High voltage power supplies for ITER RF heating and current drive systems

The RF heating and current drive (H&CD) systems to be installed for the ITER fusion machine are the electron cyclotron (EC), ion cyclotron (IC) and, although not in the first phase of the project, lower hybrid (LH). These systems require high voltage, high current power supplies (HVPS) in CW operation.These HVPS should deliver around 50 MW electrical power to each of the RF H&CD systems with stringent requirements in terms of accuracy, voltage ripple, response time, turn off time and fault energy. The PSM (Pulse Step Modulation) technology has demonstrated over the past 20 years its ability to fulfill these requirements in many industrial facilities and other fusion reactors and has therefore been chosen as reference design for the IC and EC HVPS systems.This paper describes the technical specifications, including interfaces, the resulting constraints on the design, the conceptual design proposed for ITER EC and IC HVPS systems and the current status.     

Design, fabrication and testing of UHV compatible high power RF devices for lower hybrid current drive system on SST-1 tokamak

Maintaining plasma current under steady state conditions is one of the most important pre-requisites for a tokamak-based reactor. Lower hybrid current drive (LHCD) system aims to drive tokamak plasma current by means of RF power. The LHCD system on SST-1 tokamak is based on two 500 kW, CW klystrons operating at 3.7 GHz. A waveguide transmission line transmits power from source to the antenna. A phased array waveguide antenna is used to couple power to the plasma. The antenna side of the transmission line is placed inside the tokamak vacuum vessel. The design and fabrication of this In-Vessel system has to satisfy the demands of high power RF as well as ultra high vacuum (UHV) compatibility. This paper describes some of the critical UHV compatible In-Vessel RF devices, their design, fabrication, and test results.     

Active control system upgrade design for lower hybrid current drive system on Alcator C-Mod

As a part of the scheduled expansion of the Alcator C-Mod lower hybrid current drive (LHCD) system from 12 to 16 klystrons to accommodate installation of a second LH antenna, the active control system (ACS) is being redesigned to accommodate the additional klystrons. Digitizers and output modules will be cPCI modules provided by D-tAcq Solutions. The real-time application will run on a standard PC server running Linux. Initially, the new ACS system will be designed to control 8 klystrons on the second LH antenna and the existing ACS will control the remaining 8 klystrons on the existing LH antenna. Experience gained operating the existing LHCD system has given us insight into the design of a more robust, compact, efficient and simple system for the new ACS. The design upgrade will be patterned on the digital plasma control system (DPCS [1]) in use on C-Mod.     

The lower hybrid current drive system for steady-state operation of the Vulcan tokamak conceptual design

The steady-state current drive system for the Vulcan tokamak concept has been designed, taking into account requirements of high field, small size, and high operational wall temperature (B₀ = 7 T, R₀ = 1.2 m, T_wall > 800 K). This lower hybrid current drive system allows steady-state operation by utilizing high field side launch, high RF source frequency (8 GHz), and dedicated current drive ports. An iterative MHD and current drive solver is used to determine the ideal launching spectra and location to assure strong single pass absorption. It is found that with nominal Vulcan operational parameters (n_e ≈ 4 × 10²⁰ m^?3, T_e ≈ 2.8 keV, I_p = 1.7 MA, P_LHCD = 19.8 MW) bootstrap currents of ～70% and lower hybrid current drive efficiencies of 1.16 × 10¹⁹ A W m^?2 could be achieved. The optimized solution yielded advanced tokamak profiles with q values on-axis above 2. A conceptual design of the system is presented, which takes into account space, power, cooling, and launched spectrum requirements. The system is found to be compatible with the vacuum vessel design and requires cooling power of <1 MW per waveguide bundle.     

