嵌入式Linux系统在PowerPC上的实现

介绍了一个在PowerPC上建立嵌入式Linux操作系统的应用实例.该系统的主要目的是用于VME总线设备的测试,也可用于小型数据获取与控制系统.建立系统的过程中充分考虑到软件开发以及PowerPC的特点,为开发者提供两种可选的程序编译测试方法.介绍了几种典型嵌入式操作系统,比较商业嵌入式系统与主流Linux嵌入式系统的异同以及它们的优缺点;就如何根据特定目标板建立一个可裁减的嵌入式Linux操作系统做了详细的阐述;给出了一个可以成功运行的测试实例.     

VME测试系统在嵌入式Linux上的实现

介绍了基于VME总线的测试系统,该测试系统采用PowerPC作为VME机箱的读出插件、嵌入式Linux为操作系统,实现电子学小系统测试的相关功能.首先介绍了系统相关硬件和软件框架及本系统的关键技术和各项功能模块的实现,最后介绍了图形用户界面.该系统可以用于基于VME总线的系统测试以及初步的系统控制.     

An FEL-based microwave system for fusion

This paper describes designs for 280-GHz and 560-GHz microwave sources based on free electron lasers (FELs). These 10-MW units are based on technology developed over the last 5 years. A first demonstration of high-average-power microwave production with an FEL system is expected in the Microwave Tokamak Experiment (MTX) facility. This paper gives details on the design and construction of that 250-GHz, 2-MW system and discusses specific applications for the Compact Ignition Tokamak (CIT).     

基于ARM的嵌入式Linux网络数据传输性能分析

大型高能物理实验数据采集系统普遍采用多级触发结构,各级间用网络将数据发送到更高级别的系统。以太网技术多作为触发数据传输的媒介。为提高系统传输效率,找到系统数据传输的优化方法,本文基于ARM平台,从数据流角度阐述了Linux2.4内核网络数据的接收过程,并结合测得的内核内部数据分析了网络性能变化的具体原因,给出优化系统性能的方向。     

聚变中子源驱动的次临界清洁核能系统─聚变能技术的早期应用途径

提出作为聚变能技术早期应用途径的聚变中子源驱动的清洁核能系统概念,并从国家的能源需求、国内外核电发展状况论述开发这种系统的必要性和意义,根据国内外聚变驱动器技术及次临界包层技术进展和国内多年的可行性研究结果,说明开发这种系统的现实性和基础。文中也给出了建议的发展进程。     

聚变中子源驱动的次临界清洁核能系统──聚变能技术的早期应用途径

提出作为聚变能技术早期应用途径的聚变中子源驱动的清洁核能系统概念,并从国家的能源需求、国内外核电发展状况论述开发这种系统的必要性和意义,根据国内外聚变驱动器技术及次临界包层技术进展和国内多年的可行性研究结果,说明开发这种系统的现实性和基础。文中也给出了建议的发展进程。     

基于FPGA的电子直线加速器低电平系统前馈功能的实现

用于驱动光中子源装置(TMSR Photo-Neutron Source Phase1,TDSN1)的15 Me V直线加速器,由于腔体中瞬态束流负载效应的存在使得束团在经过腔体后头部的能量过高,会导致束流能散变大,降低了束流的传输效率。"数字前馈补偿"方法在原有的数字低电平控制系统的现场可编程逻辑门阵列(Field Programmable Gate Array,FPGA)中加入前馈模块,通过直接削弱输入腔体的射频场的头部场强来达到降低束团头部能量的目的。实际数据表明,低电平系统前馈功能工作稳定,束流能散降低,束流的传输效率显著变高,克服了传统束流补偿法不能在大束团、高流强模式下工作的缺陷。     

The role of nuclear data for fusion technology studies

Nuclear data are of fundamental importance in studies of nuclear technology. In these studies, experiments to measure cross sections and decay properties and simulations of the design of fission power plants, fusion devices and accelerators are included. The large amount of data required is stored in computer readable formats in data libraries and the most common of these are the general purpose files used for neutronics or transport calculations. These files also contain the standards against which most measurements are made. The other class of libraries are the special purpose ones containing decay data, fission yields and cross section data for dosimetry and activation.This paper gives examples of what data are available and describes their use for various fusion applications. The focus will be on neutron-induced activation data with examples of how the reactions of particular importance can be identified. All data should be accompanied by estimates of the uncertainty. This is best achieved by including covariance data; however, this is extremely challenging and only a subset of the available data has such uncertainty data. The general principles of how covariance matrices are used are outlined.     

