Design and Development of a Power Supply System for NBI Test Stand of EAST

A neutral beam injector (NBI) test stand was constructed to develop a multi-megawatt prototype ion source as an auxiliary heating system on experimental advanced superconducting tokamak. A power supply system for the NBI test stand components such as a set of dc power supplies for plasma generator, a dc high voltage power supply of a tetrode accelerator, a transmission line and a surge energy suppressor. Stable arc discharges of the plasma generator with hydrogen gases for 100 s long pulse have been produced by six Langmuir probes feedback loop regulation mode to control the arc power supply. The 4 MW hydrogen ion beam of 1 s is extracted with beam energy of 80 keV and the beam current of 52 A. The dc high voltage power supply for the plasma grid of the prototype ion source was designed to contribute maximum voltage of 100 kV and current of 100 A. The high voltage power output is continuously adjustable to satisfy with plasma physics experiment in operation frequency of 10 Hz. To prevent damage of the beam source at high voltage breakdown, core snubber using deltamax soft magnetic materials have been adopted to satisfy the input energy into the accelerator from the power supply can be reduced to about 5 J in the case of breakdown at 80 kV. For the transmission line, a disc shape multi cable coaxial configuration was adopted and which the dimension of the diameter is 140 mm at the core snubber. The major issues of discharge characteristics with long pulse and beam extraction with high power for the prototype ion source were investigated on the NBI test stand.     

Passive protections against breakdowns between accelerating grids in SPIDER experiment

In the SPIDER experiment a ITER-like full size plasma source will be realized with the target to extract a Dˉ beam of 70 A and then to accelerate it to 100 keV energy. The reduction of the effects due to the frequent breakdowns between the accelerating grids is needed, because of grids damage due to energy deposition by arcing and strong electromagnetic noise (EMI) emission. The solution proposed is a comprehensive design of the circuit. Two passive components are installed: a Damping Resistor and an Output Filter in series to the Power Supplies. Then a doubled screened structure will be adopted for the 30 m long – 100 kV Transmission Line TL, which connects the Ion Source and Acceleration Power Supplies to their loads: the Inner Screen will be connected to the reference ground (the vessel) by a resistive link, the Outer Screen acting as a low-impedance ground. Finally, a Distributed Core Snubber DCS (magnetic snubber) will be installed onto the TL, aimed to increase the damping of the oscillations due to the stray inductances and capacitances. The DCS is composed of 10 magnetic alloy cores and is equipped by a biasing circuit to enhance the flux swing in unsaturated condition during the breakdown. A detailed model of the circuit is developed to evaluate the passive components parameters for protection against breakdown, in which all the magnetic and capacitive couplings between components are modeled as well as the magnetic core snubber saturation.     

Design issues of the High Voltage platform and feedthrough for the ITER NBI Ion Source

In the ITER heating Neutral Beam Injector (NBI), a High Voltage air-insulated platform (named High Voltage Deck, HVD) will be installed to host the Ion Source and Extractor Power supply system and associated diagnostics referred to ?1 MV DC potential. All power and control cables are routed from the HVD via a feedthrough (HV bushing) to the gas insulated transmission line which feeds the Injector.The paper focuses on insulation and mechanical issues for both HVD and HV bushing which are very special components, far from the present industrial standards as far as voltage (?1 MV DC) and dimensions are concerned. For this purpose, a preliminary design of the HVD has been carried out as concerns the mechanical structure and external shield. Then, the structure has been verified with a seismic analysis applying the seismic load excitation specified for the ITER construction site (Cadarache) and carrying out verifications according to relevant international standards. As regards the HV bushing design, proposals for the complex inner conductor structure and for interfaces to the HVD and transmission line are outlined; alternative installation layouts (aside or underneath the HVD) are compared from both mechanical and electrical points of view.     

