Results of Predictive Fokker?CPlanck Modelling of NBI Deuterons in ITER

First an analytical formalism is presented for calculating the source distribution of ions generated by neutral beam injection (NBI) in tokamak plasmas. A general NBI ion source term, applicable to studies in the phase space up to 6 dimensions, is provided for neutral beams with finite thickness and divergence. Further, using this source term for the envisaged NBI in ITER, we carry out 3D Fokker?CPlanck modelling of the steady-state deuteron distribution function of NBI produced fast deuterons relaxing on bulk plasma components. For two basic ITER scenarios we demonstrate the poloidal profiles of the beam deuteron density, of the NBI generated current as well as of the NBI power deposition to bulk electrons and ions. Further, we evaluate the capability of gamma and NPA diagnostics of NBI ions in ITER and demonstrate the sensitivity of the distributions of NBI generated ions to different ITER operation scenarios.     

Development of a RF-driven ion source for the ITER NBI system

Extensive R&D work on RF-driven negative hydrogen ion sources carried out at IPP Garching led to the decision of ITER to select this type of source as the new reference source for the ITER NBI system. The principle suitability of the RF source has been demonstrated in a small scale, short pulse length experiment: accelerated current densities, co-extracted electron currents at a source operation pressure, all well inside the range of the ITER requirements have been achieved simultaneously. In subsequent experiments, pulse lengths up to 1 h and the possibility of modularly extending the source to ITER source dimensions were demonstrated. The results achieved at the various IPP test beds, the lessons learnt during optimising the source for negative ion production and extraction as well as the problems still to be solved are summarized. As the next step in support of the NBI development for ITER, IPP plans to build a new test facility for beam extraction from a source of half the size for ITER.     

The negative ion source test facility ELISE

The ITER neutral beam system is using inductively coupled radio frequency (RF) ion sources, that have demonstrated the required ITER parameters on (small) sources with extraction areas up to 200 cm². As a next step towards the full size ITER source IPP is presently constructing the test facility ELISE (“Extraction from a Large Ion Source Experiment”) operating with a “half-size” source which has approximately the width but only half the height of the ITER source. The modular driver concept is expected to allow a further extrapolation to the full size in one direction to be made. The main aim of this experiment is to demonstrate the production of a large uniform negative ion beam with ITER relevant parameters in stable conditions up to one hour.Plasma operation of the source is foreseen to be performed continuously for 1 h; extraction and acceleration of negative ions up to 60 kV is only possible in pulsed mode (10 s every 180 s) due to limitations of the existing IPP HV system. The design of the source and extraction system implements a high experimental flexibility and a good diagnostic access while still staying as close as possible to the ITER design. The main differences are the source operating in air and the use of a large gate valve between the source and the target chamber.ELISE is expected to start operation at the end of 2011 and is an important step for the development of the ITER NBI system; the experience gained early will support the design as well as the commissioning and operating phases of the PRIMA NBI test facilities and the ITER neutral beam system.     

Commissioning and first results of the ITER-relevant negative ion beam test facility ELISE

The test facility ELISE which was constructed in the last three years at the Max-Planck-Institut für Plasmaphysik (IPP), Garching, is an important intermediate step of the development of the neutral beam system for ITER. ELISE allows gaining an early experience of the performance and operation of large RF driven sources for negative hydrogen ions and will give an important input for the commissioning and the design of the SPIDER and MITICA test facilities at Padua and the ITER neutral beam system. ELISE has gone recently into operation with first plasma and beam pulses. The experiments aim at the demonstration of an ion beam at the required parameters within 2 years of operation until end of 2014, the end of the service contract with F4E for the establishment and exploitation of ELISE.     

RF sources for ITER Ion Cyclotron H&CD system

The Ion Cyclotron Heating and Current Drive (IC H&CD) system for ITER will provide 20 MW to the plasma. The associated Radio Frequency (RF) source system has to be compliant with all operation modes foreseen in that frame. Their specifications are fully described in this paper and constraints on IC RF source components are detailed, in particular concerning the final stage tube of the amplifier. Results of tests performed under a collaborative work at the National Institute for Fusion Science (NIFS) facility are presented. Consequences on the procurement process by ITER India (II) are deduced.     

