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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 1,747 records · Page 97
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Non-linear magnetohydrodynamic modeling of plasma response to resonant magnetic perturbations

The interaction of static Resonant Magnetic Perturbations (RMPs) with the plasma flows is modeled in toroidal geometry, using the non-linear resistive MHD code JOREK, which includes the X-point and the scrape-off-layer. Two-fluid diamagnetic effects, the neoclassical poloidal friction and a source of toroidal rotation are introduced in the model to describe realistic plasma flows. RMP penetration is studied taking self-consistently into account the effects of these flows and the radial electric field evolution. JET-like, MAST, and ITER parameters are used in modeling. For JET-like parameters, three regimes of plasma response are found depending on the plasma resistivity and the diamagnetic rotation: at high resistivity and slow rotation, the islands generated by the RMPs at the edge resonant surfaces rotate in the ion diamagnetic direction and their size oscillates. At faster rotation, the generated islands are static and are more screened by the plasma. An intermediate regime with static islands which slightly oscillate is found at lower resistivity. In ITER simulations, the RMPs generate static islands, which forms an ergodic layer at the very edge (ψ≥0.96) characterized by lobe structures near the X-point and results in a small strike point splitting on the divertor targets. In MAST Double Null Divertor geometry, lobes are also found near the X-point and the 3D-deformation of the density and temperature profiles is observed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

High energy flux thermo-mechanical test of 1D-carbon-carbon fibre composite prototypes for the SPIDER diagnostic calorimeter

Operation of the thermonuclear fusion experiment ITER requires additional heating via injection of neutral beams from accelerated negative ions. In the SPIDER test facility, under construction in Padova, the production of negative ions will be studied and optimised. STRIKE (Short-Time Retractable Instrumented Kalorimeter Experiment) is a diagnostic used to characterise the SPIDER beam during short pulse operation (several seconds) to verify if the beam meets the ITER requirements about the maximum allowed beam non-uniformity (below ±10%). The major components of STRIKE are 16 1D-CFC (Carbon-Carbon Fibre Composite) tiles, observed at the rear side by a thermal camera. This contribution gives an overview of some tests under high energy particle flux, aimed at verifying the thermo-mechanical behaviour of several CFC prototype tiles. The tests were performed in the GLADIS facility at IPP (Max-Plank-Institut für Plasmaphysik), Garching. Dedicated linear and nonlinear simulations were carried out to interpret the experiments and a comparison of the experimental data with the simulation results is presented. The results of some morphological and structural studies on the material after exposure to the GLADIS beam are also given.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Development of negative ion extractor in the high-power and long-pulse negative ion source for fusion application

High power and long-pulse negative ion extractor, which is composed of the plasma grid (PG) and the extraction grid (EXG), is newly developed toward the neutral beam injector for heating and current drive of future fusion machines such as ITER, JT-60 Super Advanced and DEMO reactor. The PG is designed to enhance surface production of negative ions efficiently by applying the chamfered aperture. The efficiency of the negative ion production for the discharge power increased by a factor of 1.3 against that of the conventional PG. The EXG is also designed with the thermal analysis to upgrade the cooling capability for the long pulse operation of >1000 s required in ITER. Though the magnetic field for electron suppression is reduced to 0.75 of that in the conventional EXG due to this upgrade, it was experimentally confirmed that the extracted electron current can be suppressed to the allowable level for the long pulse operation. These results show that newly developed extractor has the high potential for the long pulse extraction of the negative ions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Stress-induced martensitic transformation during tensile test of full-size TF conductor jacket tube at 4.2 K

The toroidal-field (TF) conductor jacket of International Thermonuclear Experimental Reactor (ITER) is made of modified 316LN stainless steel, which is influenced by heat treatment at approximately 650 °C for 200 h to produce Nb{sub 3}Sn superconducting materials at the final stage. Due to the high electromagnetic forces arising during magnet operation, higher mechanical properties of the jacket materials at cryogenic temperatures are required. In our work, mechanical properties of the full-size TF conductor jacket tube were investigated, which satisfied the ITER requirements. Stress-induced martensitic transformation mechanism during tensile test of the conductor jacket material at 4.2 K was characterized by means of in-situ temperature dependent XRD, vibrating sample magnetometer (VSM) and in conjunction with transmission electron microscopy (TEM). The tensile behavior related to the amount of stress-induced phase transformation at cryogenic temperature was also discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Survivability of dust in tokamaks: Dust transport in the divertor sheath

