Absolute Measurements of the Magnetic Field Generated by Different Coils in the Center of EGYPTOR Tokamak

The present work is devoted to measure the absolute magnetic field produced by different coils in the EGYPTOR tokamak using a calibrated pickup coil. Scaling these measurements in different equations connected with the discharge currents from each supply system are performed. The pickup coil used in the present study is well calibrated with Helmholz coils at the IPP in Prague, Czech Republic. A 0.2% deviation has been found between an evaluation done in the present study and the calibration using Helmholz coils. Experimental measurements of the toroidal magnetic field are in good agreement with calculations to within 2%. Very low values of stray magnetic field components arising from TF and OH coils are recorded which proves that the compensation coils for these components are sufficient.     

HT－7托卡马克的八管弹丸注入器

介绍了用于ＨＴ－７托卡马克的的八管弹丸注入器的物理、工程设计原理和结构特点及配置的各种诊断手段。注入器采用气动发射技术，弹丸为１ｍｍ×１ｍｍ，１．２ｍｍ×１．２ｍｍ，１．５ｍｍ×ｌ．５ｍｍ圆柱体氢丸，丸速０．８～１．５ｋｍ／ｓ工作频率１～８Ｈｚ。     

Recent Progress of HT-7U Superconducting Tokamak

HT-7U is a superconducting tokamak. which is being constructed in Institute of Plasma Physics, Chinese Academy of Sciences. The mission of the HT-7U project is to develop a scientific and engineering basis of the steady state operation of advanced tokamak.The engineering design of the device has been optimized. The R&D program is going on. Short samples of the conductor and a CS model coil were tested. All the TF and PF coils will be manufactured and tested in Institute of Plasma Physics. Therefore, a 600-meter long jacketing line for cable-in-conduit conductors along with two winding machines, a set of VPI equipment and a test facility for the TF and PF coils are ready in ASIPP now. In this paper, the recent progress of the HT-7U is described.     

Numerical analyses of high‐beta spherical tokamaks with varied elongation

The effect of plasma elongation on the second‐stable spherical tokamak (ST) was numerically studied using the experimentally measured pressure and current profiles of ultrahigh‐beta STs. The maximum beta of ST over 50% was obtained in the TS‐3 ST/CT experiment by applying an external toroidal field to an FRC. It was found that the marginal beta for the ballooning instability increased with the plasma elongation κ of ST. The elongated STs with κ > 2 have the magnetic shear (S)–pressure gradient (α) profiles located in the second‐stable regime for the ballooning mode and the stability margin increased with κ. The close relation between the absolute minimum‐B profile and the second stability was documented. The effect of elongation on maximum beta was observed to saturate when κ exceed 3, indicating that the optimized elongation for high‐beta STs is located around 2 < κ < 3. © 2006 Wiley Periodicals, Inc. Electr Eng Jpn, 155(4): 1–6, 2006; Published online in Wiley InterScience ( www.interscience.wiley.com ). DOI 10.1002/eej.20132     

The need and prospects for improved fusion reactors

Conceptual fusion reactor studies over the past 10–15 yr have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points toward smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion reactors. The results of a number of recent conceptual designs of reversed-field pinch, spheromak, and tokamak fusion reactors are summarized as examples of more compact approaches. While a focus has been placed on increasing the fusion-power-core mass power density beyond the minimum economic threshold of 100–200 kWe/tonne, other means by which the overall attractiveness of fusion as a long-term energy source are also addressed.Nomenclature a Plasma minor radius at outboard equatorial plane (m) - A Plasma aspect ratioR _T /a - AC Annual charges ($/yr) - b Plasma minor radius in vertical direction (m) - B Magentic field at plasma or blanket (T) - B _c Magnetic field at the coil (T) - B Toroidal magnetic field (T) - B Poloidal magnetic field (T) - BOP Balance of plant - C Coil - COE Cost of electricity (mills/kWeh) - CRFPR Compact RFP reactor - CT Compact torus (FRC or spheromak) - c _FPC Unit cost of fusion power core ($/kg) - DC Direct cost ($) - DZP Dense Z-pinch - E Escalation rate (1/yr) - EDC Escalation during construction ($) - ET Elongated tokamak - F Annual fuel charges ($/yr) - FC Component of UDC not strongly dependent or FPC size ($/kWe) - FW First wall - FPC Fusion power core - f _Aux Fraction of gross electric power recirculated to BOP - f ₁ (IC+IDC+EDC)/DC - f ₂ (O&M + SCR + F)/AC - IC Indirect cost ($) - IDC Interest during construction ($) - I _w Neutron first-wall loading (MW/m²) - i Toroidal plasma current (MA) - j Plasma current density, I/a² - k _B Boltzmann constant, 1.602(10)^–16 (J/keV) - LWR Light-water (fission) reactor - MPD Mass power density 1000P_E/M_FPC (kWe/tonne) - M _N Blanket energy multiplication of 14.1-MeV neutron energy - M _FPC Mass of fusion power core (tonne) - n Plasma density (m^–3) or toroidal MHD mode number - O&M Annual operating and maintenance cost ($/yr) - p _f Plant availability factor - PFD Poloidal field dominated (CTs, RFP, DZP) - P Construction time (yr) - P_TH Thermal power (MWt) - P _E Net electric power (1-)P _ET (MWe) - P_ET Total gross electric power (MWe) - p_f Fusion power (MW) - q Tokamak safety factor (B /B _gq )(a/R _T ) - q _e EngineeringQ value, 1/e - R _T Major toroidal radius (m) - RFP Reversed-field pinch - RPE Reactor plant equipment (Account 22) - S Shield - SCR Annual spare component cost ($/yr) - SSR Second stability region for the tokamak - S/T/H Stellarator/torsatron/heliotron - ST Spherical tokamak or spherical torus - T Plasma temperature (keV) - TDC Total direct cost ($) - TOC Total overnight cost ($) - UDC Unit direct cost,TDC/10 ³ P _E ($/kWe) - V _p Plasma volume (m³) - W _p Plasma energy (GJ) - W _B Magnetic field energy (GJ) - Magnetic utilization efficiency, 2nk_BT/(B ²/2₀) - ₀ Permeability of free space, 4(10)^–7 H/m - XE Plasma confinement efficiency, a²/4_E - _e Plasma energy confinement time - _p Overall plant efficiency, _TH(1-) - _TH Thermal conversion efficiency - _FPC AverageFPC mass density (tonne/m³) - Plasma vertical elongation factor,b/a - Thickness of allFPC engineering structure surround plasma (m) - Total recirculating power fraction, (P _ET-P _E)/P _ET, or inverse aspect ratioa/R _T This work was performed under the auspices of USDOE, Office of Fusion Energy.     

