Experiments and simulations to achieve high values of plasma parameters at the Globus-M spherical tokamak (ST) at moderate absolute auxiliary heating power (up to 0.8 MW) and high specific heating power (up to 2–3 MW m−3) are described. Important distinguishing features are the low edge safety factor range, which is unusual for STs, 2.7 < q < 5 and small plasma–outer wall space (3–5 cm). High ion heating efficiency with neutral beam injection (NBI) was demonstrated. Results of numerical simulation of fast ion trajectories are described and fast ion generation during the NBI and ion cyclotron resonance heating is discussed. Investigations on their confinement and slowing down are also presented. Reasons for achievement of high IC heating efficiency are outlined. Reliable H-mode regime achievement is described. Transport ASTRA modelling demonstrated that during NB heated H-mode ion heat diffusivity remains neoclassical and the particle diffusion coefficient inside transport barrier decreases significantly. Analysis was performed of divertor tile and special probe surfaces after irradiation by plasma during a large number of shots (3000–10 000 shots). Mixed layer composition is measured and deuterium retention in different tokamak first wall areas is estimated. Plasma jet injection experiments with upgraded plasma jet are described. Jet penetration to the plasma centre with immediate increase of density and temperature drop is proved and analogy with pellet injection is outlined.
Globus-M球形托卡马克（ST）在中等辅助加热功率（最高0.8 MW）和高比加热功率（最高2–3 MW m⁻³）条件下实现高等离子体参数的实验与模拟研究已有描述。其重要特征包括较低的边缘安全因子范围（2.7 < q < 5），这在ST中并不常见，以及较小的等离子体与器壁间距（3–5 cm）。研究证明了中性束注入（NBI）下的高离子加热效率。文中描述了快离子轨道的数值模拟结果，并讨论了NBI及离子回旋共振加热期间快离子的产生。此外，还介绍了对其约束和慢化过程的研究。文中概述了实现高离子回旋（IC）加热效率的原因，并描述了可靠H模运行的实现。输运模拟（ASTRA）表明，在NBI加热的H模期间，离子热扩散系数保持新经典水平，而输运垒内的粒子扩散系数显著降低。研究还对大量放电（3000–10000次）后偏滤器靶板及专用探针表面进行了分析，测量了混合层成分，并估算了托卡马克第一壁不同区域的氘滞留量。文中还描述了使用改进型等离子体枪进行的等离子体注入实验，证明了射流可穿透至等离子体中心并导致密度立即增加和温度下降，同时概述了其与弹丸注入的类比关系。