Progress on the heating and current drive systems for ITER

The electron cyclotron (EC), ion cyclotron (IC), heating-neutral beam (H-NB) and, although not in the day 1 baseline, lower hybrid (LH) systems intended for ITER have been reviewed in 2007/2008 in light of progress of physics and technology in the field. Although the overall specifications are unchanged, notable changes have been approved. Firstly, it has been emphasized that the H&CD systems are vital for the ITER programme. Consequently, the full 73 MW should be commissioned and available on a routine basis before the D/T phase. Secondly, significant changes have been approved at system level, most notably: the possibility to operate the heating beams at full power during the hydrogen phase requiring new shine through protection; the possibility to operate IC with 2 antennas with increased robustness (no moving parts); the possible increase to 2 MW of key components of the EC transmission systems in order to provide an easier upgrading of the EC power as may be required by the project; the addition of a building dedicated to the RF power sources and to a testing facility for acceptance of diagnostics and heating port plugs. Thirdly, the need of a plan for developing, in time for the active phase, a CD system such as LH suitable for very long pulse operation of ITER was recognised. The review describes these changes and their rationale.     

多路高压电源带载测试仪

介绍了一种多路高压测试系统,并简要介绍了电路结构和电路设计、工作原理和技术指标。该仪器主要用于实验室和野外直流单路高压和多路高压系统电压值的测试,特别是对探测器偏压的测试。该仪器可以测量的额定电压最大值为6kV,电流值为0．1－10mA。仪器结构简单、集成度高、交直流供电、体积小、携带方便。     

Development of high voltage power supply for the KSTAR 170 GHz ECH and CD system

A 3.6 MW (66 kV/55 A) DC power supply system was developed for the 170 GHz EC H&CD system in KSTAR. The power supply system consists of a cathode power supply (CPS), an anode power supply (APS) and a body power supply (BPS). The cathode power supply is capable of supplying a maximum voltage of ?66 kV and a current of 55 A to the cathode with respect to the collector using pulse step modulation (PSM). The high voltage switching system for the cathode is made by a fast MOS-FET solid-state switch which can turn off the high voltage to the cathode within 3 μs in the occurrence of gyrotron faults. The APS is a voltage divider system consisting of a fixed resistor and zener diode units with the capability of 60 kV stand-off voltage. The anode voltage with respect to the cathode is controlled in a range of 0–60 kV by turning the MOS-FET switches connected in parallel to each zener diode on and off. For high frequency current modulation of the gyrotron, the parallel discharge switch is introduced between the cathode and anode in order to clamp the charged voltage in the stray capacitance. The BPS is a DC power supply with the capability of 50 kV/160 mA. The nominal operation parameter of BPS was 23 kV and 10 mA, respectively, and the voltage output is regulated with a stability of 0.025% of the rated voltage. The series MOS-FET solid-state switch is used for on/off modulation in the body voltage sychronizing with anode voltage. The parallel discharge switch is also introduced between the body and collector for high frequency RF modulation. This paper describes the key features of the high voltage power supply system of the KSTAR 170 GHz gyrotron as well as the test results of the power supply.     

Lower hybrid heating and current drive system for a tokamak reactor

A lower hybrid heating system has been designed for heating a tokamak reactor to ignition and for sustaining steady-state operation by driving the toroidal plasma current. The power spectrum from an active/passive waveguide grill is computed, and the resulting equilibrium current density profile is computed from a full electromagnetic WKB analysis of wave propagation in a cylinder. The corresponding toroidal current profile is a low-current equilibrium which is stable to various ideal modes at an economically acceptable beta. The electronic circuitry is designed to minimize the electric power required for current drive, and the resulting design appears to provide reliable operation in a reactor environment. The same system can drive current during reactor startup if some of the waveguides are modified slightly. A typical sequence of startup equilibria is calculated.     

直流高压电源纹波的实验研究

以数字示波器为主要测试设备,提出了从时域、频域、统计域三个角度描述直流高压电源纹波的方法,并分析了时域纹波频谱分析中可能存在的失真问题及解决办法.     

