M. Nakamoto, H. Kajitani, T. Suwa, Y. Takahashi, M. Yamane, T. Baba, K. Sakamoto, K. Yoshizawa, Y. Uno, A. Ishikawa, M. Nakahira, N. Koizumi, M. Inoue 1, E. Fujiwara 1, T. Shichijyo 1, K. Kuno 2, T. Minato 2, T. Hemmi 3 and C. Luongo 3
National Institutes for Quantum and Radiological Science and Technology, 801-1 Mukouyama, Naka-shi, Ibaraki 311-0193, Japan
1 Mitsubishi Heavy Industries,...
In a recently conducted test for assessing compatibility of accelerator grid of Neutral beam [1] for their performance at 150 C, failure has been evidenced across an electrodeposited (ED) bond layer, which forms a vacuum boundary with cooling medium. This happens to be the first instance where an electrodeposited bond has been subjected to Hot Helium Leak test under operational conditions of...
The present contribution is devoted to the neutral beam injectors (NBIs) for ITER heating and current-drive. First, updated information is provided about the development status of the entire NBI prototype (MITICA); starting in 2021, the first experiments will be dedicated to high-voltage holding tests in vacuum. Then the contribution describes the full-scale prototype of the NBI ion source...
Introduction
Electron cyclotron systems of fusion installations are based on powerful millimetre wave sources โ gyrotrons, which are capable to produce now megawatt microwave power in very long pulses. Gyrotrons for plasma fusion installations usually operate at frequencies 40-170 GHz. Requested output power of the tubes is about 1 MW and pulse duration is between seconds and thousands...
This paper presents a progress of the achievement of performance tests of ITER-gyrotrons developed in QST and design of dual-frequency (170 GHz and 104 GHz) gyrotron to enhance various operation scenarios in ITER such as characteristics studies of H-mode/ELM at low magnetic field. Major achievements of the ITER gyrotron developments are as follows: (i) Manufacturing of 6 out of 8 sets of ITER...
Shattered pellet injection (SPI) systems that form cryogenic pellets of low and high-Z impurities in a pipe-gun [1] for injection to mitigate disruptions have been fabricated and installed for use in thermal mitigation and runaway electron dissipation experiments on JET and KSTAR. These systems are to support disruption mitigation research for ITER and are based on an ORNL 3-barrel design for...
The mitigation of thermomechanical and runaway loads during disruptions and Vertical Displacement Events (VDEs) in ITER is essential for the project to execute the ITER Research Plan culminating (1) in the demonstration of the fusion power production goals (Q = 10 inductive operation for 300-500 s and Q = 5 for 1000 s and in steady-state up to 3000 s). To mitigate these loads ITER is equipped...
The Final Design Review (FDR) of the ITER Plasma Control System (PCS) for First Plasma will be held in July 2020 following the conceptual and preliminary designs [1,2] to prepare for First Plasma operation scheduled for the end of 2025. ITER operation follows the Staged Approach of the ITER Research Plan (IRP) [3]. The main goals of the First Plasma campaign include achieving a plasma current...
The Chinese Helium Coolant Ceramic Breeder (HCCB) Test Blanket Module (TBM) and its ancillary systems (together called Test Blanket System or TBS) is one of important steps for the China magnetic confinement fusion development, which will contribute to validate the key tritium breeding blanket technologies under the burning plasma environment, including tritium extraction, heat removal,...