Fast magnetic field computation in fusion technology using GPU technology

One of the main issues in the simulation of Tokamaks functioning is the reliable and accurate computation of actual field maps in the plasma chamber. In this paper a tool able to accurately compute magnetic field maps produced by active coils of any 3D shape, wound with high number of conductors, is presented. Under linearity assumption, the coil winding is modeled by means of “sticks”, following each conductor's shape, and the contribution of each stick is computed using high speed Graphic Computing Units (GPU's). Relevant speed enhancements with respect to standard parallel computing environment are achieved in this way.     

Linux下基于Root界面的通用探测器测试系统

介绍了一个在Linux操作系统下实现的基于Root图形界面的通用探测器测试系统。该系统可以根据不同的实验要求任意设置,实现对高压的控制,实现计数率曲线和效率曲线的自动化测量,以及在不同高压下的信号幅度、电荷量和时间量等的自动测量。     

The ARIES-AT advanced tokamak,Advanced technology fusion power plant

The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant and to identifying physics and technology areas with the highest leverage for achieving attractive and competitive fusion power in order to guide fusion R&D. The 1000-MWe ARIES-AT design has a major radius of 5.2 m, a minor radius of 1.3 m, a toroidal β of 9.2% (β_N = 5.4) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current-drive power is 35 MW. The ARIES-AT design uses the same physics basis as ARIES-RS, a reversed-shear plasma. A distinct difference between ARIES-RS and ARIES-AT plasmas is the higher plasma elongation of ARIES-AT (κ_x = 2.2) which is the result of a “thinner” blanket leading to a large increase in plasma β to 9.2% (compared to 5% for ARIES-RS) with only a slightly higher β_N. ARIES-AT blanket is a simple, low-pressure design consisting of SiC composite boxes with a SiC insert for flow distribution that does not carry any structural load. The breeding coolant (Pb–17Li) enters the fusion core from the bottom, and cools the first wall while traveling in the poloidal direction to the top of the blanket module. The coolant then returns through the blanket channel at a low speed and is superheated to ∼1100 °C. As most of the fusion power is deposited directly into the breeding coolant, this method leads to a high coolant outlet temperature while keeping the temperature of the SiC structure as well as interface between SiC structure and Pb–17Li to about 1000 °C. This blanket is well matched to an advanced Brayton power cycle, leading to an overall thermal efficiency of ∼59%. The very low afterheat in SiC composites results in exceptional safety and waste disposal characteristics. All of the fusion core components qualify for shallow land burial under U.S. regulations (furthermore, ∼90% of components qualify as Class-A waste, the lowest level). The ARIES-AT study shows that the combination of advanced tokamak modes and advanced technology leads to an attractive fusion power plant with excellent safety and environmental characteristics and with a cost of electricity (4.7 ¢/kWh), which is competitive with those projected for other sources of energy.     

核电站稳压器水位控制系统数字化的实现及优化方案

根据模拟稳压器水位控制系统结构、功能以及运行原理的分析,展开对稳压器水位控制系统数字化的研究。在选择成熟商用单板机的基础上引入了数/模,模/数转换以及控制算法等实现了稳压器水位控制系统,并和稳压器水位仿真模型构成闭环运行,通过虚拟仪器监测此控制系统的运行情况,证实了本系统达到并超过了原有模拟控制系统的功能和性能。     

The high-density Z-pinch as a pulsed fusion neutron source for fusion nuclear technology and materials testing