Conceptual design of a high-voltage compact bushing for application to future N-NBI systems of fusion reactors

The energy of future neutral beam injector (NBI) heating systems of fusion power plants ranges from 1 to 2 MeV. They are based on powerful (several tens of MW) hydrogen negative ion electrostatic accelerators where electrodes are polarized by DC high-voltage. The beam line under vacuum is supplied by HV power supplies via a transmission line pressured under SF₆ and a high voltage feedthrough called bushing. The paper presents results obtained over experimental campaigns dedicated to high voltage vacuum insulation for future NBI systems (ITER). It addresses the problematic of the electron field emission and the high voltage breakdown limit under vacuum between large electrode surfaces. The paper highlights the dependence of the electron emission (dark current) with the voltage and the background tank pressure: at low pressure (～1E?3 Pa in hydrogen), an important dark current of I ～ 100 mA has been measured at 500 kV, while at higher pressure (～0.3 Pa in helium), the dark current has been nearly suppressed (less than 3 mA of dark current at 970 kV). The paper shows that a field induced gas adsorption process could occur on the emitting surfaces (cathode), and this process tends to lower the electron field emission current by increasing the work function of the electrode surface. The Fowler–Nordheim law applied to the measured dark current indicates about 70% of work function increase at 0.3 Pa in helium. Finally, a new high-voltage bushing concept relevant to the future NBI systems is presented; it is based on these experimental findings in high voltage vacuum insulation; the main feature of the new bushing concept is to take benefit of the field induced adsorption effect, i.e., the suppression of the dark current with helium gas, in the inner part of the bushing where the electric field intensity is highest.     

Vacuum Insulation and Achievement of 980 keV, 185 A/m~2 H-Ion Beam Acceleration at JAEA for the ITER Neutral Beam Injector

Vacuum insulation of 1 MV is a common issue for the HV bushing and the accel- erator for the ITER neutral beam injector (NBI). The HV bushing as an insulating feedthrough has a five-stage structure and each stage consists of double-layered insulators. To sustain 1 MV in vacuum, reduction of electric field at several triple points existing around the double-layered insulators is a critical issue. To reduce electric field simultaneously at these points, three types of stress ring have been developed. In a voltage holding test of a full-scale mockup equipped with these stress rings, 120% of rated voltage was sustained and the voltage holding capability required in ITER was verified. In the MeV accelerator, whose target is the acceleration of a H ion beam of 1 MeV, 200 A/m 2 , the gap between the grid support was extended to suppress breakdowns triggered by electric field concentration at the edge and corner of the grid support. This modi- fication improved the voltage holding capability in vacuum, and the MeV accelerator succeeded in sustaining 1 MV stably. Furthermore, it appeared that the H ions beam was deflected and a part of the beam was intercepted at the acceleration grid. This causes high heat load on the grids and breakdowns during beam acceleration. To suppress the direct interception, a new grid was designed with proper aperture displacement based on a three dimensional beam trajectory analysis. As a result, 980 keV, 185 A/m 2 H ion beam acceleration has been demonstrated, which is close to the ITER requirement.     

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.     

Study of large gap 200 kV breakdowns in vacuum with high stored energy for ITER neutral beam accelerator

In the context of the ITER contract “ITER/CT/07/219–200 kV Stored Energy Tests”, electrical breakdown tests have been performed in vacuum with a stored energy of up to 425 J. The experiments have been conceived and performed with the collaboration of Consorzio RFX. The tests are being performed in the 1 MV test facility at IRFM, CEA-Cadarache. They should simulate the conditions that will be found in the ITER Neutral Beam accelerator, at 200 kV. This paper presents the set-up of the test bed, the choice of critical components, the diagnostic equipments and the results obtained with 200 kV applied on the anode electrode.     

Progress in CFETR Power Stations Concept Study

While EAST experiment was running in 2012, the project of the China fusion engineering test reactor (CFETR) concept design was started. This ITER-like tokamak system will be the second full superconducting tokamak in China based on EAST technology. In phase I, it has 50–200 MW heat output for demonstrating power generation. The fusion power stations contain complete structure of fusion power plant (FPP) which do not appear in the ITER and huge HV substation which receives power from the 500 kV transmission grid for powering its pulsed power electric network (PPEN) and steady-state electric power network. Furthermore, its structure of turbine generator of FPP is similar to that of a nuclear power station of the pressurized-water reactor. This paper describes the typical CFETR loads and put forward the requirements of short circuit capacity of HV grid. It analyzes different strategies of putting the generator power to the grid, i.e. on the 500 kV grid for future DEMO power structure or 66 kV medium-voltage local grid for self-use. In period between twice burning plasma, conceptual solutions are presented to maintain thermal circuit operation.     

