中国洪水管理战略框架和行动计划

“中国洪水管理战略研究”项目,针对中国洪水管理现状,从中国国情出发,广泛吸收和借鉴国内外洪水管理的成功经验,对洪水管理战略与对策进行了研究,并在深入调研的基础上提出了中国洪水管理战略框架和行动计划设想,以满足全社会日益提高的水安全保障需求,推进由控制洪水向洪水管理的战略性转变,逐步实现以科学发展观和风险管理理论为指导的洪水综合管理。     

The Electron Temperature Estimation Using Soft X-Ray Imaging in HT-7 Tokamak

In order to estimate the electron temperature soft x-ray imaging diagnostics using a double filter technique has been developed in the HT-7 tokamak. The chosen thicknesses of the Be foil are 12.5 μm and 70 μm, respectively. In this article both the main design of the diagnostic configuration and the method to estimate the electron temperature are presented. The results agree with those estimated from the soft x-ray pulse height analyzer （PHA）. The main causes of systematic error have also been investigated.     

Structural materials for fusion reactors

Fusion Reactors will require specially engineered structural materials, which will simultaneously satisfy the harsh conditions such as high thermo mechanical stresses, high heat loads and severe radiation damage without compromising on safety considerations. The fundamental differences between fusion and other nuclear reactors arise due to the 14MeV neutronics of structural materials. There exists considerable uncertainty in the nuclear data at such energies because there aren’t any strong enough sources for such neutrons except fusion reactors themselves! We thus encounter a problem of iterative nature in which we must try several experiments with the available materials in the near term. The development of such structural materials is thus going to require the experimental data of the kind that may be generated on reactors like ITER, high-performance modeling and a penetrating metallurgical insight to overcome technological challenges in terms of achieving required properties such as low activation by controlling the impurities, good thermo-mechanical properties by microstructure engineering, good chemical compatibility and high radiation resistance. These materials need to withstand a neutron wall load of the order of 2–3 MW/m², which can lead up to 30 dpa of radiation damage and 300 appm helium production per full power year in DEMO like reactors. Such conditions lead to unprecedented events related to the failure of materials due to irradiation creep, Ductile-Brittle Transition Temperature (DBTT) shift and helium embrittlement. The development of fusion materials program is oriented towards fulfilling the requirements of Test Blanket Modules, various prototype activities of SST-2 and DEMO reactor. The materials identified for first wall and blanket modules for Indian DEMO are LAFMS and ODS steels. The development program plan for these materials include (i) Manufacturing of LAFMS steel through VIM/VAR methods by controlling the impurities such as S, P and Si. (ii) ODS steel development with nano-size Y₂O₃ dispersoids in ferritic martensitic matrix by powder metallurgy route. The advanced structural materials like SiC_f /SiC composites and SiC_f /n-SiC are planned under National Fusion Program projects for indigenous development. An overview of the planned program in this direction will be presented.     

Present Status of JT-60SA Project and Development of Heating Systems for JT-60SA

Present status of the JT-60SA (JT-60 Super Advanced) project, implemented jointly by Europe and Japan since 2007, is described. The design of the main tokamak components was completed in late 2008, and all the scientific missions are preserved to contribute to ITER and DEMO reactors. The construction of the JT-60SA has begun with procurement activities for the superconducting magnet systems, vacuum vessel, in-vessel components and other components under the relevant procurement arrangements between the implementing agencies of JAEA (Japan Atomic Energy Agency) in Japan and Fusion for Energy in Europe. Designs and developments of the auxiliary heating systems for JT-60SA have been progressing at JAEA so as to provide the total injection power of 41 MW for 100 s.     

聚变裂变混合发电堆水冷包层热工水力学设计分析

一种以能量倍增为目标的聚变裂变混合发电堆(FDS-EM)概念已被提出,FDS-EM初步设计为可以产生约1.0 GW的电功率,并能实现氚自持。本文对FDS-EM水冷包层进行了热工水力学设计与分析。设计采用了压水堆的成熟技术,并给出了典型的热工设计参数,通过对典型参数下包层的数值模拟分析,得出了温度场和应力场分布,初步证明了设计的工程可行性。     

EAST钨杂质输运特性分析

用基于蒙特卡罗方法的杂质输运程序DIVIMP模拟了EAST在加热功率Pin=8 MW的情况下边界层钨杂质输运过程并分析了钨杂质输运特性,其目的是为了合理设计第一壁和偏滤器靶板部件。DIVIMP模拟得到了低再循环、高再循环和脱靶运行条件下钨杂质在边界层的分布情况。模拟结果显示,钨在脱靶情况下所产生的杂质密度比其他运行模式所产生的杂质密度小。与碳杂质输运特性的模拟结果相比,钨杂质的密度有明显的下降,总体分布在1015m-3数量级上。     

Implications of NSTX lithium results for magnetic fusion research

Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment (NSTX) for the last five years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a lithium evaporation system which can evaporate up to ∼100 g of lithium onto the lower divertor plates between lithium re-loadings. The unique feature of the lithium research program on NSTX is that it can investigate the effects of lithium in H-mode divertor plasmas. This lithium evaporation system thus far has produced many intriguing and potentially important results; the latest of these are summarized in a companion paper by H. Kugel. In this paper, we suggest possible implications and applications of the NSTX lithium results on the magnetic fusion research which include electron and global energy confinement improvements, MHD stability enhancement at high beta, edge localized mode (ELM) control, H-mode power threshold reduction, improvements in radio frequency heating and non-inductive plasma start-up performance, innovative divertor solutions and improved operational efficiency.     

Characterization of plasma current quench during disruptions at HL-2A

The most essential assumptions of physics for the evaluation of electromagnetic forces on the plasma-facing components due to a disruption-induced eddy current are characteristics of plasma current quenches including the current quench rate or its waveforms.The characteristics of plasma current quenches at HL-2A have been analyzed during spontaneous disruptions.Both linear decay and exponential decay are found in the disruptions with the fastest current quenches.However,there are two stages of current quench in the slow current quench case.The first stage with an exponential decay and the second stage followed by a rapid linear decay.The faster current quench rate corresponds to the faster movement of plasma displacement.The parameter regimes on the current quench time and the current quench rates have been obtained from disruption statistics at HL-2A.There exists no remarkable difference for distributions obtained between the limiter and the divertor configuration.This data from HL-2A provides basic data of the derivation of design criteria for a large-sized machine during the current decay phase of the disruptions.     

武钢转炉炉龄攻关总结

从本厂转炉实际情况出发,针对生产中的薄弱环节,有计划进行转炉炉龄攻关。重点在提高镁碳砖实物质量、开发复合镁碳砖和整体出钢口砖、细化综合砌筑、改进冶炼操作、采用挂渣护炉、控制末期渣成分等方面进行攻关。由于各级领导重视,各专业人员通力合作,攻关项目得以落实,创2077炉本厂历史最高炉龄和全年平均炉龄1602炉的历史最好水平。