The survivability of dust being transported in the magnetized sheath near the divertor plate of a tokamak and its impact on the desired balance of erosion and redeposition for a steady-state reactor are investigated. Two different divertor scenarios are considered. The first is characterized by an energy flux perpendicular to the plate q{sub 0}≃1 MW/m{sup 2} typical of current short-pulse tokamaks. The second has q{sub 0}≃10 MW/m{sup 2} and is relevant to long-pulse machines like ITER or Demonstration Power Plant. It is shown that micrometer dust particles can survive rather easily near the plates of a divertor plasma with q{sub 0}≃1 MW/m{sup 2} because thermal radiation provides adequate cooling for the dust particle. On the other hand, the survivability of micrometer dust particles near the divertor plates is drastically reduced when q{sub 0}≃10 MW/m{sup 2}. Micrometer dust particles redeposit their material non-locally, leading to a net poloidal mass migration across the divertor. Smaller particles (with radius ∼0.1 μm) cannot survive near the divertor and redeposit their material locally. Bigger particle (with radius ∼10 μm) can instead survive partially and move outside the divertor strike points, thus causing a net loss of divertor material to dust accumulation inside the chamber and some non-local redeposition. The implications of these results for ITER are discussed.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Towards identifying the mechanisms underlying field-aligned edge-loss of HHFW power on NSTX

Fast-wave heating will be a major heating scheme on ITER, as it can heat ions directly and is relatively unaffected by the large machine size unlike neutral beams. However, fast-wave interactions with the plasma edge can lead to deleterious effects such as, in the case of the high-harmonic fast-wave (HHFW) system on NSTX, large losses of fast-wave power in the scrape off layer (SOL) under certain conditions. In such scenarios, a large fraction of the lost HHFW power is deposited on the upper and lower divertors in bright spiral shapes. The responsible mechanism(s) has not yet been identified but may include fast-wave propagation in the scrape off layer, parametric decay instability, and RF currents driven by the antenna reactive fields. Understanding and mitigating these losses is important not only for improving the heating and current-drive on NSTX-Upgrade but also for understanding fast-wave propagation across the SOL in any fast-wave system. This talk summarizes experimental results demonstrating that the flow of lost HHFW power to the divertor regions largely follows the open SOL magnetic field lines. This lost power flux is relatively large close to both the antenna and the last closed flux surface with a reduced level in between, so the loss mechanism cannot be localized to the antenna. At the same time, significant losses also occur along field lines connected to the inboard edge of the bottom antenna plate. The power lost within the spirals is roughly estimated, showing that these field-aligned losses to the divertor are significant but may not account for the total HHFW loss. To elucidate the role of the onset layer for perpendicular fast-wave propagation with regards to fast-wave propagation in the SOL, a cylindrical cold-plasma model is being developed. This model, in addition to advanced RF codes such as TORIC and AORSA, is aimed at identifying the underlying mechanism(s) behind these SOL losses, to minimize their effects in NSTX-U, and to predict their importance in ITER.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2014

Quantitative modeling of ICRF antennas with integrated time domain RF sheath and plasma physics

Significant efforts have been made to quantitatively benchmark the sheath sub-grid model used in our time-domain simulations of plasma-immersed antenna near fields, which includes highly detailed three-dimensional geometry, the presence of the slow wave, and the non-linear evolution of the sheath potential. We present both our quantitative benchmarking strategy, and results for the ITER antenna configuration, including detailed maps of electric field, and sheath potential along the entire antenna structure. Our method is based upon a time-domain linear plasma model, using the finite-difference electromagnetic Vorpal/Vsim software. This model has been augmented with a non-linear rf-sheath sub-grid model, which provides a self-consistent boundary condition for plasma current where it exists in proximity to metallic surfaces. Very early, this algorithm was designed and demonstrated to work on very complicated three-dimensional geometry, derived from CAD or other complex description of actual hardware, including ITER antennas. Initial work with the simulation model has also provided a confirmation of the existence of propagating slow waves in the low density edge region, which can significantly impact the strength of the rf-sheath potential, which is thought to contribute to impurity generation. Our sheath algorithm is based upon per-point lumped-circuit parameters for which we have estimates and general understanding, but which allow for some tuning and fitting. We are now engaged in a careful benchmarking of the algorithm against known analytic models and existing computational techniques to insure that the predictions of rf-sheath voltage are quantitatively consistent and believable, especially where slow waves share in the field with the fast wave. Currently in progress, an addition to the plasma force response accounting for the sheath potential, should enable the modeling of sheath plasma waves, a predicted additional root to the dispersion, existing at the plasma-metal interface. Our model is presently restricted to plasma in the vicinity of metallic (highly conducting) surfaces, however we report on a new effort to include thin and thick dielectric (poorly conducting) surfaces, which are of interest to both fusion devices and to industrial plasmas, where dielectric, or dielectric-coated vessels and geometry are heavily used.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2014

