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.     

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.     

Proposed high voltage power supply for the ITER relevant lower hybrid current drive system

In the framework of the EFDA task HCD-08-03-01, the ITER lower hybrid current drive (LHCD) system design has been reviewed. The system aims to generate 24 MW of RF power at 5 GHz, of which 20 MW would be coupled to the plasmas. The present state of the art does not allow envisaging a unitary output of the klystrons exceeding 500 kW, so the project is based on 48 klystron units, leaving some margin when the transmission lines losses are taken into account. A high voltage power supply (HVPS), required to operate the klystrons, is proposed. A single HVPS would be used to feed and operate four klystrons in parallel configuration. Based on the above considerations, it is proposed to design and develop twelve HVPS, based on pulse step modulator (PSM) technology, each rated for 90 kV/90 A. This paper describes in details, the typical electrical requirements and the conceptual design of the proposed HVPS for the ITER LHCD system.     

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.     

Simulation of EAST off-axis neutral beam heating and current drive

The capability of off-axis neutral beam heating and current drive has been investigated with NUBEAM for Experimental Advanced Superconducting Tokamak (EAST). Three different approaches to realize off-axis Neutral Beam Injection (NBI) have been studied. Simulation results for on- and off-axis NBI are reported. The effects of the alignment of NBI relative to the magnetic field pitch on off-axis neutral beam heating and current drive are observed and discussed qualitatively. By comparing the numerical results, a most favorable off-axis NBI configuration is recommended. The capability to control sawtooth is also investigated by comparing locations of the q = 1 rational surface and the peak of the fast ion density profile.     

Impact of core electron temperature on current profile broadening with radio-frequency wave heating and current drive in EAST

In recent EAST experiments, current profile broadening characterized by reduced internal inductance has been achieved by utilizing radio-frequency current drives (RFCD). In contrast to previous density scan experiments, which showed an outward shift of the current density profile of lower hybrid current drive (LHCD) in higher plasma density, the core electron temperature (Te(0)) is found to affect the LHCD current profile as well. According to equilibrium reconstruction, a significant increase in on-axis safety factor (q0) from 2.05 to 3.41 is observed by careful arrangement of RFCD. Simulations using ray-tracing code GENRAY and Fokker–Planck code CQL3D have been performed to thoroughly analyze the LHCD current profile, revealing the sensitivity of the LHCD current profile to Te(0). The LHCD current density tends to accumulate in the plasma core with higher current drive efficiency benefiting from higher Te(0). With a lower Te(0), the LHCD current profile broadens due to off-axis deposition of power density. The sensitivity of the power deposition and current profile of LHCD to Te(0) provides a promising way to effectively optimize current profile via control of the core electron temperature.     

Korea's progress on the ITER tritium systems

Korea has been operating four units of CANDU nuclear power plant since 1983. Tritium generated in the heavy water of the plant has been removed by Wolsong TRF (tritium removal facility) since 2007. Korea has developed a 500-kCi BU-type tritium transport container. Furthermore, strictly controlled tritium release to the environment from the CANDU nuclear power plant in Korea will also be a helpful experience for ITER. The BU-type tritium transport container will be unpacked and the quantity of tritium in the metal tritides of the primary vessel will be measured accurately before the tritium is liberated by heating the metal tritides. In the ITER fuel cycle plant, Korea is responsible for the development and supply of the SDS (fuel storage and delivery system). We present the on-going R&D progress to use a non-nuclear material ZrCo for the storage and delivery of tritium. Especially we present the delivery characteristics of the ZrCo hydride bed by considering the fueling requirement and maintainability of the ZrCo hydride bed. We also present various test results for the experimental ZrCo beds and the test programs for the leak tight large scroll pump with a magnetic coupling-drive.     