The dense Z-pinch (DZP) is one of the earliest and simplest plasma heating and confinement schemes. Recent experimental advances based on plasma initiation from hair-like (10s m in radius) solid hydrogen filaments have so far not encountered the usually devastating MHD instabilities that plagued early DZP experimenters. These encouraging results along with the debut of a number of proof-of principle, high-current (1–2 MA in 10–100 ns) experiments have prompted consideration of the DZP as a pulsed source of DT fusion neutrons of sufficient strength (S _N 10¹⁹ n/s) to provide uncollided neutron fluxes in excess ofI _w = 5–10 MW/m² over test volumes of 10–30 liters or greater. While this neutron source would be pulsed (100s ns pulse widths, 10–100 Hz pulse rate), giving flux time compressions in the range 10⁵–10⁶, its simplicity, near-term feasibility, low cost, high-Q operation, and relevance to fusion systems thatmay provide a pulsed commercial end-product, e.g., inertial confinement or the DZP itself, together create the impetus for preliminary consideration as a neutron source for fusion nuclear technology and materials testings. The results of a preliminary parametric systems study (focusing primarily on physics issues), conceptual design, and cost vs. performance analyses are presented. The DZP promises an inexpensive and efficient means to provide pulsed DT neutrons at an average rate in excess of 10¹⁹ n/s, with neutron currents I_w10 MW/m² over volumes V_exp 30 liter using single-pulse technologies that differ little from those being used in present-day experiments.Work supported by U.S. DOE.     

FPGA横向反馈数字板在SSRF储存环的应用

上海光源（SSRF）正在构建基于数字信号处理模块的横向反馈系统。FPGA（Field Programmable Gate Array）结构可以调节数字板FIR滤波器的参数,具有较好的灵活性。图形用户界面（GUI,Graphical User Interface）程序用来实现滤波器设计,并生成供数字板下载的逗号分隔符.csv（Comma Separated Value）文件。我们通过调节横向工作点处增益和相位,使滤波效果达到最优化。本文介绍了SSRF横向反馈对单束团束流情况下的实验结果,已能够证明数字板调节的灵活性。系统在多束团情况下的研究将要进行。     

The renewed HT-7 plasma control system based on real-time Linux cluster

In order to improve the synchronization, flexibility and expansibility of the plasma control on HT-7, a new plasma control system (HT-7 PCS) was constructed. The HT-7 PCS was based on a real-time Linux cluster with a well-defined, robust and flexible software infrastructure which was adapted from DIII-D PCS. In this paper, the hardware structure and system customization details for HT-7 PCS are reported. The plasma position and current control, plasma density control and off-normal event detection, which were realized in separated systems originally, have been integrated and implemented in such HT-7 PCS. All these control algorithms have been successfully validated in the last several HT-7 experiment campaigns. Good control performance has been achieved and the experiment results are discussed in the paper.     

基于FPGA的高速串行光纤数据传输在BESIII触发系统的应用

基于FPGA内置的RocketIO高速串行收发器设计的高速串行光纤数据传输的新方法具有通道少、数据量大、传输距离远等优点,能满足现代高能物理实验的需要.本文对这种传输方法的多通道、实时数据传输进行了深入的研究,并将其应用到北京谱仪III(BESIII)触发系统中.测试结果表明:数据传输的长时间误码率好于2.3(10-16,多个通道能按照要求进行同步数据传输.该方案已经过长时间的运行考验,达到了BESIII工程的设计要求.     

OPC技术在EPICS系统中的应用

简要介绍了工控领域的OPC（OLE for process control）技术,并介绍了OPC在EPICS系统中的应用及EPICS／OPC服务器的开发。     

Developments and needs in nuclear analysis of fusion technology

Nuclear analyses provide essential input to the conceptual design, optimisation, engineering and safety case of fusion technology in current experiments, ITER, next-step devices and power plant studies. Calculations are intricate and computer-intensive, typically requiring detailed geometry models, sophisticated acceleration algorithms, high-performance parallel computations, and coupling of large and complex transport and activation codes and databases. This paper reports progress on some key areas in the development of tools and methods to meet the specific needs of fusion nuclear analyses. In particular, advances in the production and modernisation of reference models, in the preparation and quality assurance of acceleration algorithms and coupling schemes, and in the evaluation and adaptation of alternative transport codes are presented. Emphasis is given to ITER-relevant activities, which are the main driver of advances in the field. Discussion is made of the importance of efforts in these and other areas, considering some of the more pressing needs and requirements. In some cases, they call for a more efficient and coordinated use of the scarce resources available.