RF heating systems evolution for the WEST project

Tore Supra is dedicated to long pulse operation at high power, with a record in injected energy of 1 GJ (2.8 MW × 380 s) and an achieved capability of 12 MW injected power delivered by 3 RF systems: Lower Hybrid Current Drive (LHCD), Ion Cyclotron Resonance Heating (ICRH) and Electron Cyclotron Resonance Heating (ECRH). The new WEST project (W [tungsten] Environment in Steady-state Tokamak) aims at fitting Tore Supra with an actively cooled tungsten coated wall and a bulk tungsten divertor. This new device will offer to ITER a test bed for validating the relevant technologies for actively cooled metallic components, with D-shaped H-mode plasmas. For WEST operation, different scenarii able to reproduce ITER relevant conditions in terms of steady state heat loads have been identified, ranging from a high RF power scenario (15 MW, 30 s) to a high fluence scenario (10 MW, 1000 s). This paper will focus on the evolution of the RF systems required for WEST. For the ICRH system, the main issues are its ELM resilience and its CW compatibility, three new actively cooled antennas are being designed, with the aim of reducing their sensitivity to the load variations induced by ELMs. The LH system has been recently upgraded with new klystrons and the PAM antenna, the possible reshaping of the antenna mouths is presently studied for matching with the magnetic field line in the WEST configuration. For the ECRH system, the device for the poloidal movement of the mirrors of the antenna is being changed for higher accuracy and speed.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2014

Study of the absorption coefficient of alpha particles to lower hybrid waves in tokamak

Part of the energy of the Lower Hybrid (LH) waves may be absorbed by the α particles via the so-called perpendicular landau damping mechanism, which depends on various parameters of fusion reactors and the LH waves. In this article, we calculate the absorption coefficient γ{sub α} of LH waves due to α particles. Results show that, the γ{sub α} increases with the parallel refraction index n{sub ∥} while deceases with increasing the frequency of LH waves ω{sub LH} over a wide range. Higher background plasma temperature and toroidal magnetic field will increase the absorption, and there is a peak value of γ{sub α} when n{sub e}≈8×10{sup 19}m{sup −3} for ITER-like scenario. The thermal corrections to the cold plasma dispersion relation will change the damping rate to a certain extent under some specific conditions. We have also evaluated the fraction of LH power absorbed by the alpha particles, η ≈ 0.47% and 4.1% for an LH frequency of 5 GHz and 3.7 GHz respectively for ITER-like scenario. This work gives the effective reference for the choice of parameters of future fusion reactors.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2014

Study of plasma start-up initiated by second harmonic electron cyclotron resonance heating on WEGA experiment

Although both 1st harmonic ordinary mode (O1) and 2nd harmonic extra-ordinary mode (X2) have been successfully used to initiate pre-ionization and breakdown in many devices, a complete theoretical model is still missing to explain the success of this method. Moreover, some experimental observations are not completely understood, such as what occurs during the delay time between the turn-on of ECRH power and first signals of density or light measurements. Since during this free period the ECRH power has to be absorbed by in-vessel components, it is of prime importance to know what governs this delay time. Recently, dedicated start-up experiments have been performed on WEGA, using a 28 GHz ECRH system in X2-mode. This machine has the interesting capability to be run also as a tokamak allowing comparative experiments between stellarator (ι/2π > 0) and tokamak (ι/2π = 0) configurations. Different scans in heating power, neutral gas pressure, and rotational transform (ι) show clearly that the start-up is a two step process. A first step following the turn-on of the ECRH power during which no measurable electron density (or just above the noise level in some cases), ECE and radiated power is detected. Its duration depends strongly on the level of injected power. The second step corresponds to the gas ionization and plasma expansion phase, with a velocity of density build-up and filling-up of the vessel volume depending mainly on pressure, gas and rotational transform. Moreover, an interesting scenario of ECRH pre-ionization without loop voltage in tokamak configuration by applying a small optimal vertical field is relevant for start-up assistance on future experiments like ITER. The results from this experimental parametric study are useful for the modeling of the start-up assisted by the second harmonic electron cyclotron resonance heating. The aim of this work is to establish predictive scenarios for both ITER and W7-X operation.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2014