脉冲强磁场高功率脉冲电源系统研制

本文介绍了用于脉冲强磁场大功率电源系统工作原理及研制工作。该系统由58台55心电容器提供1．0MJ储能，采用真空合闸开关作为主放电开关，由大功率二极管和电阻串联构成续流回路。整个脉冲电源系统可以输出最大脉冲宽度30ms，峰值30kA可调脉冲电流，以适应不同参数脉冲磁体实验需要。初步实验已达到30ms脉宽、35T脉冲磁场，该脉冲电源系统工作稳定可靠，具有抗干扰能力，达到设计参数要求，为今后更高参数脉冲强磁场科学实验装置研制提供了重要参考。     

Development of a fast electron bremsstrahlung diagnostic system based on LYSO and silicon photomultipliers during lower hybrid current drive for tokamak

A novel real time fast electron bremsstrahlung (FEB) diagnostic system based on the lutetium yttrium oxyorthosilicate scintillators (LYSO) and silicon photomultipliers (SiPM) has been developed for tokamak. The diagnostic system is dedicated to study the FEB emission in the hard x-ray (HXR) energy range between 10 and 200 keV during the lower hybrid current drive. The system consists of a detection module and three data acquisition and processing (DAP) boards. The detection module consists of annulus LYSO-SiPM detector array and a 12-channel preamplifier module. The DAP boards upload the data to the host computer for displaying and storing through PXI bus. The time and space resolutions of the system are 10 ms and 4cm, respectively. The experimental results can show the evolution over time and the spatial distribution of FEB. This paper presents the system performance and typical discharge results.     

Concept design of fusion power shutdown system for ITER

ITER is a nuclear facility. It is essential to maintain operational safety or to bring this facility to a safe state in case of accidents or incidents. During plasma operation ITER plasma will generate significant heat loads on the plasma facing components. For a few reference accidents there is the need to stop plasma reliably within a certain time. Fusion power shutdown system is the safety system to implement this termination function for ITER. It is based on the concept of massive gas injection.This paper summarizes the safety requirements, logics and the physics requirements on this system for reliable termination of ITER plasma. With regard to the quantity of gas, transient behavior simulation is shown, subsequently providing guideline for laboratory bench-testing. Conceptual engineering design of the system is given together with instrumentation and control specifications.     

Thyristor crowbar system for the high current power supplies of ASDEX upgrade

The ohmic heating system and the poloidal field coils of ASDEX upgrade are supplied by 15 thyristor converter units with an installed apparent power of 600 MVA. To protect the thyristor converters against dc overvoltage arising from abnormal operations and resulting damages caused by the large energy stored in the AUG magnet coils an overvoltage protection system was required. The paper describes the motivation for—and the design and testing of the thyristor crowbar system representing the thyristor converter overvoltage protection system. It will present the layout, analyse the results of measurements obtained during commissioning, compare them to the calculated (design) values and report on the first experience of operation on the AUG coils improving the safety of the equipment.     

Resonant loop system for the KSTAR ICRF current drive

The phased current distribution at current straps for the KSTAR ICRF antenna causes a power imbalance at each strap owing to the mutual couplings between current straps. In order to mitigate the effect of coupling, a decoupler connecting two phased feeding lines are designed based on both a lumped element antenna model and a distributed transmission line model. Though the decoupler parameter is dependent on the loading resistance, which depends on plasma condition, an analysis shows that the decoupling is effective in the wide range of loading resistance assuming the low variation of mutual inductance between straps. A circuit analysis also shows that the RF characteristics of a complex RF transmission system are matched well for the asymmetric antenna current spectrum aiming for a non-inductive current drive of KSTAR. The calibration result of decoupler after installation is also discussed.     

