M.G. Bell

TSC simulation of ohmic discharges in TFTR

The Tokamak Simulation Code (TSC) has been used to model the time dependence of several Ohmic discharges in the TFTR experiment. The semi-empirical thermal conductivity model and the sawtooth model in TSC have been refined so that good agreement between the simulation and the experiment is obtained in the electron and ion temperature profiles and in the current profiles for the entire duration of the discharges. Neoclassical resistivity gives good agreement with the measured surface voltage and the rate of poloidal flux consumption

Simulations of alpha parameters in a TFTR DT supershot with high fusion power

A TFTR supershot with a plasma current of 2.5 MA, a neutral beam heating power of 33.7 MW and a peak DT fusion power of 7.5 MW is studied using the TRANSP plasma analysis code. Simulations of alpha parameters such as the alpha heating, pressure and distributions in energy and v1/v are given. The effects of toroidal ripple and mixing of the fast alpha particles during the sawteeth observed after the neutral beam injection phase are modelled. The distributions of alpha particles on the outer midplane are peaked near forward and backward v1/v. Ripple losses deplete the distributions in the vicinity of v1/v=-0.2. Sawtooth mixing of fast alpha particles is computed to reduce their central density and broaden their width in energy

Snowflake divertor configuration studies in National Spherical Torus Experimenta)

Experimental results from NSTX indicate that the snowflake divertor (D. Ryutov, Phys. Plasmas 14, 064502 (2007)) may be a viable solution for outstanding tokamak plasma-material interface issues. Steady-state handling of divertor heat flux and divertor plate erosion remains to be critical issues for ITER and future concept devices based on conventional and spherical tokamak geometry with high power density divertors. Experiments conducted in 4–6u2009MW NBI-heated H-mode plasmas in NSTX demonstrated that the snowflake divertor is compatible with high-confinement core plasma operation, while being very effective in steady-state divertor heat flux mitigation and impurity reduction. A steady-state snowflake divertor was obtained in recent NSTX experiments for up to 600u2009ms using three divertor magnetic coils. The high magnetic flux expansion region of the scrape-off layer (SOL) spanning up to 50% of the SOL width λq was partially detached in the snowflake divertor. In the detached zone, the heat flux profile flatt...

Beta-limiting instabilities and global mode stabilization in the National Spherical Torus Experiment

Research on the stability of spherical torus plasmas at and above the no-wall beta limit is being addressed on the National Spherical Torus Experiment [M. Ono et al., Nucl. Fusion 40, 557 (2000)], that has produced low aspect ratio plasmas, R/a∼1.27 at plasma current exceeding 1.4 MA with high energy confinement (TauE/TauE_ITER89P>2). Toroidal and normalized beta have exceeded 25% and 4.3, respectively, in q∼7 plasmas. The beta limit is observed to increase and then saturate with increasing li. The stability factor βN/li has reached 6, limited by sudden beta collapses. Increased pressure peaking leads to a decrease in βN. Ideal stability analysis of equilibria reconstructed with EFIT [L. L. Lao et al., Nucl. Fusion 25, 1611 (1985)] shows that the plasmas are at the no-wall beta limit for the n=1 kink/ballooning mode. Low aspect ratio and high edge q theoretically alter the plasma stability and mode structure compared to standard tokamak configurations. Below the no-wall limit, stability calculations show ...

Plasma shape control on the National Spherical Torus Experiment (NSTX) using real-time equilibrium reconstruction

Plasma shape control using real-time equilibrium reconstruction has been implemented on the National Spherical Torus Experiment (NSTX). The rtEFIT code originally developed for use on DIII-D was adapted for use on NSTX. The real-time equilibria provide calculations of the flux at points on the plasma boundary, which are used as input to a shape control algorithm known as isoflux control. The flux at the desired boundary location is compared with a reference flux value, and this flux error is used as the basic feedback quantity for the poloidal field coils on NSTX. The hardware that comprises the control system is described, as well as the software infrastructure. Examples of precise boundary control are also presented.

β limit disruptions in the Tokamak Fusion Test Reactor

A disruptive β limit (β=plasma pressure/magnetic pressure) is observed in high‐performance plasmas in the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Plasma Phys. Controlled Nuclear Fusion 1, 421 (1987)]. The magnetohydrodynamic character of these disruptions differs substantially from the disruptions in high‐density plasmas (density limit disruptions) on TFTR. The high β disruptions can occur with less than a millisecond warning in the form of a fast growing precursor. The precursor appears to be an n=1 kink strongly coupled through finite β effects and toroidal terms to higher m components. It does not have the ‘‘cold bubble’’ structure found in density limit disruptions. The n=1 kink, in turn, appears to excite a ballooning‐type mode that may contribute to the thermal quench.