The remote handling systems for ITER

The ITER remote handling (RH) maintenance system is a key component in ITER operation both for scheduled maintenance and for unexpected situations. It is a complex collection and integration of numerous systems, each one at its turn being the integration of diverse technologies into a coherent, space constrained, nuclearised design. This paper presents an integrated view and recent results related to the Blanket RH System, the Divertor RH System, the Transfer Cask System (TCS), the In-Vessel Viewing System, the Neutral Beam Cell RH System, the Hot Cell RH and the Multi-Purpose Deployment System.     

Numerical simulation of the combined effects of plasma heating and neutron heating loads on the ITER first wall

To understand the combined effect of plasma heating and neutron heating loadings, the distributions of temperature, stress, and strain in different two-dimensional first wall panel models under normal ITER operation condition were simulated using finite element method. The maximum temperature occurs at the Be armor, and reaches 461 °C. High thermal stresses (in the range of 80-200 MPa) are found at the interface between the Be armor and the CuCrZr layer. The maximum thermal stress reaches 324 MPa in the SS316L cooling tube (20 mm diameter), exceeding its yield strength and resulting in a maximum strain of about 1.7% at the tube inner surface. These simulation results are useful for the design and operation of ITER.     

Development of the supervisory systems for the ITER diagnostic systems in JADA

In ITER, it is important how the CODAC system conducts many plant systems including diagnostic systems. In order to establish necessary communications between the diagnostics systems and the CODAC system, Japan domestic agency (JADA) has proposed the new concept of supervisory system for the diagnostic system based on our experiences in operating plasma diagnostic systems. The supervisory system manages operation sequences, current state and configuration parameters for the measurement. JADA designed the supervisory system satisfying the requirements from both CODAC system and diagnostic systems. In our design, the tool which converts operational steps described as flowcharts into the EPICS (experimental physics and industrial control system) records source codes is introduced. This tool will ensure reduction of the system designers’ efforts. We designed a communication protocol to configure measurement parameters and proposed configuration parameter validation function. We also analyzed the management of the central/local control mode for the diagnostic systems. The function which selects the adequate limit values and consistency check algorithms in accordance with the conditions of the diagnostics system is proposed. JADA will develop a prototype of the supervisory system and validate the design in 2013.     

基于CODAC的ITER电流引线测控系统

中国科学院等离子体物理研究所(Institute of Plasma Physical,Chinese Academy of Sciences,ASIPP)负责国际热核聚变实验堆(International Thermonuclear Experimental,ITER)60根高温超导电流引线(High Temperature Superconducting Current Lead,HTSCL)产品的研制与测试,并在2013–2015年间开展了三对三种电流等级(68 k A、55 k A和10 k A)的高温超导电流引线认证制造。为检验电流引线的低温大电流性能,ASIPP与印度塔塔咨询服务公司(Tata Consultancy Service,TCS)及ITER合作开发了基于CODAC(Control,Data Access and Communication)框架的ITER高温超导电流引线测控系统。该系统包括传统西门子PLC300工艺过程测控系统、基于Lab VIEW的失超保护系统、基于PLC400冗余设计的互锁系统和基于NI c系列模块的快速控制系统(Plant system Controller,Fast Controls,PCF)。目前本系统已通过三轮验收测试并在2015年1月份的ITER CC 10 k A电流引线原型件和同年7月份的ITER TF 68 k A电流引线原型件中成功应用。结果表明,本系统能很好地满足电流引线的实验需求,得到ITER国际认同。电流引线测控系统软硬件遵照ITER的CODAC标准进行设计,是CODAC和互锁保护规范的首次在ITER真实组件物理性能测试的联合应用案例,可作为ITER采购包出厂验收推行的CODAC模范。     

