Session UI1: Plasma Technology: Magnetic Confinement

Human Outreach through Physics

In this talk unique methods for human outreach through physics are described. The focus is on identifying young talented researchers and colleagues around the globe and nourish them for the purpose of diffusing physics knowledge. The goal can be achieved through the organization of international conferences, workshops, seminars, and colleagues, at different locations, invite young and experienced researchers to those meetings, invite them to your home institution, in addition to visiting their universities/laboratories for mentoring and exchanging physics knowledge. The scientific part shall deal with collective processes and coherent nonlinear effects in space and laboratory plasmas.   

Advanced density profile reflectometry; the state-of-the-art and measurement prospects for ITER

Dramatic progress in millimeter-wave technology has allowed the realization of a key goal for ITER diagnostics, the routine measurement of the plasma density profile from millimeter-wave radar (reflectometry) measurements. In reflectometry, the measured round-trip group delay of a probe beam reflected from a plasma cutoff is used to infer the density distribution in the plasma. Reflectometer systems implemented by UCLA on a number of devices employ frequency-modulated continuous-wave (FM-CW), ultrawide-bandwidth, high-resolution radar systems. One such system on DIII-D has routinely demonstrated measurements of the density profile over a range of electron density of 0-6.4$\times$10$^{19}\,$ m$^{-3}$, with $\sim$25~$\mu$s time and $\sim$4~mm radial resolution, meeting key ITER requirements. This progress in performance was made possible by multiple advances in the areas of millimeter-wave technology, novel measurement techniques, and improved understanding, including: (i)~fast sweep, solid-state, wide bandwidth sources and power amplifiers, (ii)~dual polarization measurements to expand the density range, (iii)~adaptive radar-based data analysis with parallel processing on a Unix cluster, (iv)~high memory depth data acquisition, and (v)~advances in full wave code modeling. The benefits of advanced system performance will be illustrated using measurements from a wide range of phenomena, including ELM and fast-ion driven mode dynamics, L-H transition studies and plasma-wall interaction. The measurement capabilities demonstrated by these systems provide a design basis for the development of the main ITER profile reflectometer system. This talk will explore the extent to which these reflectometer system designs, results and experience can be translated to ITER, and will identify what new studies and experimental tests are essential.   

New Design Methods for Magnetic Flux Loop Arrays in the NCSX Experiment

Magnetic pickup loops on the vacuum vessel (VV) can provide an abundance of equilibrium information for stellarators. A substantial effort has gone into designing flux loops for NCSX, a 3-field period quasi-axisymmetric stellarator under construction at PPPL. The design philosophy, to measure all of the magnetic field distributions normal to the VV that can be measured, has necessitated the development of Singular Value Decomposition algorithms for identifying efficient loop locations. The fields are expected to be predominantly stellarator symmetric (SS) - the symmetry of the machine design - with toroidal mode numbers per torus, n, equal to a multiple of 3 and possessing reflection symmetry in a period. However, plasma instabilities and coil imperfections will generate non-SS fields which must also be diagnosed. The measured symmetric fields will yield important information on the plasma current and pressure profile as well as on the plasma shape. All fields that obey the design symmetries could be measured by placing flux loops in a single half-period of the VV, but accurate resolution of non-symmetric modes, quantified by the condition number of a matrix, requires re-positioning loops to equivalent locations on the full torus. A sub-array of loops located along the inside wall of the vertically elongated cross-section was designed to detect n=3 m=5 or 6 resonant field perturbations that can cause important islands. Additional sub-arrays included are continuous in the toroidal and poloidal directions. Loops are also placed at symmetry points of the VV to obtain maximal sensitivity to asymmetric perturbations. Combining results from various calculations which have made extensive use of a database of 2500 free-boundary VMEC equilibria, has led to the choice of 227 flux loops for NCSX, of which 151 have distinct shapes.   

