A small,novel, remote in-vessel inspection system for the Alcator C-Mod tokamak

A small robust system has been constructed for in-situ visual inspection of the Alcator C-Mod tokamak. The system consists of a small, light, wide-angle high definition camera and LED package housed in a nacelle on the end of thin, rigid, 3.5 m long support pole. The nacelle has two actuated degrees of freedom allowing the camera to observe nearly 4π steradians. The support pole has a specific slight curve that allows it to pass to either side of the center column of the tokamak to observe the entirety of the vessel interior, while still fitting through the small aspect ratio Alcator C-Mod vacuum port structure. The support pole and camera can enter the vessel through any horizontal vacuum port with an inner diameter greater than 4 cm, thus a dedicated port is not required. The inspection is typically undertaken during maintenance periods when the vessel is filled with a noble gas near atmospheric pressure thus minimizing the influx of water vapor and the concomitant loss of wall conditioning. The system is operated manually, producing photos and video which are reviewed in near real-time. Nearly the entire vessel, including the plasma facing components, can be carefully inspected in 3–5 h. The system provides improved characterization of the interior components and surfaces of the tokamak with a modest engineering and operational effort. Information gathered from the system has identified damage to plasma facing components that were interfering with tokamak operation, as well as damage to mechanical components which were redesigned during the remainder of the campaign, thereby enhancing program planning.     

Power supplies for plasma column control in the COMPASS tokamak

The main magnetic fields in COMPASS – i.e. The Toroidal, Magnetising, Equilibrium, and Shaping Fields – are created by a set of four corresponding thyristor power supplies controlled in a 0.5 ms loop.The plasma position has to be controlled both radially and vertically by two additional magnetic fields provided by two fast amplifiers (FAs) based on MOSFET technology, each supplying ±100 V and up to ±5 kA.Currently, an ongoing project aims at ELM triggering by fast changes of the vertical position of the plasma column, also referred to as vertical kicks. For this purpose, a new Vertical Kicks Power Supply (VKPS) capable of quick change of vertical plasma position is being constructed. This power supply should operate at up to 1.2 kV with switching frequency up to 5 kHz. It is designed as a H-bridge but based on IGBT transistors which can be operated at higher voltages than MOSFETs.In this paper, we focus on the FAs and VKPS engineering design and required output parameters. Both the power supplies are based on modern components with highest available ratings in their categories. Unique design of the power supplies takes advantage of the short duration of the COMPASS discharge by overloading the transistors above their maximal steady-state rating.The FA is regularly operating, so that in addition to describing its design, we also describe the achieved performance parameters. Finally, the common controller unit, communication, and error handling is described.     

Rationale for nuclear reactor control room data displays

One of the ‘lessons learned’ from the Three Mile Island accident focuses upon the need for a validated source of plant-status information in the control room. The utilization of computer-generated graphics to display the reduced readings of the plant instrumentation has introduced the need for a set of guidelines that focuses upon the mental image of plant conditions that the operators receive from viewing the computer monitors. The principles that govern the design of displays are similar to those employed in the education process because the objective is the same, namely, to transfer information.The philosophy for the development of displays to portray the status of major plant systems bases the level of detail upon the needs of the user. Graphic displays that relate the plant system parameter to the plant systems are recommended along with bar-chart type representations of the relationship between plant parameters and allowable limits.     

Water hydraulic manipulator for fail safe and fault tolerant remote handling operations at ITER

Department of Intelligent Hydraulics and Automation (IHA) of Tampere University of Technology has been involved in the European Fusion program since 1994 within the ITER reactor maintenance activities. In this paper we discuss the design and development of a six degrees of freedom water hydraulic manipulator with a force feedback for teleoperation tasks. The manipulator is planned to be delivered to Divertor Test Platform 2 (DTP2) during year 2008. The paper also discusses the possibility to improve the fail safe and redundant operation of the manipulator. During the design of the water hydraulic manipulator, special provisions have been made in order to meet the safety requirements such as servo valve block for redundant operation and safety vane brakes for fail safe operation.     