ITER operational space for full plasma current H-mode operation

Sensitivity studies performed as part of the ITER IO design review highlighted a very stiff dependence of the maximum Q attainable on the machine parameters. In particular, in the considered range, the achievable Q scales with I_p^{^}4. As a consequence, the achievement of the ITER objective of Q = 10 requires the machine to be routinely operated at a nominal current I_p of 15 MA, and at full toroidal field BT of 5.3 T. This paper analyses the capabilities of the poloidal field (PF) system (including the central solenoid) of ITER against realistic full current plasma scenarios. An exploration of the ITER operational space for the 15 and 17 MA inductive scenario is carried out. An extensive analysis includes the evaluation of margins for the closed loop shape control action. The overall results of this analysis indicate that the control of a 15 MA plasma in ITER is likely to be adequate in the range of li 0.7–0.9 whereas, for a 17 MA plasma, control capabilities are strongly reduced. The ITER operational space, provided by the reference pre-2008 PF system, was rather limited if compared to the range of parameters normally observed in present experiment. Proposals for increasing the current and field limits on PF2, PF5 and PF6, adjustment on the number of turns in some of the PF coils, changes to the divertor dome geometry, to the conductor of PF6 to Nb3Sn, moving PF6 radially and/or vertically are described and evaluated in the paper. Some of them have been included in 2008 ITER revised configuration.     

Progress of IRSN R&D on ITER Safety Assessment

The French “Institut de Radioprotection et de S?reté Nucléaire” (IRSN), in support to the French “Autorité de S?reté Nucléaire”, is analysing the safety of ITER fusion installation on the basis of the ITER operator’s safety file. IRSN set up a multi-year R&D program in 2007 to support this safety assessment process. Priority has been given to four technical issues and the main outcomes of the work done in 2010 and 2011 are summarized in this paper: for simulation of accident scenarios in the vacuum vessel, adaptation of the ASTEC system code; for risk of explosion of gas-dust mixtures in the vacuum vessel, adaptation of the TONUS-CFD code for gas distribution, development of DUST code for dust transport, and preparation of IRSN experiments on gas inerting, dust mobilization, and hydrogen-dust mixtures explosion; for evaluation of the efficiency of the detritiation systems, thermo-chemical calculations of tritium speciation during transport in the gas phase and preparation of future experiments to evaluate the most influent factors on detritiation; for material neutron activation, adaptation of the VESTA Monte Carlo depletion code. The first results of these tasks have been used in 2011 for the analysis of the ITER safety file. In the near future, this R&D global programme may be reoriented to account for the feedback of the latter analysis or for new knowledge.     

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.     

Progress on the ITER ECRH upper launcher steering mirror identification and control

The main objective of the ITER ECRH upper launcher (UL) is to control magnetohydrodynamic activity, in particular neoclassical tearing modes (NTMs), by driving several MW of EC current near the q = 1, 3/2, 2 flux surfaces, where NTMs are expected to occur.The steering of the EC power is done by the steering mechanism assembly (SMA) that comprises a reflecting mirror and a frictionless and backlash free pneumo-mechanical system actuated with pressurised helium gas. The control requirements for this component in terms of steering accuracy and speed are reviewed. With respect to these requirements, the performance of the first SMA prototype is assessed in a mock up of the UL pneumatic configuration.The expected design characteristics of the SMA have been verified and an overall satisfactory performance has been assessed. Furthermore, the main challenges for the future work, such as the pressure and angular position control, have been identified.     

Simulation-based design process for the verification of ITER remote handling systems

The work behind this paper takes place in the EFDA's European Goal Oriented Training programme on Remote Handling (RH) “GOT-RH”. The programme aims to train engineers for activities supporting the ITER project and the long-term fusion programme. One of the projects of this programme focuses on the verification and validation (V&V) of ITER RH system requirements using digital mock-ups (DMU). The purpose of this project is to study and develop efficient approach of using DMUs in the V&V process of ITER RH system design utilizing a System Engineering (SE) framework. Complex engineering systems such as ITER facilities lead to substantial rise of cost while manufacturing the full-scale prototype. In the V&V process for ITER RH equipment, physical tests are a requirement to ensure the compliance of the system according to the required operation. Therefore it is essential to virtually verify the developed system before starting the prototype manufacturing phase. This paper gives an overview of the current trends in using digital mock-up within product design processes. It suggests a simulation-based process design centralized around a simulation lifecycle management system. The purpose of this paper is to describe possible improvements in the formalization of the ITER RH design process and V&V processes, in order to increase their cost efficiency and reliability.     