Ignited Spherical Tokamaks as a Reactor Development Facility

The talk presents the concept of Ignited Spherical Tokamaks (IST), which can serve as a neutron fusion source for the Reactor Development Facility. The IST would be uniquely consistent with three objectives of magnetic fusion, i.e., (a) high power density plasma regime (5-10 MW/m$^3$), and high neutron flux (b) for designing the ``first wall'' of the reactor (up to the fluence of 15 MW year/m$^2$), and (c) for developing the tritium cycle. The lithium-based plasma facing components (LiWalls) of an IST provide the pumping boundary conditions for the plasma. When combined with central fueling of the plasma by low energy (E=45-50 keV) neutral beam injection (NBI), the LiWall environment leads to a flat plasma temperature T=E/3. This results in a super-critical ignition regime, with ion-temperature gradient turbulence eliminated, when the energy confinement is close to neo-classical, while the high current density at the separatrix robustly stabilizes the edge-localized modes. Unlike conventional approach to magnetic fusion, the super-critical ignition regime relies on core fueling by NBI and fast expulsion of the $\alpha$-particles, rather than on their heating the plasma. In this regards, the IST configuration (for the neutron source purposes) and stellarators (as power reactors), rather than tokamaks, have similarity regarding super-critical ignition regime. A separate national program ($\simeq$\$2-2.5 B for $\simeq$ 15 years) can realistically develop an Ignited Spherical Tokamak as a fusion neutron source for reactor R\&D in 3 steps (two with DD, and one with DT plasmas), i.e., \begin{enumerate} \item A spherical tokamak, targeting achievement of absorbing wall regime with neo-classical confinement in a DD plasma and $Q_{DT-equiv}=1$, \item Full scale DD-prototype of IST for demonstration of all aspects of stationary super-critical regime with $Q_{DT-equiv}\simeq50$. \item IST itself with a DT plasma and $Q_{DT}\simeq50$ for reactor technology and $\alpha$-particle power extraction studies. \end{enumerate}   

FTU experimental results using a lithium liquid limiter

For the first time in a medium size tokamak as FTU a liquid lithium limiter has acted as main limiter. The main goal of the experiment aims to test the capillary porous system (CPS) configuration. The experiment has two main aspects: a technological one related to the use of a liquid metal such as the capability to control high heat loads and to ensure the self-regeneration of plasma facing components.The other, more physical, to test the CPS LL behaviour and to use it like a conditioning system to deposit a lithium film on the chamber walls In this first experimental campaign the LL has been tested in ohmic plasma discharges with B$_{T }$= 6T, Ip = 0.5- 0.9 MA and average electron density n$_{e}$ from 0.15 up to 2.6x10$^{20}$m$^{-3}$. The experimental results confirm the strong capability of lithium to pump hydrogen particles allowing to extend the range of plasma operations to the lowest electron density never reached on FTU (1.5x10$^{19}$m$^{-3})$. As consequences of this pumping effect we have measured significant modifications of the scrape off layer, mainly of electron temperature.Furthermore reductions in the total radiated power and in the loop voltage have been observed due to low Z$_{eff}$ values.After litization, Z$_{eff}$ is equal to 1.5 at low density (.5x10$^{20}$m$^{-3})$ and 1. at higher density (1x10$^{20}$m$^{-3})$.The behaviour of the lithium limiter as first wall material has been successfully tested for thermal loads in the range 1-10 MW/m$^{2}$.Thermal analysis and modeling will be discussed. \newline \newline Co-authors: M.L. Apicella, M.Marinucci, C. Mazzotta, V.Pericoli Ridolfini, O.Tudisco, R. Zagorski, V. Lazarev, A. Vertkov, A. Alekseyev.   

On Heat Loading, Novel Divertors, and Fusion Reactors

A new magnetic divertor geometry has been proposed to solve reactor heat exhaust problems, which are far more severe for a reactor than for ITER. Using reactor-compatible coils to generate an extra X-point downstream from the main X-point, the new X-divertor (XD) is shown to greatly expand magnetic flux at the divertor plates. As a result, the heat is distributed over a larger area and the line length is greatly increased. The heat-flux limitations of a standard divertor (SD) force a high core radiation fraction ($f_{Rad}$) in most reactor designs that necessarily have a several times higher ratio of heating power to radius (P/R) than ITER. It is argued that such high values of $f_{Rad}$ will probably have serious deleterious consequences on the core confinement and stability of a burning plasma. Operation with internal transport barriers (ITBs) does not appear to overcome this problem. By reducing the core $f_{Rad}$ within an acceptable range, the X-divertor is shown to substantially lower the core confinement requirement for a fusion reactor. As a bonus, the XD also enables the use of liquid metals by reducing the MHD drag. A possible series of experiments for an efficient and attractive path to practical fusion power is suggested.