A remotely steered millimetre wave launcher for electron cyclotron heating and current drive on ITER

High-power millimetre wave beams employed on ITER for heating and current drive at the 170 GHz electron cyclotron resonance frequency require agile steering and tight focusing of the beams to suppress neoclassical tearing modes. This paper presents experimental validation of the remote steering (RS) concept of the ITER upper port millimetre wave beam launcher. Remote steering at the entrance of the upper port launcher rather than at the plasma side offers advantages in reliability and maintenance of the mechanically vulnerable steering system. A one-to-one scale mock-up consisting of a transmission line, mitre bends, remote steering unit, vacuum window, square corrugated waveguide and front mirror simulates the ITER launcher design configuration. Validation is based on low-power heterodyne measurements of the complex amplitude and phase distribution of the steered Gaussian beam. High-power (400 kW) short pulse (10 ms) operation under vacuum, diagnosed by calorimetry and thermography of the near- and far-field beam patterns, confirms high-power operation, but shows increased power loss attributed to deteriorating input beam quality compared with low-power operation. Polarization measurements show little variation with steering, which is important for effective current drive requiring elliptical polarization for O-mode excitation. Results show that a RS range of up to −12° to +12° can be achieved with acceptable beam quality. These measurements confirm the back-up design of the ITER ECRH&CD launcher with future application for DEMO.     

Consideration and research of high voltage insulation strategy for ITER Feeder busbar joint

The ITER Feeder busbar joint is an important component which connects different sections of busbar which run from the high temperature superconducting current leads to the magnet terminals. High voltage insulated bus bar sections will be assembled onto the feeder support structure (called the containment duct), and the joints between sections will be made in situ. The high voltage insulation over the joint is the final step to complete the insulation of the whole bus bar system, and its quality is key to achieving the required electrical performance of the system. The final joint has a complex external configuration with an irregular structure in places, which makes it difficult for the application of the insulation. The limited access in the Feeder and the complex joint external structure make the implementation of the joint insulation hard to solve. The Feeder group at ASIPP has carried out significant R&D for the joint insulation, involving research on the selection of materials, the design of the insulation structure, the curing technology and the quality inspection method. After many rounds of manufacturing and testing, the insulation strategy for the Feeder joint will focus on the so-called “wet-winding” technology. Despite the fact that this process has some intrinsic disadvantages, it is more appropriate for the joint structure compared to other methods. The sub-suppliers to ASIPP have chosen this impregnation method for the feeder qualification activities based on guidance from ASIPP. This paper describes the research procedure of the ITER Feeder busbar joint insulation, summarizes the activities on the joint insulation structure design, describes the trials of the impregnation method, and describes the pre-qualification tests. Even for the preferable “wet-winding” process, some improvements have been applied due to recommendations from ITER IO experts. For example, both the use of a vacuum bag during the process and investigation using non-destructive tests (NDT) have been implemented.     

Design of a high power millimeter wave launcher for EC H&CD system on ITER

An optimization procedure of the quasi-optical system for a millimeter wave launcher is developed for the ITER electron cyclotron heating and current drive (EC H&CD) launcher. In the launcher, the radiated RF beams from eight corrugated waveguides are reflected sequentially by two mirrors and injected into a plasma through a small aperture in the blanket shield module on the top of the launcher. Using a steepest decent method, the heat load on the mirrors is successfully reduced to the acceptable level by flattening the RF power profile on the mirrors keeping the scattering of the RF power to a minimum from the mirrors. It is found that 20 MW injection will be acceptable even when the resistivity of 2.64 × 10⁻⁸ Ωm for the surface of the mirror (dispersion strengthened copper, 151 °C assumed) is increased by a factor of ∼10 with a contamination.     

New high power CW test facilities for ITER ICRH components testing

First CW test bed, devoted for ion cyclotron resonance heating (ICRH), is being built at CEA Cadarache. It has been designed for testing the ICRH antenna sub assemblies under ITER relevant conditions (vacuum, cooling and RF).This paper presents a technical overview of these facilities and discusses their future operations in the framework of the ITER ICRH European R&D program.