Progress in control and data acquisition for the ITER neutral beam test facility

SPIDER, the ion source test bed in the ITER neutral beam test facility, is under construction and its operation is expected to start in 2014. Control and data acquisition for SPIDER are undergoing final design. SPIDER CODAS, as the control and data acquisition system is referred to, is requested to manage 25 plant units, to acquire 1000 analogue signals with sampling rates ranging from a few S/s to 10 MS/s, to acquire images with up to 100 frames per second, to operate with long pulses lasting up to 1 h, and to sustain 200 MB/s data throughput into the data archive with an annual data storage amount of up to 50 TB. SPIDER CODAS software architecture integrates three open-source software frameworks each addressing specific system requirements. Slow control exploits the synergy among EPICS and Siemens S7 programmable controllers. Data handling is by MDSplus a data-centric framework that is geared towards the collection and organization of scientific data. Diagnostics based on imaging drive the design of data throughput and archive size. Fast control is implemented by using MARTe, a data-driven, object-oriented, real-time environment. The paper will describe in detail the progress of the system hardware and software architecture and will show how the software frameworks interact to provide the functions requested by SPIDER CODAS. The paper will focus on how the performance requirements can be met with the described SPIDER CODAS architecture, describing the progress achieved by carrying out prototyping activities.     

Status and progress of Indian LLCB test blanket systems for ITER

The lead–lithium ceramic breeder (LLCB) TBM and its auxiliary systems are being developed by India for testing in ITER machine. The LLCB TBM consists of lithium titanate as ceramic breeder (CB) material in the form of packed pebble beds. The FW structural material is ferritic martensitic steel cooled by high-pressure helium gas and lead–lithium eutectic (Pb–Li) flowing separately around the ceramic breeder pebble bed to extract the nuclear heat from the CB zones. Low-pressure helium is purged inside the CB zone for in situ extraction of bred tritium. Currently the LLCB blanket design optimization is under progress. The performance of tritium breeding and high-grade heat extraction is being evaluated by neutronic analysis and thermal–hydraulic calculations for different LLCB cooling configurations and geometrical design variants. The LLCB TBM auxiliary systems such as, helium cooling system (HCS), lead–lithium cooling system (LLCS), tritium extraction system (TES) process design are under progress. Safety analysis of the LLCB test blanket system (TBS) is under progress for the contribution to preliminary safety report of ITER-TBMs. This paper will present the status of the LLCB TBM design, process integration design (PID) of the auxiliary systems and preliminary safety analysis results.     

Development on systems configuration in ITER Tokamak Complex and Auxiliary Buildings

The ITER site consists of almost 30 buildings to service the Tokamak machine which is located in the centre of the Tokamak Complex facility with the Tokamak-, Diagnostic- and Tritium building.The design of a large part of the ITER plant systems will be executed by the ITER Domestic Agencies or their industrial suppliers under functional specifications provided by the ITER Organization. At the same time, the detailed design of the building is carried out by the European Domestic Agency ‘Fusion for Energy’ (F4E).In order to allow an efficient identification of the ITER configuration as well as to manage the concurrent engineering activities and to simplify the identification and assessment of changes, the design of each ITER plant systems is described in the so-called Configuration Management Models (CMM). These are light CATIA^® 3D models that define the required space envelope and the physical interfaces in-between the systems and the buildings.The paper describes the procedure adopted for the control of the baseline configuration of the Tokamak Complex facility and Auxiliary Buildings with their associated plant systems and illustrates the current status as well as recent developments in the different systems.