俄罗斯是热核聚变研究资历最深的国家之一,也是世界首个托卡马克 装置的诞生地.从最初的几十年占据世界受控热核聚变研究最先进的位置,到受经济衰退影响萎靡不振,再到21世纪后重振旗鼓,积极通过国际合作振兴本国核聚 变研究事业,俄罗斯的所有努力都为推动世界核聚变事业贡献了力量.介绍了俄罗斯核聚变研究的发展历史、主要研究机构、主要装置沿革以及突出成果,分析了俄 罗斯从最早开始核聚变研究至今各种变化的政治和经济原因以及经费投入,并简要阐述了俄罗斯未来的核聚变战略规划.

The requirement at CERN for 1 kA range High Temperature Superconducting (HTS) cables optimized for long electrical transfer has led to the design and assembly of a novel type of cable that can be made from pre-reacted MgB2, Bi-2223 or YBCO tapes. The cable consists of an assembly of twisted pairs, each of which is made from three superconducting tapes with the required copper stabilizer. The twisted pair cable is designed to transfer a DC current of 600 A in helium gas environment. The paper reports on the results of the electrical tests performed on twisted-pair cables of identical structure and made from commercially available MgB2, Bi-2223 and YBCO tapes. The twist pitch of the cables is adapted to match the mechanical properties of the different superconductors. Critical current tests were performed at both liquid helium and liquid nitrogen temperature. The electrical performance of several cables made from different conductors is reported and compared.

Coated conductor applications such as fusion magnets, particle accelerator magnets and generator windings require high current-carrying capabilities. This requirement can be fulfilled by various cable concepts using commercial long length REBCO coated conductors with high current-carrying performance. In the past few years, our group has successfully developed the Roebel cable concept for coated conductors. The design advantages of such a cable are high current-carrying capability and low alternating current (AC) losses. Unfortunately, for large-scale applications, the possibilities of a simple scale-up of the Roebel geometry are limited and additional design ideas are needed. One way to reach the required high currents is the Rutherford cable concept. In this concept a conductor is wound with transposition on a flat metal former. In order to design the former, the bending properties of the Roebel assembled coated conductor cables (RACC) must be measured and characterized. This allows the identification of a destruction-free interval for the Roebel cable, in terms of bending angle and transposition length. In this work we designed and assembled a demonstrator of a coated conductor Rutherford cable (CCRC) with three RACC cables. We measured the critical current and the AC losses of the cable demonstrator. Our results show that, despite still needing efforts in terms of reproducibility of the assembly process and of AC loss reduction, this design is a promising and viable solution for high current-capacity cables made of coated conductors.

It has long been known that high temperature superconductors (HTS) could have an important role to play in the future of tokamak fusion research. Here we report on first results of the use of HTS in a tokamak magnet and on the progress in design and construction of the first fully-HTS tokamak. (C) 2013 Elsevier B.V. All rights reserved.

The objective of the ITER TBM Program is to provide the first experimental data on the performance of the breeding blankets in the integrated fusion nuclear environment. Such information is essential to design and predict the performance of DEMO and future fusion reactors. It foresees to test six mock-ups of breeding blankets, called Test Blanket Module (TBM), in three dedicated ITER equatorial ports from the beginning of the ITER operation. The TBM and its associated ancillary systems, including cooling system and tritium extraction system, forms the Test Blanket System (TBS) that will be fully integrated in the ITER machine and buildings. This paper describes the main features of the six TBSs that are presently planned for installation and operation in ITER, the main interfaces with other ITER systems and the main aspects of the TBM Program management.

Europe has developed two reference tritium breeder blankets concepts for a DEMO fusion reactor: the Helium-Cooled Lithium–Lead and the Helium-Cooled Pebble-Bed. Both will be tested in ITER under the form of Test Blanket Modules (TBMs). The paper reviews the current status of development and qualification of the EU TBMs functional materials; i.e. ceramic solid breeder materials, beryllium/beryllides multiplier materials and Lithium–Lead liquid metal breeder material Pb–15.7Li. For each functional material the main functional/performance requirements with key qualification issues, current status of the R&D activities and the EU development strategy are presented. In the development strategy major steps considered are listed pointing out importance of the ‘Development/qualification/procurement plan’, currently under elaboration, for definition of a roadmap of further activities aiming at delivery of qualified functional materials to be used in the European TBMs in ITER.

In ITER, as in any tokamak, the first wall and divertor plasma-facing components (PFC) must provide adequate protection of in-vessel structures, sufficient heat exhaust capability and be compatible with the requirements of plasma purity. These functions take on new significance in ITER, which will combine long pulse, high power operation with severe restrictions on permitted core impurity concentrations and which, in addition, will produce transient energy loads on a scale unattainable in today devices. The current ITER PFC design has now reached a rather mature stage following the 2007 ITER Design Review. This paper presents the key elements of the design, reviews the physics drivers, essentially thermal load specifications, which have defined the concept and discusses a selection of material and design issues.