Detailed design of the RF source for the 1 MV neutral beam test facility

In the framework of the EU activities for the development of the Neutral Beam Injector for ITER, the detailed design of the Radio Frequency (RF) driven negative ion source to be installed in the 1 MV ITER Neutral Beam Test Facility (NBTF) has been carried out.Results coming from ongoing R&D on IPP test beds [A. Stäbler et al., Development of a RF-Driven Ion Source for the ITER NBI System, this conference] and the design of the new ELISE facility [B. Heinemann et al., Design of the Half-Size ITER Neutral Beam Source Test Facility ELISE, this conference] brought several modifications to the solution based on the previous design.An assessment was carried out regarding the Back-Streaming positive Ions (BSI+) that impinge on the back plates of the ion source and cause high and localized heat loads. This led to the redesign of most heated components to increase cooling, and to different choices for the plasma facing materials to reduce the effects of sputtering.The design of the electric circuit, gas supply and the other auxiliary systems has been optimized. Integration with other components of the beam source has been revised, with regards to the interfaces with the supporting structure, the plasma grid and the flexible connections.In the paper the design will be presented in detail, as well as the results of the analyses performed for the thermo-mechanical verification of the components.     

Design of the Prototype Negative Ion Source for Neutral Beam Injector at ASIPP

In order to support the design,manufacture and commissioning of the negativeion-based neutral beam injection(NBI) system for the Chinese Fusion Engineering Test Reactor(CFETR),the Hefei utility negative ion test equipment with RF source(HUNTER) was proposed at ASIPP.A prototype negative ion source will be developed at first.The main bodies of plasma source and accelerator of the prototype negative ion source are similar to that of the ion source for EAST-NBI.But instead of the filament-arc driver,an RF driver is adopted for the prototype negative ion source to fulfill the requirement of long pulse operation.A cesium seeding system and a magnetic filter are added for enhancing the negative ion density near the plasma grid and minimizing co-extracted electrons.Besides,an ITER-like extraction system is applied inside the accelerator,where the negative ion beam is extracted and accelerated up to 50 kV.     

RADI—A RF source size-scaling experiment towards the ITER neutral beam negative ion source

IPP Garching is currently developing a negative hydrogen ion RF source for the ITER neutral beam system. The source demonstrated already current densities in excess of the ITER requirements (>200 A/m² D⁻) at the required source pressure and electron/ion ratio, but with only small extraction area and limited pulse length. A new test facility (RADI) went recently into operation for the demonstration of the required (plasma) homogeneity of a large RF source and the modular driver concept.The source with the dimension of 0.8 m × 0.76 m has roughly the width and half the height of the ITER source; its modular driver concept will allow an easy extrapolation in only one direction to the full size ITER source. The RF power supply consists of two 180 kW, 1 MHz RF generators capable of 30 s pulses. A dummy grid matches the conductance of the ITER source. Full size extraction is presently not possible due to the lack of an insulator, a large size extraction system and a beam dump.The main parameters determining the performance of this “half-size” source are the negative ion and electron density in front of the grid as well as the homogeneity of their profiles across the grid. Those will be measured by optical emission and cavity ring down spectroscopy, by Langmuir probes and laser detachment. These methods have been calibrated to the extracted current densities achieved at the smaller source test facilities at IPP for similar source parameters. However, in order to get some information about the possible ion and electron currents, local single aperture extraction with a Faraday cup system is planned.     

