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Overview of results from the National Spherical Torus Experiment (NSTX)

机译：美国国家球形圆环实验（NSTX）的结果概述

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The mission of the National Spherical Torus Experiment (NSTX) is the demonstration of the physics basis required to extrapolate to the next steps for the spherical torus (ST), such as a plasma facing component test facility (NHTX) or an ST based component test facility (ST-CTF), and to support ITER. Key issues for the ST are transport, and steady state high β operation. To better understand electron transport, a new high-k scattering diagnostic was used extensively to investigate electron gyro-scale fluctuations with varying electron temperature gradient scale length. Results from n = 3 braking studies are consistent with the flow shear dependence of ion transport. New results from electron Bernstein wave emission measurements from plasmas with lithium wall coating applied indicate transmission efficiencies near 70% in H-mode as a result of reduced collisionality. Improved coupling of high harmonic fast-waves has been achieved by reducing the edge density relative to the critical density for surface wave coupling. In order to achieve high bootstrap current fraction, future ST designs envision running at very high elongation. Plasmas have been maintained on NSTX at very low internal inductance l_i ～ 0.4 with strong shaping (k ～ 2.7, δ ～ 0.8) with β_N approaching the with-wall β-limit for several energy confinement times. By operating at lower collisionality in this regime, NSTX has achieved record non-inductive current drive fraction f_(NI) ～ 71%. Instabilities driven by super-Alfvenic ions will be an important issue for all burning plasmas, including ITER. Fast ions from NBI on NSTX are super-Alfvenic. Linear toroidal Alfven eigenmode thresholds and appreciable fast ion loss during multi-mode bursts are measured and these results are compared with theory. The impact of n > 1 error fields on stability is an important result for ITER. Resistive wall mode/resonant field amplification feedback combined with n = 3 error field control was used on NSTX to maintain plasma rotation with β above the no-wall limit. Other highlights are results of lithium coating experiments, momentum confinement studies, scrape-off layer width scaling, demonstration of divertor heat load mitigation in strongly shaped plasmas and coupling of coaxial helicity injection plasmas to ohmic heating ramp-up. These results advance the ST towards next step fusion energy devices such as NHTX and ST-CTF.
机译：国家球形圆环实验（NSTX）的任务是证明推断球形圆环（ST）下一步所需的物理基础，例如面向等离子体的组件测试设备（NHTX）或基于ST的组件测试设施（ST-CTF），并支持ITER。 ST的关键问题是运输和稳态高β运行。为了更好地理解电子传输，广泛使用了一种新型的高k散射诊断程序来研究随着电子温度梯度标度长度的变化而产生的电子陀螺标度波动。 n = 3次制动研究的结果与离子传输的流动剪切相关性一致。施加锂壁涂层的等离子体在电子伯恩斯坦波发射测量中获得的新结果表明，由于降低了碰撞性，H模式下的传输效率接近70％。通过相对于表面波耦合的临界密度降低边缘密度，已实现了高谐波快速波的改进耦合。为了实现高自举电流分数，未来的ST设计设想以很高的伸长率运行。 NSTX上的等离子体保持在非常低的内部电感l_i〜0.4时，具有很强的整形（k〜2.7，δ〜0.8），β_N接近壁的β极限数次。通过在这种情况下以较低的碰撞性运行，NSTX达到了创纪录的无感电流驱动比f_（NI）〜71％。对于所有燃烧的等离子体（包括ITER），由超级Alf离子驱动的不稳定性将是一个重要的问题。 NSTX上来自NBI的快速离子是超级Alfvenic离子。测量了线性环形Alfven本征模式阈值和多模式猝发过程中明显的快速离子损失，并将这些结果与理论进行了比较。对于ITER，n> 1误差场对稳定性的影响是重要的结果。在NSTX上使用电阻壁模式/共振场放大反馈与n = 3误差场控制相结合，以使血浆旋转保持β超过无壁极限。其他亮点还包括锂涂层实验，动量限制研究，刮除层宽度缩放，在强形等离子体中减轻偏滤器热负荷以及同轴螺旋注入等离子体与欧姆加热加速耦合的结果。这些结果使ST迈向了下一步的融合能源设备，例如NHTX和ST-CTF。

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