Striation pattern of target particle and heat fluxes in three dimensional simulations for DIII-D

The application of resonant magnetic perturbations results in a non-axisymmetric striation pattern of magnetic field lines from the plasma interior which intersect the divertor targets. The impact on related particle and heat fluxes is investigated by three dimensional computer simulations for two different recycling conditions (controlled via neutral gas pumping). It is demonstrated that a mismatch between the particle and heat flux striation pattern (splitting vs. no splitting), as is repeatedly observed in ITER similar shape H-mode plasmas at DIII-D, can be reproduced by the simulations for high recycling conditions at the onset of partial detachment. These results indicate that a detailed knowledge of the particle and energy balance is at least as important for realistic simulations as the consideration of a change in the magnetic field structure by plasma response effects.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Artifact reduction in industrial computed tomography via data fusion

As the most stressed part of a gas turbine the first row of turbine blades is not only a challenge for the materials used. Also the testing of these parts have to meet the highest standards. Computed tomography (CT) as the technique which could reveal the most details also provides the biggest challenges [1]: A full penetration of large sized turbine blades is often only possible at high X-ray voltages causing disproportional high costs. A reduction of the X-ray voltage is able to reduce these arising costs but yields non penetration artifacts in the reconstructed CT image. In most instances, these artifacts manifests itself as blurred and smeared regions at concave edges due to a reduced signal to noise ratio. In order to complement the missing information and to increase the overall image quality of our reconstruction, we use further imaging modalities such as a 3-D Scanner and ultrasonic imaging. A 3-D scanner is easy and cost effective to implement and is able to acquire all relevant data simultaneously with the CT projections. If, however, the interior structure is of supplemental interest, an ultrasonic imaging method is additionally used. We consider this data as a priori knowledge to employ them in an iterative reconstruction. To do so, standard iterative reconstruction methods are modified to incorporate the a priori data in a regularization approach in combination with minimizing the total variation of our image. Applying this procedure on turbine blades, we are able to reduce the apparent artifacts almost completely.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

Non-perturbative modelling of energetic particle effects on resistive wall mode: Anisotropy and finite orbit width

A non-perturbative magnetohydrodynamic-kinetic hybrid formulation is developed and implemented into the MARS-K code [Liu et al., Phys. Plasmas 15, 112503 (2008)] that takes into account the anisotropy and asymmetry [Graves et al., Nature Commun. 3, 624 (2012)] of the equilibrium distribution of energetic particles (EPs) in particle pitch angle space, as well as first order finite orbit width (FOW) corrections for both passing and trapped EPs. Anisotropic models, which affect both the adiabatic and non-adiabatic drift kinetic energy contributions, are implemented for both neutral beam injection and ion cyclotron resonant heating induced EPs. The first order FOW correction does not contribute to the precessional drift resonance of trapped particles, but generally remains finite for the bounce and transit resonance contributions, as well as for the adiabatic contributions from asymmetrically distributed passing particles. Numerical results for a 9MA steady state ITER plasma suggest that (i) both the anisotropy and FOW effects can be important for the resistive wall mode stability in ITER plasmas; and (ii) the non-perturbative approach predicts less kinetic stabilization of the mode, than the perturbative approach, in the presence of anisotropy and FOW effects for the EPs. The latter may partially be related to the modification of the eigenfunction of the mode by the drift kinetic effects.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Benchmarking kinetic calculations of resistive wall mode stability