Comparison of mainline and alternate approaches to fusion energy

The tokamak and tandem mirror concepts are compared with alternate confinement concepts using the criteria established in DOE/ET-0047, An Evaluation of Alternate Magnetic Fusion Concepts 1977. The concepts are evaluated and rated in each of three broad categories: confidence in physics and technology, and reactor desirability. The STARFIRE and MARS reactors are used as a basis for comparing the mainline tokamak and tandem mirror concepts with the alternate concepts evaluated in DOE/ET-0047. Two recent alternate concepts, theohmically heated toroidal experiment (OHTE) and thecompact reversed field pinch reactor (CRFPR), are also evaluated. Results indicate that the physics of the mainline tokamaks and tandem mirrors is better understood than that of most alternate concepts. Both mainline concepts rank near the middle for technology requirements, and both rank near or at the bottom when compared with the reactor desirability of alternate concepts.     

Overview of Tiber-II—Evolution toward an Engineering Test Reactor

TIBER II is designed to be a minimum size and cost candidate for an international Engineering Test Reactor. High-current density Nb₃Sn superconducting magnets with radiation-tolerant polymide insulation is combined with a minimum thickness tungsten inbored shield and a common, external vacuum boundary to minimize the inner radial build of the tokamak core. This results in a major radius of 3 m, compared to 5 m for previous ETR designs such as INTOR, with correspondingly lower costs expected. Cyclic stress fatigue limits the number of pulses so that steady-state current drive, based on a combination of neutral beams, lower hybrid and ECH, is designed to achieve reactor-relevant nuclear testing conditions (Fluence 3MW yr/m², rwall > 1 MW/m² in steady state).This report is abstracted from a more complete information document UCID-20863 with numerous authors. See Ref. l for complete credits.     

A low wall-loading DEMO reactor design with high priority for early and reliable realization of a tokamak fusion reactor over the cost performance

Based on scientific databases adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R&D program, a low wall-loading DEMO fusion reactor has been designed, where high priority has been given to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the major radius of this DEMO reactor is chosen to be 10 m, plasma ignition is achievable with a low fusion power of 0.8 GW and an operation period of 4–5 hours is available only with inductive current drive. The low ignition power makes it possible to adopt a first wall with an austenitic stainless steel, for which significant databases and operating experience exists, due to its use in the presence of neutron irradiation in fission reactors. In step with development of advanced materials, a step-wise increase of the fusion power seems to be feasible and realistic, because this DEMO reactor has the potential to produce a fusion power of 5 GW.     

基于LightGBM的EAST杂质破裂预警

对全超导托卡马克核聚变实验装置东方超环(EAST)运行放电期间发生的杂质破裂进行预测对未来的聚变装置的长脉冲稳态放电有重要意义. 根据杂质破裂的物理特性筛选出的2018年的334炮杂质破裂炮数据以及2021年的1628炮非破裂炮作为训练炮, 再由等离子体平衡、密度、电流以及辐射等8种诊断信号组成的训练样本以LightGBM算法训练出杂质破裂预测模型. 实验结果表明LightGBM算法模型可以对杂质破裂进行准确预测(成功预测率96.29%), 非破裂炮的误判率6.87%. 研究结果证明利用LightGBM进行EAST等离子体杂质破裂预警是可行的方案.     

Initial high beta operation of the HBT-EP Tokamak

HBT-EP is a new research tokamak designed and built to investigate passive and active feedback techniques to control MHD instabilities. In particular, HBT-EP will be able to test techniques to control fast MHD instabilities occurring at high Troyon-normalized beta, _N Ba/I_p [Tm/MA], since it is equipped with a thick, close-fitting, and adjustable conducting shell. The major goals of the initial operation of HBT-EP have been the achievement of high beta operation (_N 3) using only ohmic heating and the observation of MHD instabilities. By using a unique fast startup technique, we have successfully achieved these goals. A variety of MHD phenomena were observed during the high beta operation of HBT-EP. At modest beta (_N 2), discharges have been maintained for more than 10 msec, and these discharges exhibit saturated resistive instabilities. When _N approaches 3, major disruptions occur preceded by oscillating, growing precursors. During start-up, one or more minor disruptions are usually observed. A 1-D transport code has been used to simulate the evolution of the current profile, and these early minor instabilities are predicted to be double tearing modes. The simulation also reproduces the observed high beta operation when saturated neo-Alcator energy confinement scaling is assumed.