核电厂数字化控制室界面管理任务对控制室运行人员的影响

讨论了数字化技术引入核电站控制室后,控制室人机界面管理任务对控制室运行人员和核电站安全产生的不良影响,以及在设计中如何考虑并减轻其不良影响。     

A 600 MWe pulsed tokamak reactor

The choice of parameters for a pulsed tokamak reactor with net electrical output of 600 MW_e is considered. The physics and engineering assumptions are described in detail with particular emphasis on the magnetic field design. This leads to reactor parameters with 2 m minor radius, an aspect ratio of 3.9, and energy multiplication factorQ of 13. The sensitivity of the chosen parameters to various changes in the assumptions is analyzed. A comparison is made with a recent study of the reversed field pinch reactor, showing that the two containment systems appear to have similar potential as fusion power sources.     

Design and realization of the J-TEXT tokamak central control system

The Joint Texas Experimental Tokamak (J-TEXT), a medium-sized conventional tokamak, serves as a user experimental facility in the China-USA fusion research community. Development of a flexible and easy-to-use J-TEXT central control system (CCS) is of supreme importance for users to coordinate the experimental scenarios with full integration into the discharge operation. This paper describes in detail the structure and functions of the J-TEXT CCS system as well as the performance in practical implementation. Results obtained from both commissioning and routine operations show that the J-TEXT CCS system can offer a satisfactory and effective control that is reliable and stable. The J-TEXT tokamak achieved high-quality performance in its first-ever experimental campaign with this CCS system.     

Cluster storage for COMPASS tokamak

The COMPASS tokamak is expected to produce several gigabytes of data per shot in near future. A new storage system is needed to accommodate and access all the data. It should be scalable, fault-tolerant, and parallel. It should not be based on proprietary solutions to maintain independence from hardware and software manufacturers and preferably it should be built on inexpensive commodity hardware. One of the promising distributed parallel fault-tolerant file systems, GlusterFS, was selected for testing. The aim of the work was to make initial tests of a particular small GlusterFS setup to confirm its aptitude for the COMPASS storage system. Aggregated reading throughput from multiple NFS clients was one of the most important figures that were benchmarked, it scaled well with the number of clients, starting just above 60 MiB/s¹ for 1 client and going slightly over 300 MiB/s for 6 clients.     

Upgrade of plasma density feedback control systemin HT-7 tokamak

The HT-7 is a superconducting tokamak in China used to make researches on the controlled nuclear fusion as a national project for the fusion research. The plasma density feedback control subsystem is the one of the subsystems of the distributed control system in HT-7 tokamak (HT7DCS). The main function of the subsystem is to control the plasma density on real-time. For this reason, the real-time capability and good stability are the most significant factors, which will influence the control results. Since the former plasma density feedback control system (FPDFCS) based on Windows operation system could not fulfill such requirements well, a new subsystem has to be developed. The paper describes the upgrade of the plasma density feedback control system (UPDFCS), based on the dual operation system (Windows and Linux), in detail.     

A new Q-band comb-based multi-channel microwave Doppler backward scattering diagnostic developed on the HL-3 tokamak

The Doppler backscattering(DBS) diagnostic is widely used to measure the localized density fluctuations and the propagation velocity of turbulent structures. Microwave is launched at a frequency that approaches a cutoff layer in the plasma at an angle oblique to the cutoff layer. A new Q-band multichannel DBS system based on a comb generator has been designed and tested for application on the HL-3 tokamak. With the comb generator and heterodyne scheme, the stability and flexibility of the new DBS system are improved. The new DBS diagnostic has a high output power(～ 10 dBm), good power flatness(< 5 dB in Q-band), and frequency stability,and the inter-frequency separation is tunable remotely. This article introduces the system design,laboratory test results, and initial experimental results from the HL-3 tokamak. With the help of the newly developed multichannel DBS, the turbulence information can be studied with high temporal and spatial resolution.     

A multi-channel real-time digital integrator for magnetic diagnostics in HL-2A tokamak

A novel full-digital integrator has been developed for the magnetic diagnostics in HL-2A. Based on the pipeline processing of the field-programmable gate array and high-speed PCI extensions for instrumentation platform, the digital integrator has realized octal-channel10-k Hz real-time integration and data transmission. In order to reduce the integration drift, a 24-bit analog-todigital converter and simple analog processing circuits are applied for high-precision sampling, while certain correction algorithms are used to minimize the drift. With simple and highly integrated circuits and high-performance digital processor, the digital integrator is of high stability and functional expansibility which greatly simplifies the operation procedure. The digital integrator has been tested in the plasma discharge experiments, and the experimental results have confirmed that the drift performance and accuracy of the digital integrator could fully meet the requirements of HL-2A.     

