Reassessment of steady-state operation in ITER with NBI and EC heating and current drive

Nuclear Fushion

article

A.R. Polevoi¹  A.A. Ivanov²  S.Yu. Medvedev²  G.T.A. Huijsmans⁴  S.H. Kim¹  A. Loarte¹  E. Fable⁵  A.Y. Kuyanov³ 

[1] ITER Organisation;Keldysh Institute of Applied Mathematics;NRC ‘Kurchatov Institute’;CEA;Max-Planck Inst. für Plasmaphysik

关键词: ITER;    steady-state;    transport modelling;    scenarios;    MHD stability;   

DOI  :  10.1088/1741-4326/aba335

来源: Institute of Physics Publishing Ltd.

【 摘 要 】

One of the three goals of the ITER project is to demonstrate fusion gainQ≥ 5 in steady-state operation (SSO). A reassessment of the necessary conditions forQ≥ 5 SSO in ITER has been carried out for steady-state scenarios assuming no internal transport barrier in the core plasma and with the sole use of neutral beam injection (NBI) and electron cyclotron heating and current drive (EC H&CD) as heating and current drive sources in these scenarios. The parametric operational space for SSO in ITER has been reassessed utilizing an inverse evaluation approach that takes into account the baseline design of the NBI and EC H&CD systems withP_NBI= 33 MW,P_EC= 20 MW and their possible upgrade capabilities included in their design,P_NBI≤ 49.5 MW, andP_EC≤ 30 MW. The optimal operational points from this evaluation approach have been chosen for detailed 1.5-D transport modelling by ASTRA and followed-up by magnetohydrodynamic (MHD) stability analysis to demonstrate the conditions in which theQ= 5 SS goal can be achieved in ITER where the plasma current is fully accounted for by the driven current by NBI and EC H&CD and the self-driven bootstrap current. The ideal MHD stability of plasma configurations with self-consistently simulated plasma profiles and equilibrium for the chosen OPs has been analysed by the KINX code. Using this analysis, the possibility to control the MHD stability of these steady-state plasmas by tailoring the current profile with the flexibility provided by design of the systems for electron cyclotron current drive and neutral beam current drive in ITER has been demonstrated. Issues related to the realization of such scenarios from the point of view of plasma physics, experimental demonstration and design limits of the ITER systems and components is discussed.

【 授权许可】

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