Recent progress in the NSTX/NSTX-U lithium programme and prospects for reactor-relevant liquid-lithium based divertor development

Developing a reactor-compatible divertor has been identified as a particularly challenging technology problem for magnetic confinement fusion. Application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and other plasma performance benefits. During the 2010 NSTX campaign, application of a relatively modest amount of Li (300xa0mg prior to the discharge) resulted in a ∼50% reduction in heat load on the liquid lithium divertor (LLD) attributable to enhanced divertor bolometric radiation. These promising Li results in NSTX and related modelling calculations motivated the radiative LLD concept proposed here. Li is evaporated from the liquid lithium (LL) coated divertor strike-point surface due to the intense heat flux. The evaporated Li is readily ionized by the plasma due to its low ionization energy, and the poor Li particle confinement near the divertor plate enables ionized Li ions to radiate strongly, resulting in a significant reduction in the divertor heat flux. This radiative process has the desired effect of spreading the localized divertor heat load to the rest of the divertor chamber wall surfaces, facilitating the divertor heat removal. The LL coating of divertor surfaces can also provide a ‘sacrificial’ protective layer to protect the substrate solid material from transient high heat flux such as the ones caused by the edge localized modes. By operating at lower temperature than the first wall, the LL covered large divertor chamber wall surfaces can serve as an effective particle pump for the entire reactor chamber, as impurities generally migrate towards lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity (e.g., ∼1xa0lxa0s−1 for ∼1% level ‘impurities’) is envisioned for a steady-state 1xa0GW-electric class fusion power plant.

Overview of experiments with radiation cooling at high confinement and high density in limited and diverted discharges

An overview is presented of recent experiments with radiating mantles on limiter and divertor machines, realizing simultaneously high confinement and high density at high-radiation levels. A variety of operational regimes has been observed and the characteristics of each are documented. High-performance plasmas (i.e. edge localized mode (ELM)-free H-mode confinement quality and normalized beta values simultaneously) with radiating mantles have been demonstrated under quasistationary conditions during the maximum flattop time of the machine (equal to tens of confinement times) on DIII-D and TEXTOR-94. Maximum values for up to 4 and for the advanced tokamak confinement-stability product up to 13, have been obtained in very high confinement mode (VH-mode) like discharges with radiating mantles in DIII-D. There is a striking similarity between improved ohmic confinement discharges (with or without Ne seeding) and radiating mantle discharges, indicating a possible common origin for the confinement improvement observed. Possible scenarios for the application of radiating mantles on larger machines such as JET and JT-60U are indicated.

Neoclassical tearing modes in Tokamak Fusion Test Reactor experiments. I. Measurements of magnetic islands and Δ

Tearing-type modes are observed in most high confinement operation regimes in the Tokamak Fusion Test Reactor (TFTR) [Nucl. Fusion 35, 1429 (1995)]. Three different methods are used to measure the magnetic island widths: external magnetic coils, internal temperature fluctuation from electron cyclotron emission (ECE) diagnostics and an experiment where the plasma major radius is rapidly shifted (“Jog” experiments). A good agreement between the three methods is observed. Numerical and analytic calculations of Δ′ (the tearing instability index) are compared with an experimental measurement of Δ′ using the tearing mode eigenfunction mapped from the jog data. The obtained negative Δ′ indicates that the observed tearing modes cannot be explained by the classical current-gradient-driven tearing theory.

New capabilities and results for the National Spherical Torus Experiment

The National Spherical Torus Experiment (NSTX) produces plasmas with toroidal aspect ratio as low as 1.25, which can be heated by up to 6 MW high-harmonic fast waves and up to 7 MW of deuterium neutral beam injection. Using new poloidal field coils, plasmas with cross-section elongation up to 2.7, triangularity 0.8, plasma currents Ip up to 1.5 MA and normalized currents Ip/aBT up to 7.5 MA/mT have been achieved. A significant extension of the plasma pulse length, to 1.5 s at a plasma current of 0.7 MA, has been achieved by exploiting the bootstrap and NBI-driven currents to reduce the dissipation of poloidal flux. Inductive plasma startup has been supplemented by coaxial helicity injection (CHI) and the production of persistent current on closed flux surfaces by CHI has now been demonstrated in NSTX. The plasma response to magnetic field perturbations with toroidal mode numbers n = 1 or 3 and the effects on the plasma rotation have been investigated using three pairs of coils outside the vacuum vessel. Recent studies of both MHD stability and of transport benefitted from improved diagnostics, including measurements of the internal poloidal field using the motional Stark effect (MSE). In plasmas with a region of reversed magnetic shear in the core, now confirmed by the MSE data, improved electron confinement has been observed.