Effect of Plasma Rotation on Neutral Beam Heating and Current Drive in Tokamaks

For a rapidly rotating plasma, the effects of the resulting Doppler shift have to be included in the neoclassical theory of neutral beam heating, current drive, and plasma transport. In this paper, an improved simulation of neutral beam injection (NBI) and current drive in rotating plasmas is introduced. NBI is simulated using the Monte Carlo code NUBEAM along with the transport code ONETWO. The physical characteristics of heating and current drive for co- and counter-NBI are investigated for non-rotating, co-rotating, and counter-rotating plasmas, all of which can take place in the experiments. In general, it is found that rotation of the plasma can increase the NBI power deposition on the plasma electrons but has little effect on the ions. Moreover, plasma heating by co-NBI is more efficient than that by counter-NBI. For neutral beam current drive, because of the Doppler shift, co-rotation (counter-rotation) of the bulk plasma tends to decrease the co-NBI (counter-NBI) driven current. On the other hand, due to trapping and orbit loss of the fast ions, co-rotation (counter-rotation) has little effect on the counter-NBI (co-NBI) driven current. The results are applied to the forthcoming NBI heating and current drive experiments of the EAST tokamak and should also be useful in the design of experiments in ITER.     

5 MW CW supply system for the ITER gyrotrons Test Facility

ECH (Electron Cyclotron Heating) for ITER will deliver into the plasma 20 MW of RF power. The procurement of the RF sources will be shared equally between the three following partners: Europe, Japan and Russia. Moreover, Europe decided to develop a RF source capable of 2 MW CW of RF power, based on the design of a coaxial gyrotron with a depressed collector. In order to be able to develop and test these RF sources, a Test Facility (TF) has been built at the CRPP premises in Lausanne (CH).The present paper will first remind the main operation conditions considered to test safely a gyrotron. The power supplies parameters allowing to fulfill these conditions will be reviewed. The core of the paper content will describe the newly installed Main High Voltage Power Supply (MHVPS), to be connected to the gyrotron cathode and capable of ?60 kV/80 A-CW. The principle, the characteristics, the on-site test results will be described at the light of the requirements imposed by the gyrotron testing. Particular aspects of the installation and commissioning on-site will be highlighted in comparison with the ITER environment. The synchronized operation of the MHVPS and the BPS (Body Power Supply) on dummy load, piloted through the TF remote control, will be presented and commented.Since the TF supply structure has been built integrating the particular conditions and requirements expected for ITER, a conclusion will summarize the performances obtained at the light of these criteria.     

Numerical Simulation of Non-Inductive Current Driven Scenario in EAST Using Neutral Beam Injection

For achieving the scientific mission of long pulse and high performance operation,experimental advanced superconducting tokamak(EAST) applies fully superconducting magnet technology and is equiped with high power auxiliary heating system.Besides RF(Radio Frequency) wave heating,neutral beam injection(NBI) is an effective heating and current drive method in fusion research.NBCD(Neutral Beam Current Drive) as a viable non-inductive current drive source plays an important role in quasi-steady state operating scenario for tokamak.The non-inductive current driven scenario in EAST only by NBI is predicted using the TSC/NUBEAM code.At the condition of low plasma current and moderate plasma density,neutral beam injection heats the plasma effectively and NBCD plus bootstrap current accounts for a large proportion among the total plasma current for the flattop time.     

Preliminary design of the control and data acquisition systems for the Neutral Beam Test Facility

The Neutral Beam Test Facility, which will be built in Padova, Italy, is aimed at developing the ITER heating neutral beam injector (HNB) and at testing and optimizing its operation up to nominal performance before installation on ITER. It requires the development of two independent experiments referred to as SPIDER (source for production of ions of deuterium extracted from Rf plasma) and MITICA (megavolt ITer injector & concept advancement). SPIDER will explore the full-size negative ion source for ITER, whereas MITICA will explore the full-size ITER neutral beam injector. Both experiments will be designed for long-pulse operation, up to 3600 s, as ITER itself. MITICA includes three functional components: the heating neutral beam injector plant system (HNB), which is the device under test; the auxiliary plant system (AUX), which includes all equipment to operate the HNB in the test facility (e.g. the local electric grid to feed the HNB power supplies), and MITICA supervisory system that is an electronics/informatics infrastructure to operate the facility. The paper introduces the requirements for the control and data acquisition systems of the experiments and proposes a preliminary design for both systems. SPIDER, which is preparatory to MITICA and will be developed on a shorter time scale, has no constraints coming from ITER CODAC, whereas MITICA includes the ITER neutral beam injector and therefore must be fully compatible with ITER CODAC.     