EAST-NBI抑制极电源IGBT串联技术的研究

中性束注入器(Neutral Beam Injector,NBI)是东方超环(Experimental Advanced Superconducting Tokamak,EAST)核聚变实验装置辅助加热的重要组成部分。目前NBI离子源引出系统采用四电极结构,即加速电极、梯度电极、抑制电极和地电极。抑制极电源是为其中的抑制电极提供负电位的高压直流电源。根据抑制极电源输出特性的要求,输出端采用串联绝缘栅双极型晶体管(Insulated Gate Bipolar Transistor,IGBT)作为调制开关。为研究IGBT串联技术对均压效果和抑制极电源输出特性的影响,采用PSpice软件对IGBT开关进行了建模,并进行了不同电路参数下的仿真。仿真表明:一定条件下,电阻电容二极管(Resistance Capacitance Diode,RCD)缓冲电路中电容参数对动态均压效果和电源关断特性具有决定性影响,缓冲电阻影响电容的放电时间及放电电流峰值。最后给出了相应的实验测试结果。该研究结果可以明确缓冲电路参数与均压效果以及抑制极电源开关特性之间的定量关系,为抑制极电源开关特性的进一步优化及其与加速极电源的特性匹配提供数据指导,对于NBI离子源的安全稳定运行具有重要意义。     

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.     

An Improved Parallelogram Hysteresis Loop Used in the HVDC Snubber

The snubber is a transformer-like fault-protection device in the heating neutral beam of the fusion device, such as ITER, which is designed to suppress the short-circuit current and protect the device during HV sparking. In process of modeling the nonlinear equivalent inductance of the snubber, the permeability changes stepwise at the turn points of the parallelogram hysteresis loop. Due to the step change of the permeability, the overshoot of the short-circuit current occurs in simulation. In order to eliminate the current overshoot, the piecewise linear interpolation method is adopted to optimize the hysteresis loop. Simulation results show that the current overshoot is eliminated and its good accordance with the test result verifies the feasibility and accuracy of the improved parallelogram hysteresis loop.     

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.     

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.     

EAST强流离子源电源系统的初步测试运行

测试NBI大功率强流离子源的综合测试台正在建设,已研制了离子源等离子体发生器电源系统、等离子体电极高压电源及梯度极分压器、抑制极负高压电源等电源系统,以及高压传输线及缓冲器,在测试台上开展了对EAST中性束注入器第一台兆瓦级强流离子源样机进行整体电源系统测试和离子源起弧放电的初步测试,完成了离子源电源系统初步性能测试及...     

Plasma scenarios, equilibrium configurations and control in the design of FAST

The Fusion Advanced Studies Torus (FAST) conceptual study has been proposed [A. Pizzuto on behalf of the Italian Association, The Fusion Advanced Studies Torus (FAST): a proposal for an ITER Satellite facility in support of the development of fusion energy, in: Proceedings of 22nd IAEA Fusion Energy Conference, Geneva, Switzerland, October 13–18, 2008; Nucl. Fusion, submitted for publication] as possible European ITER Satellite facility with the aim of preparing ITER operation scenarios and helping DEMO design and R&D. Insights into ITER regimes of operation in deuterium plasmas can be obtained from investigations of non linear dynamics that are relevant for the understanding of alpha particle behaviours in burning plasmas by using fast ions accelerated by heating and current drive systems.FAST equilibrium configurations have been designed in order to reproduce those of ITER with scaled plasma current, but still suitable to fulfil plasma conditions for studying burning plasma physics issues in an integrated framework. In this paper we report the plasma scenarios that can be studied on FAST, with emphasis on the aspect of its flexibility in terms of both performance and physics that can be investigated. All plasma equilibria satisfy the following constraints: (a) minimum distance of 3 energy e-folding length (assumed to be 1 cm on the equatorial plane) between plasma and first wall to avoid interaction between plasma and main chamber; (b) maximum current density in the poloidal field coils, transiently, up to around 30 MA/m². The discharge duration is always limited by the heating of the toroidal field coils that are inertially cooled by helium gas at 30 K. The location of the poloidal field coils has been optimized in order to: minimize the magnetic energy; produce enough magnetic flux (up to 35 Wb stored) for the formation and sustainment of each scenario; produce a good field null at the plasma break-down (B_P/B_T < 2 × 10⁻⁴ at low field, i.e. B_T = 4 T and E_T = 2 V/m for at least 40 ms).Plasma position and shape control studies will also be presented. The optimization of the passive shell position slows the vertical stability growth time down to 100 ms.     

