V.S. Chan

The Spherical Tokamak Path to Fusion Power

AbstractThe low-aspect-ratio tokamak or spherical torus (ST) approach offers the two key elements needed to enable magnetic confinement fusion to make the transition from a government-funded research program to the commercial marketplace: a low-cost, low-power, small-size market entry vehicle and a strong economy of scale in larger devices. Within the ST concept, a very small device (A = 1.4, major radius ~1 m, similar size to the DIII-D tokamak) could be built that would produce ~800 MW(thermal), 200 MW(net electric) and would have a gain, defined as QPLANT = (gross electric power/recirculating power), of ~2. Such a device would have all the operating systems and features of a power plant and would therefore be acceptable as a pilot plant, even though the cost of electricity would not be competitive. The ratio of fusion power to copper toroidal field (TF) coil dissipation rises quickly with device size (like R3 to R4, depending on what is held constant) and can lead to 4-GW(thermal) power plants with QPL...

ELECTRON CYCLOTRON CURRENT DRIVE EFFICIENCY IN GENERAL TOKAMAK GEOMETRY

Green’s-function techniques are used to calculate electron cyclotron current drive (ECCD) efficiency in general tokamak geometry in the low-collisionality regime. Fully relativistic electron dynamics is employed in the theoretical formulation. The high-velocity collision model is used to model Coulomb collisions and a simplified quasilinear rf diffusion operator describes wave–particle interactions. The approximate analytic solutions which are benchmarked with a widely used ECCD model, facilitate time-dependent simulations of tokamak operational scenarios using the noninductive current drive of electron cyclotron waves.

Optimized profiles for improved confinement and stability in the DIII-D tokamak

Simultaneous achievement of high energy confinement, tau E, and high plasma beta, beta , leads to an economically attractive compact tokamak fusion reactor. High confinement enhancement, H= tau E/ tau E-ITER89P=4, and high normalized beta beta N beta /(I/aB)=6%-m-T/MA have been obtained in DIII-D experimental discharges. These improved confinement and/or improved stability limits are observed in several DIII-D high performance operational regimes: VH-mode, high li H-mode, second stable core, and high beta poloidal. We have identified several important features of the improved performance in these discharges: details of the plasma shape, toroidal rotation or E*B flow profile, q profile and current density profile, and pressure profile. From our improved physics understanding of these enhanced performance regimes, we have developed operational scenarios which maintain the essential features of the improved confinement and which increase the stability limits using localized current profile control. The stability limit is increased by modifying the interior safety factor profile to be nonmonotonic with high central q, while maintaining the edge current density consistent with the improved transport regimes and the high edge bootstrap current. We have calculated high beta equilibria with beta N=6.5, stable to ideal n=1 kinks and stable to ideal ballooning modes. The safety factor at the 95% flux surface is 6, the central q value is 3.9 and the minimum in q is 2.6.

Fokker-Planck simulations mylb of knock-on electron runaway avalanche and bursts in tokamaks

The avalanche of runaway electrons in an ohmic tokamak plasma triggered by knock-on collisions of traces of energetic electrons with the bulk electrons is simulated by the bounce averaged Fokker-Planck code, CQL3D. It is shown that even when the electric field is small for the production of Dreicer runaways, the knock-on collisions can produce significant runaway electrons in a fraction of a second at typical reactor parameters. The energy spectrum of these knock-on runaways has a characteristic temperature. The growth rate and temperature of the runaway distribution are determined and compared with theory. In simulations of pellet injection into high temperature plasmas, it is shown that a burst of Dreicer runaways may also occur depending on the cooling rate due to the pellet injection. Implications of these phenomena on disruption control in reactor plasmas are discussed.

Fusion Nuclear Science Facility Candidates

Abstract To move to a fusion DEMO power plant after ITER, a Fusion Nuclear Science Facility (FNSF) is needed in addition to ITER and research in operating tokamaks and those under construction. The FNSF will enable research on how to utilize and deal with the products of fusion reactions, addressing such issues as how to extract the energy from neutrons and alpha particles into high-temperature process heat streams to be either used directly or converted to electricity, how to make tritium from the neutrons and lithium, how to deal with the effects of the neutrons on the blanket structures, and how to manage the first wall surface erosion caused by the alpha particle heat appearing as low-energy plasma fluxes to those surfaces. Two candidates for the FNSF are considered in this paper: normal and low aspect ratio copper magnet tokamaks. The methods of selecting optimum machine design points versus aspect ratio are fully presented. The two options are compared and contrasted; both options appear viable.

