An Indian test facility to characterise diagnostic neutral beam for ITER

The diagnostic neutral beam (DNB) line shall be used to diagnose the He ash content in the D–T phase of the ITER machine using the charge exchange recombination spectroscopy (CXRS). Implementation of a successful DNB at ITER requires several challenges related to the production, neutralization and transport of the neutral beam over path lengths of 20.665 m, to be overcome. The delivery is aided if the above effects are tested prior to onsite commissioning. As DNB is a procurement package for INDIA, an ITER approved Indian test facility, INTF, is under construction at Institute for Plasma Research (IPR), India and is envisaged to be operational in 2015. The timeline for this facility is synchronized with the RADI, ELISE (IPP, Garching), SPIDER (RFX, Padova) in a manner that best utilization of configurational inputs available from them are incorporated in the design. This paper describes the facility in detail and discusses the experiments planned to optimise the beam transmission and testing of the beam line components using various diagnostics.     

Gas flow and related beam losses in the ITER neutral beam injector

The gas flow in the ITER neutral beam injectors has been studied using a 3D Monte Carlo code to define a number of key parameters affecting the design and operation of the injector. This paper presents the results of calculations of the gas density in the two accelerator concepts presently considered as options for the ITER injectors, and the resultant stripping losses of the negative ions during their acceleration to 1 MeV. The sensitivity of the model to various parameters has been studied, including the gas temperature in the ion source and the subsequent accommodation by collisions with the accelerator structure, and the degree of dissociation of the D₂ or H₂ in the ion source, and subsequent recombination during collisions with the accelerator structure. Additionally the sensitivity of the losses to details of the beam source design and operating parameters are examined for both accelerator concepts.     

Efficiency of neutral beam neutralizers in JET and ITER

Neutral beam neutralizer efficiency plays a key role in determining the overall efficiency of neutral beam systems. Understanding the shortfall in neutralization efficiency encountered in positive ion neutral beam systems at JET is therefore of importance in ensuring the adequacy of the ITER design and in formulating beam-line designs for DEMO. Experimentation has previously demonstrated both the presence of background plasma and elevated gas temperatures, suggesting that the reduced efficiency may be due to a reduction in gas density. However, historical modifications to the neutralizer design at JET in accordance with observations from models produced little improvement in the neutral beam power delivered to the tokamak.This paper describes the development of the neutralizer models from an initial global heating balance for the gas alone, through to the recent application of computational fluid dynamic (CFD) to provide a consistent beam–plasma–gas system able to capture details of the neutral gas flow within the neutralizer. It is demonstrated that for the JET neutralizer a full 3D computation is necessary to correctly capture the behavior of the beam–plasma–gas system. The analysis is also extended to the ITER neutralizer.Overall, the importance of capturing the full complexity of neutral beam neutralizers is highlighted. The necessity of developing 3D modeling capability to support the design of future DEMO systems is demonstrated not only for beam neutralizers but for other beam components that include a fluid element, such as the duct.     

Design of the in-wall shielding for the ITER neutral beam port

The ITER neutral beam port is composed of connecting duct, port extension and port stub extension. The spaces between inner and outer shells of the port extension and port stub extension are filled with pre-assembled blocks, called in-wall shielding. The main purpose of IWS is to provide neutron shielding for the superconducting magnet, thermal shield and cryostat from the main vessel during plasma operation. In order to provide effective neutron shielding capability with the cooling water, 40 mm thick flat plates (steel type 304B4) are used in almost all areas of the volume between port shells. The IWS is composed of shield plates, upper/lower brackets and bolt/nut/washers. Major activities during design work are to develop installation concept of the IWS blocks for easy assembly into port structures and to perform structural analysis to assess sufficient strength, fabrication feasibility study and 3D modeling including drawing works.In this paper, major results of mechanical design are introduced. First, the design requirements for IWS and the developed IWS designs for easy assembly into the port structure are introduced. Second, is introduced the engineering analysis results to assess structural integrity. And then the fabrication feasibility study results are presented for major fabrication processes. Lastly, conclusion and future works are mentioned.     