The current magnetic confinement nuclear fusion power reactor concepts going beyond ITER are based on assumptions about the availability of materials with extreme mechanical, heat, and neutron load capacity. In Europe, the development of such structural and armour materials together with the necessary production, machining, and fabrication technologies is pursued within the EFDA long-term fusion materials programme. This paper reviews the progress of work within the programme in the area of tungsten and tungsten alloys. Results, conclusions, and future projections are summarized for each of the programme main subtopics, which are: (1) fabrication, (2) structural W materials, (3) W armour materials, and (4) materials science and modelling. It gives a detailed overview of the latest results on materials research, fabrication processes, joining options, high heat flux testing, plasticity studies, modelling, and validation experiments.

The use of tungsten (W) as material for plasma-facing components (PFM) in fusion devices is reviewed with respect to its plasma and material compatibility under burning plasmas conditions. Fusion-relevant plasma operation with W walls is characterised by the need to operate at high edge densities, no or moderate density peaking, and external tools to control the W transport in the plasma core. Several surface and material issues related with the high particle fluencies in fusion devices needs further R&D but are not considered from present view to seriously limit the use of W as PFM. Reliable control of Edge Localised Modes (ELMs) and disruptions is indispensable for the application of W, both to control the W transport in the edge and to avoid target melting in uncontrolled events which can seriously detoriate the operational performance of the device. For DEMO and reactors, the behaviour of W under large neutron fluencies has to be further clarified and measures must be developed to mitigate degradation of material properties by neutron damage.

Ultra-fine grained tungsten specimens with a grain size from the submicron range to several microns were fabricated by resistance sintering under ultra-high pressure. Transient heat loads were applied on the tungsten specimens at room temperature for a pulsed duration of 5 ms at different power density of 0.22, 0.33, 0.44 and 0.55 GW m 612, respectively, by using an electron beam facility. The crack formations and surface melting behaviors under transient heat load were investigated.

This chapter presents tritium permeation characteristics of various materials used in fusion reactors, including plasma-facing, structural, and barrier materials. The parameters necessary for tritium release calculations for various regions of a fusion reactor are given.

Nanostructured ferritic alloy is a subcategory of oxide dispersion strengthened steels intended for advanced reactor applications. The complex ultrafine grained microstructure of an advanced nanostructured ferritic alloy, as determined by electron microscopy and atom probe tomography, is summarised. Three distinct populations of precipitates were observed: 20-50 nm Ti(N,O,C), 5-10 nm diameter Y2Ti2O7/Y2TiO5 and 1-4 nm diameter Ti,Y, O enriched nanoclusters. The first two populations were predominantly located along grain boundaries together with Cr, W and C segregation. A dense population of nanoclusters was observed both in the grain interior as well as on the grain boundaries. These nanoclusters are highly tolerant to high dose irradiation at elevated temperatures.

Several Fe–14Cr based alloys with varying compositions were processed using a combined route of mechanical alloying and spark plasma sintering. Microstructural characteristics of the consolidated alloys were examined via transmission electron microscopy and atom probe tomography, and mechanical properties evaluated using microhardness testing. Lanthanum oxide (0.5wt.%) was added to Fe–14Cr leading to improvement in microstructural stability and mechanical properties mainly due to a high number density of La–Cr–O-enriched nanoclusters. The combined addition of La, Ti (1wt.%) and Mo (0.3wt.%) to the Fe–14Cr base composition further enhanced the microstructural stability and mechanical properties. Nanoclusters enriched in Cr–Ti–La–O with a number density of 1.4×1024m613 were found in this alloy with a bimodal grain size distribution. After adding Y2O3 (0.3wt.%) along with Ti and Mo to the Fe–14Cr matrix, a high number density (1.5×1024m613) of Cr–Ti–Y–O-enriched NCs was also detected. Formation mechanism of these nanoclusters can be explained through the concentrations and diffusion rates of the initial oxide species formed during the milling process and initial stages of sintering as well as the thermodynamic nucleation barrier and their enthalpy of formation.

ITER包层屏蔽块全尺寸原型 件应基于当前的设计方案,满足其物理功能,并符合包层界面的设计要求。另外,在屏蔽块全尺寸原型件的设计中,还要充分考虑关键制造技术的研发结果,例如深 孔钻、TIG焊接、NDT检测等技术,这些关键制造技术的研发结果,为设计提供了技术保障。该全尺寸原型件的顺利完成并通过ITER相应认证程序,是中方 签署采购的必由环节,也为今后完成采购包奠定了基础。