a steep temperature pedestal, while particle confinement remains at L-mode levels, giving stationary density and avoiding impurity accumulation. I-mode has now been obtained over nearly all of the magnetic fields and currents possible in this high field tokamak (I p 0.55–1.7 MA, B T 2.8–8 T) using a configuration with B × ∇B drift away from the X-point. Results at 8 T confirm that the L–I power threshold varies only weakly with B T , and that the power range for I-mode increases with B T ; no 8 T discharges transitioned to H-mode. Parameter dependences of energy confinement are investigated. Core transport simulations are giving insight into the observed turbulence reduction, profile stiffness and confinement improvement. Pedestal models explain the observed stability to ELMs, and can simulate the observed weakly coherent mode. Conditions for I–H transitions have complex dependences on density as well as power. I-modes have now been maintained in near-DN configurations, leading to improved divertor power flux sharing. Overall, prospects for I-mode on future fusion devices such as ITER and ARC are encouraging. Further experiments on other tokamaks are needed to improve confidence in extrapolation.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Surface heat flux feedback controlled impurity seeding experiments with Alcator C-Mod’s high-Z vertical target plate divertor: performance, limitations and implications for fusion power reactors

The Alcator C-Mod team has recently developed a feedback system to measure and control surface heat flux in real-time. The system uses real-time measurements of surface heat flux from surface thermocouples and a pulse-width modulated piezo valve to inject low-Z impurities (typically N2) into the private flux region. It has been used in C-Mod to mitigate peak surface heat fluxes >40 MW/m2 down to <10 MW/m2 while maintaining excellent core confinement, H98>1. While the system works quite well under relatively steady conditions, use of it during transients has revealed important limitations on feedback control of impurity seeding in conventional vertical target plate divertors. In some cases, the system is unable to avoid plasma reattachment to the divertor plate or the formation of a confinement-damaging x-point MARFE. This is due to the small operational window for mitigated heat flux in the parameters of incident plasma heat flux, plasma density, and impurity density as well as the relatively slow response of the impurity gas injection system compared to plasma transients. Given the severe consequences for failure of such a system to operate reliably in a reactor, there is substantial risk that the conventional vertical target plate divertor will not provide an adequately controllable system in reactor-class devices. These considerations motivate the need to develop passively stable, highly compliant divertor configurations and experimental facilities that can test such possible solutions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Surface heat flux feedback controlled impurity seeding experiments with Alcator C-Mod's high-Z vertical target plate divertor: performance, limitations and implications for fusion power reactors

The Alcator C-Mod team has recently developed a feedback system to measure and control surface heat flux in real-time. The system uses real-time measurements of surface heat flux from surface thermocouples and a pulse-width modulated piezo valve to inject low-Z impurities (typically N 2 ) into the private flux region. It has been used in C-Mod to mitigate peak surface heat fluxes >40 MW m -2 down to <10 MW m -2 while maintaining excellent core confinement, H 98 > 1. While the system works quite well under relatively steady conditions, use of it during transients has revealed important limitations on feedback control of impurity seeding in conventional vertical target plate divertors. In some cases, the system is unable to avoid plasma reattachment to the divertor plate or the formation of a confinement-damaging x-point MARFE. This is due to the small operational window for mitigated heat flux in the parameters of incident plasma heat flux, plasma density, and impurity density as well as the relatively slow response of the impurity gas injection system compared to plasma transients. Given the severe consequences for failure of such a system to operate reliably in a reactor, there is substantial risk that the conventional vertical target plate divertor will not provide an adequately controllable system in reactor-class devices. Finally, these considerations motivate the need to develop passively stable, highly compliant divertor configurations and experimental facilities that can test such possible solutions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2013

Alcator C-Mod Experiments in Support of the ITER Baseline 15 MA Scenario

Experiments on Alcator C-Mod have addressed several issues for the ITER 15 MA baseline scenario from 2009-2012. Rampup studies show ICRF can save significant V-s, and that an H-mode in the ramp can be utilized to save 50% more. ICRF modifications to li(1) are minimal, although the Te profile is peaked relative to ohmic in the plasma center, and alter sawtooth onset times. Rampdown studies show H-modes can be routinely sustained, avoiding an OH coil over-current associated with the H-L transition, that fast rampdowns are preferred, the density drops with Ip, and that the H-L transition occurs at P loss /P thr,LH ~ 1.0-1.3 at n/n Gr ~ 0.85. Flattop plasmas targeting ITER baseline parameters have been sustained for 20 τ E or 8-13 τ CR , but only reach H 98 ~ 0.6 at n/n Gr = 0.85, rising to 0.9 at n/n Gr = 0.65.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