Korean high heat flux test facility by using electron beam system for ITER first wall semi-prototype

A Korean high heat flux test facility for the semi-prototype (SP) qualification of an ITER first wall (FW) will be constructed to evaluate the fabrication technologies required for the ITER FW, and the acceptance of these developed technologies will be established for the ITER FW manufacturing procedure. Korea participated in this qualification program, and is responsible for suitable arrangements for the heat flux test of our fabricated SPs. Qualification testing can be started provided that adequate quality and control measures are implemented and validated by the ITER Organization (IO). The controlling measures required for all heat flux tests shall be concrete and demonstrate the satisfaction of the IO test programs. Each country shall provide a test plan covering the quality and controlling measures in the high heat flux test facility to be implemented throughout the test program. Korean high heat flux testing for these ITER plasma facing materials will be performed by using a 60 kV electron beam and a power supply system of 300 kW, where the allowable target dimension is 70 cm × 50 cm in a vacuum chamber. In addition, this facility needs a cooling system for a high-temperature target and decontamination system for beryllium filtration.     

Recent improvements to the ITER neutral beam system design

The ITER [1] fusion device is expected to demonstrate the feasibility of magnetically confined deuterium–tritium plasma as an energy source which might one day lead to practical power plants. Injection of energetic beams of neutral atoms (up to 1 MeV D⁰ or up to 870 keV H⁰) will be one of the primary methods used for heating the plasma, and for driving toroidal electrical current within it, the latter being essential in producing the required magnetic confinement field configuration. The design calls for each beamline to inject up to 16.5 MW of power through the duct into the tokamak, with an initial complement of two beamlines injecting parallel to the direction of the current arising from the tokamak transformer effect, and with the possibility of eventually adding a third beamline, also in the co-current direction. The general design of the beamlines has taken shape over the past 17 years [2], and is now predicated upon an RF-driven negative ion source based upon the line of sources developed by the Institute for Plasma Physics (IPP) at Garching during recent decades [3], [4], [5], and a multiple-aperture multiple-grid electrostatic accelerator derived from negative ion accelerators developed by the Japan Atomic Energy Agency (JAEA) across a similar span of time [6], [7], [8]. During the past years, the basic concept of the beam system has been further refined and developed, and assessment of suitable fabrication techniques has begun. While many design details which will be important to the installation and implementation of the ITER beams have been worked out during this time, this paper focuses upon those changes to the overall design concept which might be of general interest within the technical community.     

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 progress on ITER port plug test facility

To achieve the overall ITER machine availability target, the availability of diagnostics and heating port plugs shall be as high as 99.5%. To fulfill these requirements, it is mandatory to test the port plugs at operating temperature before installation on the machine and after refurbishment.The ITER port plug test facility (PPTF) provides the possibility to test upper and equatorial port plugs before installation on the machine. The port plug test facility is composed of several test stands. These test stands are first used in the domestic agencies and on the ITER Organization site to test the port plugs at the end of manufacturing. Two of these stands are installed later in the ITER hot cell facility to test the port plugs after refurbishment. The port plugs to be tested are the Ion Cyclotron (IC) heating and current drive antennas, Electron Cyclotron (EC) heating and current drive launchers, diagnostics and test blanket modules port plugs.Test stands shall be capable to perform environmental and functional tests. The test stands are composed of one vacuum tank (3.3 m in diameter, 5.6 m long) and the associated heating, vacuum and control systems. The vacuum tank shall achieve an ultimate pressure of 1 × 10^?5 Pa at 100 °C containing a port plug. The heating system shall provide water at 240 °C and 4.4 MPa to heat up the port plugs. Openings are provided on the back of the vacuum tank to insert probes for the functional tests.This paper describes the tests to be performed on the port plugs and the conceptual design of the port plug test facility. The configuration of the standalone test stands and the integration in the hot cell facility are presented.     

Thermo-hydraulic design verification of the neutral beam liner for the ITER vacuum vessel

The ITER vacuum vessel has upper, equatorial and lower port structures used for equipment installation, utility feedthroughs, vacuum pumping and access inside the vessel for maintenance. A neutral beam (NB) port of equatorial ports provides a path of neutral beam for plasma heating and current drive. An internal duct liner is built in the NB ports, and copper alloy panels are placed in the top and bottom of the liner to protect duct surface from high-power heat loads. Global NB liner models for the upper panel and the lower panel have been developed, and flow field and conjugate heat transfer analyses have been performed to find out pressure drop and heat transfer characteristics of the liner. Heat loads such as NB power, volumetric heating and surface heat flux are applied in the analyses for beam steering and misalignment conditions. For the upper panel, it is found that unbalanced flow distribution occurs in each flow path, and this causes poor heat transfer performance as well. In order to improve flow distribution and to reduce pressure losses, hydraulic analyses for modified cooling path schemes have been carried out, and design recommendations are made based on the analysis results. For the lower panel, local flow distributions and pressure drop values at each header and branch of the tube are obtained by applying design coolant flow rate. Together with the coolant flow field, temperature and heat transfer coefficient distributions are also acquired from the analyses. Based on the analysis results, it is concluded that the lower panel for the NB liner is relatively well designed even though the given heat loads are very severe.     