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.     

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.     

A Design Model of Core Snubber for EAST NBI

A transformer type iron core snubber, as a protective device against the stray capacitance during the breakdown in EAST, is analyzed in detail. Three kinds of topology are presented. Then with the analysis for equivalent circuit, the ranges of three key parameters, i.e., secondary side resistance, leakage inductance and snubber inductance, are determined. By considering the saturation of the magnetic material, a design principle is also presented. A nearly 1:10 core snubber is tested. It is proved that a high permeability core with secondary resistor can restrain the discharge current effectively.     

Study on high conversion type core of innovative water reactor for flexible fuel cycle (FLWR) for minor actinide (MA) recycling

In order to ensure sustainable energy supplies in the future based on the well-established light water reactor (LWR) technologies, conceptual design studies have been performed on the innovative water reactor for flexible fuel cycle (FLWR) with the high conversion ratio core. For early introduction of FLWR without a serious technical gap from the LWR technologies, the conceptual design of the high conversion type one (HC-FLWR) was constructed to recycle reprocessed plutonium. Furthermore, an investigation of minor actinide (MA) recycling based on the HC-FLWR core concept has been performed and is presented in this paper. Because HC-FLWR is a near-term technology, it would be a good option in the future if HC-FLWR can recycle MAs. In order to recycle MAs in HC-FLWR, it has been found that the core design should be changed, because the loaded MA makes the void reactivity coefficient worse and decreases the discharge burn-up. To find a promising core design specification, the investigation on the core characteristics were performed using the results from parameter surveys with core burn-up calculations. The final core designs were established by coupled three dimensional neutronics and thermal–hydraulics core calculations. The major core specifications are as follows. The plutonium fissile (Puf) content is 13 wt%. The discharge burn-up is about 55 GWd/t. Around 2 wt% of Np or Am can be recycled. The MA conversion ratios are around unity. In particular, it has been found that loaded Np can be transmuted effectively in this core concept. Therefore, these concepts would be a good option to reduce environmental burdens.     

The use of an electrothermal plasma gun to simulate the extremely high heat flux conditions of a tokamak disruption

Disruption damage conditions for future large tokamaks like ITER are nearly impossible to simulate on current tokamaks. The electrothermal plasma source SIRENS has been designed, constructed, and operated to produce high density (> 10²⁵/m³), low temperature (1–3 eV) plasma formed by the ablation of the insulator with currents of up to 100 kA (100 s pulse length) and energies up to 15 kJ. The source heat fluence (variable from 0.2 to 7 MJ/m²) is adequate for simulation of the thermal quench phase of plasma disruption in future fusion tokamaks. Different materials have been exposed to the high heat flux in SIRENS, where comparative erosion behavior was obtained. Vapor shield phenomena has been characterized for different materials, and the energy transmission factor through the shielding layer is obtained. The device is also equipped with a magnet capable of producing a parallel magnetic field (up to 16 T) over a 8 msec pulse length. The magnetic field is produced to decrease the turbulent energy transport through the vapor shield, which provides further reduction of surface erosion (magnetic vapor shield effect).