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Lee, J. P.

Publications and source records attributed to Lee, J. P..

A sustained high-temperature fusion plasma regime facilitated by fast ions

Nuclear fusion is one of the most attractive alternatives to carbon-dependent energy sources. Harnessing energy from nuclear fusion in a large reactor scale, however, still presents many scientific challenges despite the many years of research and steady advances in magnetic confinement approaches. State-of-the-art magnetic fusion devices cannot yet achieve a sustainable fusion performance, which requires a high temperature above 100 million kelvin and sufficient control of instabilities to ensure steady-state operation on the order of tens of seconds. Here, in this study, we report experiments at the Korea Superconducting Tokamak Advanced Research device producing a plasma fusion regime that satisfies most of the above requirements: thanks to abundant fast ions stabilizing the core plasma turbulence, we generate plasmas at a temperature of 100 million kelvin lasting up to 20 seconds without plasma edge instabilities or impurity accumulation. A low plasma density combined with a moderate input power for operation is key to establishing this regime by preserving a high fraction of fast ions. This regime is rarely subject to disruption and can be sustained reliably even without a sophisticated control, and thus represents a promising path towards commercial fusion reactors.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

HIS-TORIC: Extending core ICRF wave simulation to include realistic SOL plasmas

This paper presents a novel approach to incorporating an arbitrarily shaped edge scrape-off-layer (SOL) plasma and an ion cyclotron range of frequency (ICRF) antenna structure into existing core ICRF wave simulation models. We partition the entire computation domain into two sub-domains: a core and an edge region. Simulations in each domain are performed separately with appropriate numerical solvers. For the core, the TORIC ICRF solver [M. Brambilla, Plasma Phys. Contrl. Fusion 41, 1 (1999)] was modified to impose an essential (Dirichlet) boundary condition at its interface with the edge domain. In the edge, a finite element method (FEM) is used to solve a cold collisional plasma model. The domains are then joined together using the continuity boundary condition for the tangential electric and magnetic fields at their interfaces (Hybrid Integration of SOL to TORIC: HIS-TORIC). The model developed here was tested using an ICRH H minority heating scenario on the Alcator C-Mod tokamak [I. Hutchinson et. al., Phys. Plasmas 1, 1511 (1994)]. The simulated pattern of core wave propagation agrees well with a standard TORIC simulation. This approach opens the possibility of using a realistic diverted SOL plasma and a complicated 3D RF antenna together with a rigorous hot core plasma model, while requiring only minimal modification to existing RF codes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

HIS-TORIC: extending core ICRF wave simulation to include realistic SOL plasmas

Here, this paper presents a novel approach to incorporating an arbitrarily shaped edge scrape-off-layer (SOL) plasma and an ion cyclotron range of frequency (ICRF) antenna structure into existing core ICRF wave simulation models. We partition the entire computation domain into two sub-domains: a core and an edge region. Simulations in each domain are performed separately with appropriate numerical solvers. For the core, the TORIC ICRF solver (Brambilla 1999 Plasma Phys. Control. Fusion 41 1) was modified to impose an essential (Dirichlet) boundary condition at its interface with the edge domain. In the edge, a finite element method is used to solve a cold collisional plasma model. The domains are then joined together using the continuity boundary condition for the tangential electric and magnetic fields at their interfaces (Hybrid Integration of SOL to TORIC: HIS-TORIC). The model developed here was tested using an ICRH H minority heating scenario on the Alcator C-Mod tokamak (Hutchinson et al 1994 Phys. Plasmas 1 1511). The simulated pattern of core wave propagation agrees well with a standard TORIC simulation. This approach opens the possibility of using a realistic diverted SOL plasma and a complicated 3D RF antenna together with a rigorous hot core plasma model, while requiring only minimal modification to existing RF codes.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Full-wave simulations of ICRF heating regimes in toroidal plasma with non-Maxwellian distribution functions

At the power levels required for significant heating and current drive in magnetically-confined toroidal plasma, modification of the particle distribution function from a Maxwellian shape is likely (Stix 1975 Nucl. Fusion 15 737), with consequent changes in wave propagation and in the location and amount of absorption. In order to study these effects computationally, both the finite-Larmor-radius and the high-harmonic fast wave (HHFW), versions of the full-wave, hot-plasma toroidal simulation code TORIC (Brambilla 1999 Plasma Phys. Control. Fusion 41 1 and Brambilla 2002 Plasma Phys. Control. Fusion 44 2423), have been extended to allow the prescription of arbitrary velocity distributions of the form f(v(parallel to), v(perpendicular to) , psi, theta). For hydrogen (H) minority heating of a deuterium (D) plasma with anisotropic Maxwellian H distributions, the fractional H absorption varies significantly with changes in parallel temperature but is essentially independent of perpendicular temperature. On the other hand, for HHFW regime with anisotropic Maxwellian fast ion distribution, the fractional beam ion absorption varies mainly with changes in the perpendicular temperature. The evaluation of the wave-field and power absorption, through the full wave solver, with the ion distribution function provided by either a Monte-Carlo particle and Fokker-Planck codes is also examined for Alcator C-Mod and NSTX plasmas. Non-Maxwellian effects generally tend to increase the absorption with respect to the equivalent Maxwellian distribution.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Effects of the q profile on toroidal rotation in Alcator C-Mod LHCD plasmas

