R. Andre

Bio: R. Andre is an academic researcher from Princeton Plasma Physics Laboratory. The author has contributed to research in topics: Tokamak & Neutral beam injection. The author has an hindex of 13, co-authored 20 publications receiving 1019 citations.

Experimental test of the neoclassical theory of impurity poloidal rotation in tokamaksa)

Abstract: Despite the importance of rotation in fusion plasmas, our present understanding of momentum transport is inadequate. The lack of understanding is in part related to the difficulty of performing accurate rotation measurements, especially for poloidal rotation. Recently, measurements of poloidal rotation for impurity ions (Z>1) have been obtained in the core of DIII-D [J. L. Luxon, Nucl. Fusion 42, 6114 (2002)] plasmas using charge exchange recombination spectroscopy. The inferred poloidal rotation is based on careful consideration of the effective energy-dependent cross section and of the gyromotion of the ions. The rotation measurements are found to be consistent with the radial electric field determined independently from multiple impurity species as well as from motional Stark effect spectroscopic measurements. The poloidal rotation measurements have been compared with predictions based on the neoclassical theory of poloidal rotation from the code NCLASS [W. A. Houlberg et al., Phys. Plasmas 4, 3230 (1997)]. The comparison shows that the neoclassically predicted poloidal rotation is in general significantly smaller than the actual measurements.

Comparison of poloidal velocity measurements to neoclassical theory on the National Spherical Torus Experimenta)

Abstract: Knowledge of poloidal velocity is necessary for the determination of the radial electric field, which along with its gradient is linked to turbulence suppression and transport barrier formation. Recent measurements of poloidal flow on conventional tokamaks have been reported to be an order of magnitude larger than expected from neoclassical theory. In contrast, poloidal velocity measurements on the NSTX spherical torus [Kaye et al., Phys. Plasmas 8, 1977 (2001)] are near or below neoclassical estimates. A novel charge exchange recombination spectroscopy diagnostic is used, which features active and passive sets of up/down symmetric views to produce line-integrated poloidal velocity measurements that do not need atomic physics corrections. Inversions are used to extract local profiles from line-integrated active and background measurements. Poloidal velocity measurements are compared with neoclassical values computed with the codes NCLASS [Houlberg et al., Phys. Plasmas 4, 3230 (1997)] and GTC-NEO [Wang et...

Exploration of the equilibrium operating space for NSTX-Upgrade

Abstract: This paper explores a range of high-performance equilibrium scenarios achievable with neutral beam heating in the NSTX-Upgrade device (Menard J.E. 2012 Nucl. Fusion 52 083015). NSTX-Upgrade is a substantial upgrade to the existing NSTX device (Ono M. et al 2000 Nucl. Fusion 40 557), with significantly higher toroidal field and solenoid capabilities, and three additional neutral beam sources with significantly larger current-drive efficiency. Equilibria are computed with free-boundary TRANSP, allowing a self-consistent calculation of the non-inductive current-drive sources, the plasma equilibrium and poloidal-field coil currents, using the realistic device geometry. The thermal profiles are taken from a variety of existing NSTX discharges, and different assumptions for the thermal confinement scalings are utilized. The no-wall and ideal-wall n = 1 stability limits are computed with the DCON code. The central and minimum safety factors are quite sensitive to many parameters: they generally increase with large outer plasma-wall gaps and higher density, but can have either trend with the confinement enhancement factor. In scenarios with strong central beam current drive, the inclusion of non-classical fast-ion diffusion raises qmin, decreases the pressure peaking, and generally improves the global stability, at the expense of a reduction in the non-inductive current-drive fraction; cases with less beam current drive are largely insensitive to additional fast-ion diffusion. The non-inductive current level is quite sensitive to the underlying confinement and profile assumptions. For instance, for BT = 1.0 T and Pinj = 12.6 MW, the non-inductive current level varies from 875 kA with ITER-98y,2 thermal confinement scaling and narrow thermal profiles to 1325 kA for an ST specific scaling expression and broad profiles. Scenarios are presented which can be sustained for 8–10 s, or (20–30) τCR, at βN = 3.8–4.5. The value of qmin can be controlled at either fixed non-inductive fraction of 100% or fixed plasma current, by varying which beam sources are used, opening the possibility for feedback control of the current profile. In terms of quantities like collisionality, neutron emission, non-inductive fraction, or stored energy, these scenarios represent a significant performance extension compared with NSTX and other present spherical torii.