Validating the calculations of kinetic resistive wall mode (RWM) stability is important for confidently predicting RWM stable operating regions in ITER and other high performance tokamaks for disruption avoidance. Benchmarking the calculations of the Magnetohydrodynamic Resistive Spectrum—Kinetic (MARS-K) [Y. Liu et al., Phys. Plasmas 15, 112503 (2008)], Modification to Ideal Stability by Kinetic effects (MISK) [B. Hu et al., Phys. Plasmas 12, 057301 (2005)], and Perturbed Equilibrium Nonambipolar Transport (PENT) [N. Logan et al., Phys. Plasmas 20, 122507 (2013)] codes for two Solov'ev analytical equilibria and a projected ITER equilibrium has demonstrated good agreement between the codes. The important particle frequencies, the frequency resonance energy integral in which they are used, the marginally stable eigenfunctions, perturbed Lagrangians, and fluid growth rates are all generally consistent between the codes. The most important kinetic effect at low rotation is the resonance between the mode rotation and the trapped thermal particle's precession drift, and MARS-K, MISK, and PENT show good agreement in this term. The different ways the rational surface contribution was treated historically in the codes is identified as a source of disagreement in the bounce and transit resonance terms at higher plasma rotation. Calculations from all of the codes support the present understanding that RWM stability can be increased by kinetic effects at low rotation through precession drift resonance and at high rotation by bounce and transit resonances, while intermediate rotation can remain susceptible to instability. The applicability of benchmarked kinetic stability calculations to experimental results is demonstrated by the prediction of MISK calculations of near marginal growth rates for experimental marginal stability points from the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 (2000)].

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Modeling of divertor geometry effects in China fusion engineering testing reactor by SOLPS/B2-Eirene

The China Fusion Engineering Testing Reactor (CFETR) is currently under design. The SOLPS/B2-Eirene code package is utilized for the design and optimization of the divertor geometry for CFETR. Detailed modeling is carried out for an ITER-like divertor configuration and one with relatively open inner divertor structure, to assess, in particular, peak power loading on the divertor target, which is a key issue for the operation of a next-step fusion machine, such as ITER and CFETR. As expected, the divertor peak heat flux greatly exceeds the maximum steady-state heat load of 10 MW/m{sup 2}, which is a limit dictated by engineering, for both divertor configurations with a wide range of edge plasma conditions. Ar puffing is effective at reducing divertor peak heat fluxes below 10 MW/m{sup 2} even at relatively low densities for both cases, favoring the divertor configuration with more open inner divertor structure.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2014

Preliminary design of laser-induced breakdown spectroscopy for proto-Material Plasma Exposure eXperiment

Laser-induced breakdown spectroscopy (LIBS) is a technique for measuring surface matter composition. LIBS is performed by focusing laser radiation onto a target surface, ablating the surface, forming a plasma, and analyzing the light produced. LIBS surface analysis is a possible diagnostic for characterizing plasma-facing materials in ITER. Oak Ridge National Laboratory has enabled the initial installation of a laser-induced breakdown spectroscopy diagnostic on the prototype Material-Plasma Exposure eXperiment (Proto-MPEX), which strives to mimic the conditions found at the surface of the ITER divertor. This paper will discuss the LIBS implementation on Proto-MPEX, preliminary design of the fiber optic LIBS collection probe, and the expected results.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

Performance study of LMS based adaptive algorithms for unknown system identification

Adaptive filtering techniques have gained much popularity in the modeling of unknown system identification problem. These techniques can be classified as either iterative or direct. Iterative techniques include stochastic descent method and its improved versions in affine space. In this paper we present a comparative study of the least mean square (LMS) algorithm and some improved versions of LMS, more precisely the normalized LMS (NLMS), LMS-Newton, transform domain LMS (TDLMS) and affine projection algorithm (APA). The performance evaluation of these algorithms is carried out using adaptive system identification (ASI) model with random input signals, in which the unknown (measured) signal is assumed to be contaminated by output noise. Simulation results are recorded to compare the performance in terms of convergence speed, robustness, misalignment, and their sensitivity to the spectral properties of input signals. Main objective of this comparative study is to observe the effects of fast convergence rate of improved versions of LMS algorithms on their robustness and misalignment.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2014

Thomson scattering for core plasma on DEMO

This paper describes the challenges of Thomson scattering implementation for core plasma on DEMO and evaluates the capability to measure extremely high electron temperature range 0.5-40keV. A number of solutions to be developed for ITER diagnostics are suggested in consideration of their realization for DEMO. New approaches suggested for DEMO may also be of interest to ITER and currently operating magnetic confinement devices.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