Changes in the core toroidal rotation profiles following injection of lower hybrid (LH) waves have been documented in Alcator C-Mod plasmas. Shot by shot scans of LH input power have been performed at fixed magnetic field and electron density for several plasma currents. For sawtoothing target plasmas, if the input power is low enough that the central safety factor q 0 remains below 1, the change in the core rotation is in the counter-current direction, consistent in sign, magnitude and LH power scaling with direct momentum input from the LH waves. If the power level is high enough that there are significant changes to the q profile, including the termination of sawtooth oscillations, the change in the toroidal rotation is in the co-current direction, consistent with changes in the momentum flux through its dependence on the current density profile. Further, the direction of the rotation changes depends on whether q 0 is below or above unity, and seemingly not on the magnetic shear, nor the Ohmic confinement regime of the target plasma.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Tokamak elongation – how much is too much? Part 1. Theory

In this and the accompanying paper, the problem of the maximally achievable elongation κ in a tokamak is investigated. The work represents an extension of many earlier studies, which were often focused on determining κ limits due to (i) natural elongation in a simple applied pure vertical field or (ii) axisymmetric stability in the presence of a perfectly conducting wall. The extension investigated here includes the effect of the vertical stability feedback system which actually sets the maximum practical elongation limit in a real experiment. A basic resistive wall stability parameter, γτ w , is introduced to model the feedback system which although simple in appearance actually captures the essence of the feedback system. Elongation limits in the presence of feedback are then determined by calculating the maximum κ against n=0 resistive wall modes for fixed γτ w . The results are obtained by means of a general formulation culminating in a variational principle which is particularly amenable to numerical analysis. The principle is valid for arbitrary profiles but simplifies significantly for the Solov’ev profiles, effectively reducing the 2-D stability problem into a 1-D problem. Finally, the accompanying paper provides the numerical results and leads to a sharp answer of ‘how much elongation is too much’?

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Tokamak elongation – how much is too much? Part 2. Numerical results

We report the analytic theory presented in Paper I is converted into a form convenient for numerical analysis. A fast and accurate code has been written using this numerical formulation. The results are presented by first defining a reference set of physical parameters based on experimental data from high performance discharges. Scaling relations of maximum achievable elongation (κ max ) versus inverse aspect ratio (ε) are obtained numerically for various values of poloidal beta (β p ), wall radius (b/a) and feedback capability parameter (γ τ w ) in ranges near the reference values. It is also shown that each value of κ max occurs at a corresponding value of optimized triangularity (δ), whose scaling is also determined as a function of ε. The results show that the theoretical predictions of κ max are slightly higher than experimental observations for high performance discharges, as measured by high average pressure. The theoretical δ values are noticeably lower. We suggest that the explanation is associated with the observation that high performance involves not only n = 0 MHD stability, but also n ≥ 1 MHD modes described by β N in the Troyon limit and transport as characterized by τ E . Operation away from the n = 0 MHD optimum may still lead to higher performance if there are more than compensatory gains in β N and τ E . Unfortunately, while the empirical scaling of β N and τ E with the elongation (κ) has been determined, the dependence on δ has still not been quantified. This information is needed in order to perform more accurate overall optimizations in future experimental designs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Quantum photonics hybrid integration platform

Fundamental to integrated photonic quantum computing is an on-chip method for routing and modulating quantum light emission. We demonstrate a hybrid integration platform consisting of arbitrarily designed waveguide circuits and single-photon sources. InAs quantum dots (QD) embedded in GaAs are bonded to a SiON waveguide chip such that the QD emission is coupled to the waveguide mode. The waveguides are SiON core embedded in a SiO{sub 2} cladding. A tuneable Mach Zehnder interferometer (MZI) modulates the emission between two output ports and can act as a path-encoded qubit preparation device. The single-photon nature of the emission was verified using the on-chip MZI as a beamsplitter in a Hanbury Brown and Twiss measurement.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

Virtual environment architecture for rapid application development

We describe the MITRE Virtual Environment Architecture (VEA), a product of nearly two years of investigations and prototypes of virtual environment technology. This paper discusses the requirements for rapid prototyping, and an architecture we are developing to support virtual environment construction. VEA supports rapid application development by providing a variety of pre-built modules that can be reconfigured for each application session. The modules supply interfaces for several types of interactive I/O devices, in addition to large-screen or head-mounted displays.

Grinstein, Georges G.↗