Commissioning of the first KSTAR neutral beam injection system and beam experiments

The neutral beam injection (NBI) system was designed to provide plasma heating and current drive for high performance and long pulse operation of the Korean Superconducting Tokamak Advanced Research (KSTAR) device using two co-current beam injection systems. Each neutral beam injection system was designed to inject three beams using three ion sources and each ion source has been designed to deliver more than 2.0 MW of deuterium neutral beam power for the 100-keV beam energy. Consequently, the final goal of the KSTAR NBI system aims to inject more than 12 MW of deuterium beam power with the two NBI for the long pulse operation of the KSTAR. As an initial step toward the long pulse (～300 s) KSTAR NBI system development, the first neutral beam injection system equipped with one ion source was constructed for the KSTAR 2010 campaign and successfully commissioned. During the KSTAR 2010 campaign, a MW-deuterium neutral beam was successfully injected to the KSTAR plasma with maximum beam energy of 90 keV and the L-H transition was observed with neutral beam heating. In recent 2011 campaign, the beam power of 1.5 MW is injected with the beam energy of 95 keV. With the beam injection, the ion and electron temperatures increased significantly, and increase of the toroidal rotation speed of the plasma was observed as well. This paper describes the design, construction, commissioning results of the first NBI system leading the successful heating experiments carried in the KSTAR 2010 and 2011 campaign and the trial of 300-s long pulse beam extraction.     

Status of the ITER Ion Cyclotron H&CD system

The ongoing design of the ITER Ion Cyclotron Heating and Current Drive system (20 MW, 40–55 MHz) is rendered challenging by the wide spectrum of requirements and interface constraints to which it is subject, several of which are conflicting and/or still in a high state of flux. These requirements include operation over a broad range of plasma scenarios and magnetic fields (which prompts usage of wide-band phased antenna arrays), high radio-frequency (RF) power density at the first wall (and associated operation close to voltage and current limits), resilience to ELM-induced load variations, intense thermal and mechanical loads, long pulse operation, high system availability, efficient nuclear shielding, high density of antenna services, remote-handling ability, tight installation tolerances, and nuclear safety function as tritium confinement barrier. R&D activities are ongoing or in preparation to validate critical antenna components (plasma-facing Faraday screen, RF sliding contacts, RF vacuum windows), as well as to qualify the RF power sources and the transmission and matching components. Intensive numerical modeling and experimental studies on antenna mock-ups have been conducted to validate and optimize the RF design. The paper highlights progress and outstanding issues for the various system components.     

Potential failure mode and effects analysis for the ITER NB injector

The failure mode and effects analysis (FMEA) is a widely used analytical technique that helps in identifying and reducing the risks of failure in a system, component or process.The application of a systematic method like the FMEA was deemed necessary and adequate to support the design process of the ITER NBI (neutral beam injector). The approach adopted was to develop a FMEA at a general “system level”, focusing the study on the main functions of the system and ensuring that all the interfaces and interactions are covered among the various subsystems. The FMEA was extended to the whole NBI system taking into account the present design status. The FMEA procedure will be then applied to the detailed design phase at the component level, in particular to identify (or define) the ITER Class of Risk.Several important failure modes were evidenced, and estimates of subsystems and components reliability are now available. FMEA procedure resulted essential to identify and confirm the diagnostic systems required for protection and control, and the outcome of this analysis will represent the baseline document for the design of the NBI and NBTF integrated protection system.In the paper, rationale and background of the FMEA for ITER NBI are presented, methods employed are described and most interesting results are reported and discussed.     