Overview of NSTX Upgrade initial results and modelling highlights

Abstract: Author(s): Menard, JE; Allain, JP; Battaglia, DJ; Bedoya, F; Bell, RE; Belova, E; Berkery, JW; Boyer, MD; Crocker, N; Diallo, A; Ebrahimi, F; Ferraro, N; Fredrickson, E; Frerichs, H; Gerhardt, S; Gorelenkov, N; Guttenfelder, W; Heidbrink, W; Kaita, R; Kaye, SM; Kriete, DM; Kubota, S; Leblanc, BP; Liu, D; Lunsford, R; Mueller, D; Myers, CE; Ono, M; Park, JK; Podesta, M; Raman, R; Reinke, M; Ren, Y; Sabbagh, SA; Schmitz, O; Scotti, F; Sechrest, Y; Skinner, CH; Smith, DR; Soukhanovskii, V; Stoltzfus-Dueck, T; Yuh, H; Wang, Z; Waters, I; Ahn, JW; Andre, R; Barchfeld, R; Beiersdorfer, P; Bertelli, N; Bhattacharjee, A; Brennan, D; Buttery, R; Capece, A; Canal, G; Canik, J; Chang, CS; Darrow, D; Delgado-Aparicio, L; Domier, C; Ethier, S; Evans, T; Ferron, J; Finkenthal, M; Fonck, R; Gan, K; Gates, D; Goumiri, I; Gray, T; Hosea, J; Humphreys, D; Jarboe, T; Jardin, S; Jaworski, MA; Koel, B; Kolemen, E; Ku, S; La Haye, RJ; Levinton, F; Luhmann, N; Maingi, R; Maqueda, R; McKee, G; Meier, E; Myra, J; Perkins, R | Abstract: 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 H98y,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. Lastly, 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.

Digital control of dynamic systems

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Chapter 6: Steady state operation

Abstract: Significant progress has been made in the area of advanced modes of operation that are candidates for achieving steady state conditions in a fusion reactor. The corresponding parameters, domain of operation, scenarios and integration issues of advanced scenarios are discussed in this chapter. A review of the presently developed scenarios, including discussions on operational space, is given. Significant progress has been made in the domain of heating and current drive in recent years, especially in the domain of off-axis current drive, which is essential for the achievement of the required current profile. The actuators for heating and current drive that are necessary to produce and control the advanced tokamak discharges are discussed, including modelling and predictions for ITER. The specific control issues for steady state operation are discussed, including the already existing experimental results as well as the various strategies and needs (qψ profile control and temperature gradients). Achievable parameters for the ITER steady state and hybrid scenarios with foreseen heating and current drive systems are discussed using modelling including actuators, allowing an assessment of achievable current profiles. Finally, a summary is given in the last section including outstanding issues and recommendations for further research and development.

Integrated modeling applications for tokamak experiments with OMFIT

Abstract: One modeling framework for integrated tasks (OMFIT) is a comprehensive integrated modeling framework which has been developed to enable physics codes to interact in complicated workflows, and support scientists at all stages of the modeling cycle. The OMFIT development follows a unique bottom-up approach, where the framework design and capabilities organically evolve to support progressive integration of the components that are required to accomplish physics goals of increasing complexity. OMFIT provides a workflow for easily generating full kinetic equilibrium reconstructions that are constrained by magnetic and motional Stark effect measurements, and kinetic profile information that includes fast-ion pressure modeled by a transport code. It was found that magnetic measurements can be used to quantify the amount of anomalous fast-ion diffusion that is present in DIII-D discharges, and provide an estimate that is consistent with what would be needed for transport simulations to match the measured neutron rates. OMFIT was used to streamline edge-stability analyses, and evaluate the effect of resonant magnetic perturbation (RMP) on the pedestal stability, which have been found to be consistent with the experimental observations. The development of a five-dimensional numerical fluid model for estimating the effects of the interaction between magnetohydrodynamic (MHD) and microturbulence, and its systematic verification against analytic models was also supported by the framework. OMFIT was used for optimizing an innovative high-harmonic fast wave system proposed for DIII-D. For a parallel refractive index , the conditions for strong electron-Landau damping were found to be independent of launched and poloidal angle. OMFIT has been the platform of choice for developing a neural-network based approach to efficiently perform a non-linear multivariate regression of local transport fluxes as a function of local dimensionless parameters. Transport predictions for thousands of DIII-D discharges showed excellent agreement with the power balance calculations across the whole plasma radius and over a broad range of operating regimes. Concerning predictive transport simulations, the framework made possible the design and automation of a workflow that enables self-consistent predictions of kinetic profiles and the plasma equilibrium. It is found that the feedback between the transport fluxes and plasma equilibrium can significantly affect the kinetic profiles predictions. Such a rich set of results provide tangible evidence of how bottom-up approaches can potentially provide a fast track to integrated modeling solutions that are functional, cost-effective, and in sync with the research effort of the community.

Progress of the CFETR design

Abstract: The Chinese Fusion Engineering Testing Reactor (CFETR), complementing the ITER facility, is aiming to demonstrate fusion energy production up to 200 MW initially and to eventually reach DEMO relevant power level 1 GW, to manifest a high duty factor of 0.3–0.5, and to pursue tritium self-sufficiency with tritium breeding ratio (TBR) >1. The key challenge to meet the missions of the CFETR is to run the machine in steady state (or long pulse) and high duty factor. By using a multi-dimensional code suite with physics-based models, self-consistent steady-state and hybrid mode scenarios for CFETR have been developed under a high magnetic field up to 6.5 T. The negative-ion neutral beam injection together with high frequency electron cyclotron wave and lower hybrid wave (and/or fast wave) are proposed to be used to drive the current. Subsequently the engineering design of CFETR including the magnet system, vacuum system, tritium breeding blanket, divertor, remote handling and maintenance system will be introduced. Some research and development (R&D) activities are also introduced in this paper.