Long pulse enhanced confinement discharges in the HT-7 superconducting tokamak by ion Bernstein wave heating and lower hybrid wave current drive

Abstract

Significant progress has been made on the HT-7 superconducting tokamak (R=1.22 m, a=0.27 m) in the past year toward obtaining advanced operating modes in steady state. By combining ion Bernstein wave and lower hybrid current drive edge-localized-mode-free limiter H-mode discharges with H89∼2, βN*H89>3 have been obtained that lasted for 53 τE. A well-boronized wall prevented further impurity accumulation and the large pumping kept the recycling at a very low level. The density was kept almost constant with the feedback control. The noninductive driven current fraction is about 60% and bootstrap fraction is about 18%. The central electron temperature is about 2.6 keV and density above 2×1019 m−3. With reduced plasma parameters of central Te∼1.5 keV and ne0∼2×1019 m−3, H89∼1.45, βN∼1.1, the improved confinement phase lasts for more than 100 τE with the noninductive driven current fraction ∼70%.

Study of confinement and LHCD efficiency on the HT-7 tokamak

“…It has high thermal conductivity up to 180 W mK −1 , whose good thermal shock resistance can withstand 6 MW m −2 high heat loads for 60 s. All carbon tiles were coated with 100 µm SiC functional gradient coating, which was also used for the limiter and the Faraday screen of the IBW antenna. Twenty four pieces of a ferromagnetic material (ferritic steel) were installed inside the vacuum chamber, to be used as plasma facing components (PFCs) and for the reduction of the magnetic ripple [8]. About 1 min duration of long-pulse plasmas was achieved on HT-7 in 2003 as shown in figure 1(a).…”

“…Efforts to further optimize plasmas will benefit from the following aspects: (1) higher power, longer duration, (2) the development of feedback control of the current profile and (3) high LHCD efficiency. Experiments in many tokamaks such as Tore Supra [1], TRIAM-1M [2], FTU [3], JET [4], JT-60U [5] and HT-7 [6][7][8][9][10] have shown that LHCD is one of the most efficient methods to drive non-inductive current in tokamak plasma. In a Tore Supra tokamak [1], up to 0.5 MA full non-inductively driven current with 6 min pulse length has been achieved.…”

Single Gaussian process method for arbitrary tokamak regimes with a statistical analysis

“…Equilibrium reconstruction has emerged as a critical tool for tokamak control [4] and in plasma performance optimization [5][6][7][8]. The most widely-used code for equilibrium reconstruction is EFIT [2,3], which has been used at the DIII-D [3,9], C-MOD [10], JET [11], HIT [12], START [13,14], MAST [15], KSTAR [16,17], JT-60U [18], NSTX [19], TORE SUPRA [8,20], HT-7 [21], HL-2A [22], and EAST [23,24] devices. Equilibrium reconstruction is a critical part of the development of the fusion.…”

Single Gaussian Process Method for Arbitrary Tokamak Regimes with a Statistical Analysis

“…HIT [14], START [15,16], MAST [17], KSTAR [18,19], JT-60U [20], NSTX [21], TORE SUPRA [8,22], HT-7 [23], HL-2A [24], and EAST [25,26] devices. Equilibrium reconstruction is a critical part of the development of the fusion.…”