Technical diagnosis system for EAST tokamak

Technical diagnosis system (TDS) is one of the important subsystems of EAST (experimental advanced superconducting tokamak) device, main function of which is to monitor status parameters in EAST device. Those status parameters include temperature of different positions of main components, resistance of each superconducting (SC) coils, joint resistance of SC coils and high-temperature superconducting (HTS) current leads, strain of cold-quality components endured force, and displacement and current of toroidal field (TF) coils in EAST device, which are analog input signals. In addition there are still some analog and digital output signals. The TDS monitors all of those signals in the period of EAST experiments. TDS data monitoring is described in detail for it plays important role during EAST campaign. And how to protect the SC magnet system during each plasma discharging is presented with data of temperature of coolant inlet and outlet of SC coils and feeders and cases of the TF coils and temperature in the upper and middle and bottom of the TF coil case.During construction of the TDS primary difficulties come from installation of Lakeshore Cernox temperature sensors, strain measurement of central solenoid coils support legs and installation of co-wound voltage sensors for quench detection. While during operation since the first commissioning big challenges are from temperature measurement changes in current leads and quench detection of PF coils. Those difficulties in both stages are introduced which are key to make the TDS reliable. Meanwhile analysis of experimental data like temperature as a back up to testify quench occurrence and stress on vacuum vessel thermal shield and vacuum vessel have also been discussed.     

A functional approach for managing ITER operations

ITER is currently the most ambitious project on nuclear fusion research. Its objective is to demonstrate the feasibility of fusion as an energy source for the future. The complexity of the systems required to meet this challenge present many opportunities for omissions or incorrect assumptions. System engineering allows the engineer to deal with such a complexity by developing a Functional Breakdown Structure (FBS). Unlike a Plant Breakdown Structure (PBS), the FBS is a function-oriented tree, not a product-oriented tree. It details operations or activities that have to be performed as needed functions of the architecture, allowing identification of any missing elements, defining the personnel skills required to operate the architecture and managing the machine availability.     

Module of lithium divertor for KTM tokamak

Activity on projects of ITER and DEMO reactors has shown that solution of problems of divertor target plates and other plasma facing elements (PFEs) based on the solid plasma facing materials cause serious difficulties. Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium–metal with low Z. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life; to reduce power flux due to intensive re-irradiation on lithium atoms in plasma periphery that will essentially facilitate a problem of heat removal from PFE; to reduce Z_eff of plasma to minimally possible level close to 1; to exclude tritium accumulation, that is provided with absence of dust products and an opportunity of the active control of the tritium contents in liquid lithium. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) – new material in which liquid lithium fill a solid matrix from porous material. The progress in development of lithium technology and also activity in lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, HT-7 and stellarator TJ II permits of solving the problems in development of steady-state operating lithium divertor module project for Kazakhstan tokamak KTM. At present the lithium divertor module for KTM tokamak is under development in the framework of ISTC project # K-1561. Initial heating up to 200 °C and lithium surface temperature stabilization during plasma interaction in the range of 350–550 °C will be provided by external system for thermal stabilization due to circulation of the Na–K heat transfer media. Lithium filled tungsten felt is offered as the base plasma facing material of divertor. Development, creation and experimental research of lithium divertor model for KTM will allow to solve existing problems and to fulfill the basic approaches to designing of lithium divertor and in-vessel elements of new fusion reactor generation, to investigate plasma physics aspects of lithium influence, to develop technology of work with lithium in tokamak conditions. Results of this project addresses to the progress in the field of fusion neutrons source and fusion energy source creation.     

Limiting heat loads for graphite tokamak limiters

Translated from Atomnaya Énergiya, Vol. 69, No. 3, pp. 187–188, September, 1990.     

Assembly study for JT-60SA tokamak

The assembly scenarios and assembly tools of the major tokamak components for JT-60SA are studied in the following. (1) The assembly frame (with a dedicated 30-tonne crane), which is located around the JT-60SA tokamak, is adopted for effective assembly works in the torus hall and the temporary support of the components during assembly. (2) Metrology for precise positioning of the components is also studied by defining the metrology points on the components. (3) The sector segmentation for weld joints and positioning of the vacuum vessel (VV), the assembly scenario and tools for VV thermal shield (TS), the connection of the outer intercoil structure (OIS) and the installation of the final toroidal field coil (TFC) are studied, as typical examples of the assembly scenarios and tools for JT-60SA.