Realization of minute-long steady-state H-mode discharges on EAST

In the 2016 EAST experimental campaign,a steady-state long-pulse H-mode discharge with an ITER-like tungsten divertor lasting longer than one minute has been obtained using only RF heating and current drive,through an integrated control of the wall conditioning,plasma configuration,divertor heat flux,particle exhaust,impurity management,and effective coupling of multiple RF heating and current drive sources at high injected power.The plasma current (Ip ～ 0.45 MA) was fully-noninductively driven (Vloop ＜ 0.0 V) by a combination of ～2.5 MW LHW,～0.4 MW ECH and ～0.8 MW ICRF.This result demonstrates the progress of physics and technology studies on EAST,and will benefit the physics basis for steady state operation of ITER and CFETR.     

Progress of the RF negative hydrogen ion source for fusion at HUST

Huazhong University of Science and Technology has developed an experimental setup of a radio frequency (RF) driven negative hydrogen ion source, to investigate the physics of production and extraction of the H− ions for neutral beam injection in nuclear fusion reactors. The main design parameters of the ion source are: RF power ≤40 kW; extraction voltage ≤10 kV; accelerator voltage ≤20 kV. This paper gives an overview of the progress of the ion source with particular emphasis on some issues. The RF driver and source plasma are analyzed and optimized in terms of impedance matching, plasma characteristics and power coupling. In regard to the simulation analysis, a plasma model based on the particle-in-cell method and a beam trajectory model considering beam stripping loss are developed to investigate the plasma and negative ions transport inside the ion source. Furthermore, a collisional radiative model of H and H2 is built for plasma optical diagnosis.     

Achievement of 1000 s plasma generation of RF source for neutral beam injector

A radio frequency(RF) ion source was developed for neutral beam injector in Institute of Plasma Physics, Chinese Academy of Sciences(ASIPP). A cylindrical driver based RF plasma source was tested and optimized for long pulse operation. Recently, the plasma source achieved 1000 s stable plasma discharge with RF power of 35 k W and source pressure of 0.7 Pa for the first time.The heat loading on each of driver components such as the Faraday shield(FS), RF coil and expansion chamber was measured by the water flow calorimeter system. The experiment results showed that FS is the component with highest heat loading and the heat loading on FS was about 65.8% of total heat loading on source. The details of the long pulse operation of RF plasma source are presented in this paper.     

R&D progress of high power ion source on EAST-NBI

The neutral beam injector(NBI) system was designed and developed mainly for the plasma heating on the Experimental Advanced Superconducting Tokamak(EAST). The high power ion source is the key part of the NBI. A hot cathode ion source was used on the EAST-NBI. The ion source was conditioned on the ion source test bed with hydrogen gas and achieved the designed parameters. The deuterium gas was used when it moved to the EAST-NBI. The main performance of the ion source on EAST is presented in this paper. The highest beam power of 4.5 MW in NBI-1 and 2.75 MW in NBI-2 was achieved. The total neutral beam power is about 4.5 MW. The long pulse beam of 100 s is injected into the EAST plasma too.     

Predictive Simulations of Plasma Heating and Current Drive by Neutral Beam Injection on EAST

Long pulse and high performance steady-state operation is the main scientific mission of experimental advanced superconducting tokamak (EAST). In order to achieve this objective, high-power auxiliary heating systems are essential. Radio frequency (RF) wave heating and neutral beam injection (NBI) are two principal methods. NBI is an effective method of plasma heating and current drive, and it has been used in many magnetic confinement fusion devices. Based on the plasma equilibrium of EAST (Li et al., Plasma Phys Control Fusion 55:125008, 2013) plus previous EAST experimental data used as initial conditions, the NBI module (Polevoi et al., JAERI-Data, 1997) employed in automated system for transport analysis (ASTRA) code (Pereverzev et al., IPP-Report, 2002) is applied to predict the effects of plasma heating and current drive with different neutral beam injection power levels. At certain levels of plasma densities and plasma current densities, the simulation results show that the NBI heats plasma effectively, also increases the proportions of NB current and bootstrap current among total current significantly.