Generation of plasma rotation in a tokamak by ion-cyclotron absorption of fast Alfvén waves

A mechanism is proposed and evaluated for driving rotation in tokamak plasmas by minority ion-cyclotron heating, even though this heating introduces negligible angular momentum. The mechanism has two elements: First, angular momentum transport is governed by a diffusion equation with a boundary condition at the separatrix. Second, Monte Carlo calculations show that ion-cyclotron energized particles will provide a torque density source which has a zero volume integral but separated positive and negative regions. With such a source, a solution of the diffusion equation predicts that ion-cyclotron heating will cause a rotational shear layer to develop. The corresponding jump in plasma rotation ΔΩ is found to be negative outwards when the ion-cyclotron surface lies on the low-field side of the magnetic axis and positive outwards with the resonance on the high-field side. The magnitude of the jump ΔΩ=(4qmaxWJ2*) (eBR3a2ne(2π)2)−1(τM/τE) where |J2*|≈2–4 is a nondimensional rotation frequency calculated by the M...

Theory of fast wave current drive for tokamak plasmas

The paper presents calculations of the efficiency of fast wave current drive at reactor-like densities and temperatures, including toroidal effects. Accessibility and competitive absorption mechanisms are estimated. Two bands of frequencies are found to be of interest for reactor applications – one in the ion cyclotron range of frequencies and the other at higher harmonics but below the lower hybrid frequency.

Evaluation of CFETR as a Fusion Nuclear Science Facility using multiple system codes

This paper presents the results of a multi-system codes benchmarking study of the recently published China Fusion Engineering Test Reactor (CFETR) pre-conceptual design (Wan et al 2014 IEEE Trans. Plasma Sci. 42 495). Two system codes, General Atomics System Code (GASC) and Tokamak Energy System Code (TESC), using different methodologies to arrive at CFETR performance parameters under the same CFETR constraints show that the correlation between the physics performance and the fusion performance is consistent, and the computed parameters are in good agreement. Optimization of the first wall surface for tritium breeding and the minimization of the machine size are highly compatible. Variations of the plasma currents and profiles lead to changes in the required normalized physics performance, however, they do not significantly affect the optimized size of the machine. GASC and TESC have also been used to explore a lower aspect ratio, larger volume plasma taking advantage of the engineering flexibility in the CFETR design. Assuming the ITER steady-state scenario physics, the larger plasma together with a moderately higher BT and Ip can result in a high gain Qfus ~ 12, Pfus ~ 1 GW machine approaching DEMO-like performance. It is concluded that the CFETR baseline mode can meet the minimum goal of the Fusion Nuclear Science Facility (FNSF) mission and advanced physics will enable it to address comprehensively the outstanding critical technology gaps on the path to a demonstration reactor (DEMO). Before proceeding with CFETR construction steady-state operation has to be demonstrated, further development is needed to solve the divertor heat load issue, and blankets have to be designed with tritium breeding ratio (TBR) >1 as a target.

Modelling of feedback and rotation stabilization of the resistive wall mode in tokamaks

This paper describes the modelling of the feedback control and rotational stabilization of the resistive wall mode (RWM) in tokamaks. A normal mode theory for the feedback stabilization of the RWM has been developed for an ideal plasma with no toroidal rotation. This theory has been numerically implemented for general tokamak geometry and applied to the DIII-D tokamak. A general formulation is further developed for the feedback stabilization of a tokamak with toroidal rotation and plasma dissipation. It has been used to understand the role of the external resonant field in affecting the plasma stability and compared with the resonant field amplification phenomenon observed in DIII-D. The effectiveness of a differentially rotating resistive wall in stabilizing the RWM has also been studied numerically. It is found that for a non-circular tokamak, a wide range of flow patterns are all effective. The structure of the RWM predicted from ideal MHD theory has been compared with signals from various diagnostics. It is also projected that based on DIII-D results scaled up to the ITER-FEAT, 33 MW of 1 MeV negative neutral beam injection will be able to sustain plasma rotation sufficient to stabilize the RWM without relying on feedback.

Physics basis for the advanced tokamak fusion power plant, ARIES-AT

Abstract The advanced tokamak is considered as the basis for a fusion power plant. The ARIES-AT design has an aspect ratio of A ≡ R / a = 4.0 , an elongation and triangularity of κ = 2.20 , δ = 0.90 (evaluated at the separatrix surface), a toroidal beta of β = 9.1 % (normalized to the vacuum toroidal field at the plasma center), which corresponds to a normalized beta of β N ≡ 100 × β / ( I P ( M A ) / a ( m ) B ( T ) ) = 5.4 . These beta values are chosen to be 10% below the ideal MHD stability limit. The bootstrap-current fraction is f BS ≡ I BS / I P = 0.91 . This leads to a design with total plasma current I P = 12.8  MA, and toroidal field of 11.1 T (at the coil edge) and 5.8 T (at the plasma center). The major and minor radii are 5.2 and 1.3 m. The effects of H-mode edge gradients and the stability of this configuration to non-ideal modes is analyzed. The current drive system consists of ICRF/FW for on-axis current drive and a Lower Hybrid system for off-axis. Transport projections are presented using the drift-wave based GLF23 model. The approach to power and particle exhaust using both plasma core and scrape-off-layer radiation is presented.