Recent developments in the structural design and optimization of ITER neutral beam manifold

This paper describes a new design of the neutral beam manifold based on a more optimized support system.A proposed alternative scheme has presented to replace the former complex manifold supports and internal pipe supports in the final design phase.Both the structural reliability and feasibility were confirmed with detailed analyses.Comparative analyses between two typical types of manifold support scheme were performed.All relevant results of mechanical analyses for typical operation scenarios and fault conditions are presented.Future optimization activities are described,which will give useful information for a refined setting of components in the next phase.     

BEPC单粒子试验束的数据获取

阐述了Linux操作系统下BEPC电子直线加速器上单粒子试验束数据获取系统的硬件组成及软件设计与实现．     

New CAMAC based data acquisition and beam control system for Lund nuclear microprobe

A new CAMAC based data acquisition system has been installed at the Lund Nuclear Microprobe facility. This paper reports on the development and present status of the data acquisition system. The system is a true multiparameter CAMAC based system with fast Fera bus readout and in crate memory buffer. The user interface is based on Sparrow Kmax software for a Power Macintosh platform. The system read out and tag the event data with position on-line, which make fast on-line monitoring of spectra or element maps possible.Simultaneously, all data can be saved event by event for off-line analysis. The beam scanning part is software controlled through a timed D/A converter, this allows fast scanning of the beam. With a CCD-camera and video card the area to be analysed could be defined directly from the image, and the sample position can be moved. Any kind of irregular scan patterns could be defined.     

BEPC试验束数据获取系统改进及离线数据分析

运行在BEPC电子直线加速器上的试验束,主要利用在线探测器和离线数据分析选择单粒子事例,满足多种单粒子束流试验.本文介绍了试验束改进后的数据获取系统,以及离线数据分析中的飞行时间谱分析,单粒子判选和粒子击中位置定位等.     

ITER氚增殖实验包层设计研究进展

国际热核实验反应堆(ITER)为人类开发聚变能提供重要的物理和工程技术实验平台,ITER氚增殖实验包层模块(TBM)技术是必须掌握的关键技术.参与ITER计划的成员国根据本国商用演示堆包层发展策略,分别提出了各自的实验包层概念,以便在ITER运行期间进行实验.本文对ITER-TBM目前已经开展和正在进行的主要设计研究工作进展进行总结,介绍了各方提出的设计方案、支撑设计的相关技术研究进展,以及合作实验窗口的分配现状.     

ITER数据归档系统的应用

国际热核实验堆(International Thermonuclear Experimental Reactor,ITER)是世界上最大的超导托卡马克装置,其中央控制系统CODAC(Control,Data Access and Communication)为ITER装置及其各个子系统提供了丰富的控制系统设计与开发工具。ITER极向场整流器系统作为其子系统之一在运行的过程中会产生大量的实验数据,大量数据的高速采集给数据的存储造成很大的困难,其控制系统需要及时地将采集到的数据进行有效的存储,并提供相应数据的查看方式。本文采用CODAC提供的数据归档系统对极向场控制系统数据进行有效的管理。在充分了解该数据归档系统的基础上,对该数据归档系统的使用和可用性进行了详细的介绍。     

Progress in development and design of the neutral beam injector for JT-60SA

Neutral beam (NB) injectors for JT-60 Super Advanced (JT-60SA) have been designed and developed. Twelve positive-ion-based and one negative-ion-based NB injectors are allocated to inject 30 MW D⁰ beams in total for 100 s. Each of the positive-ion-based NB injector is designed to inject 1.7 MW for 100 s at 85 keV. A part of the power supplies and magnetic shield utilized on JT-60U are upgraded and reused on JT-60SA. To realize the negative-ion-based NB injector for JT-60SA where the injection of 500 keV, 10 MW D⁰ beams for 100 s is required, R&Ds of the negative ion source have been carried out. High-energy negative ion beams of 490–500 keV have been successfully produced at a beam current of 1–2.8 A through 20% of the total ion extraction area, by improving voltage holding capability of the ion source. This is the first demonstration of a high-current negative ion acceleration of >1 A to 500 keV. The design of the power supplies and the beamline is also in progress. The procurement of the acceleration power supply starts in 2010.     

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.