Effect of 10 MW/m2 plasma exposure on W-SiC/SiC dual layer tiles

W-SiC/SiC dual layer tile has many advantages as a high heat flux component (HHFC) material for fusion, in theory. However, due to insufficient data known, its high potentiality and near term availability has not been well recognized. This work provides the recent materials R&D status and the first plasma exposure test result from the world largest helical device, large helical device of National Institute for Fusion Science in Japan. Tungsten armor with SiC/SiC substrate layer survived during the LHD plasma exposure with 10 MW/m² maximum heat load for the 5.3-s operation cycle. The macro and microstructure evolution, including crack and pore formation, was analyzed and an excellent high heat load resistance was demonstrated.     

Prospects of High Temperature Superconductors for fusion magnets and power applications

During the last few years, progress in the field of second-generation High Temperature Superconductors (HTS) was breathtaking. Industry has taken up production of long length coated REBCO conductors with reduced angular dependency on external magnetic field and excellent critical current density jc. Consequently these REBCO tapes are used more and more in power application.For fusion magnets, high current conductors in the kA range are needed to limit the voltage during fast discharge. Several designs for high current cables using High Temperature Superconductors have been proposed. With the REBCO tape performance at hand, the prospects of fusion magnets based on such high current cables are promising. An operation at 4.5 K offers a comfortable temperature margin, more mechanical stability and the possibility to reach even higher fields compared to existing solutions with Nb₃Sn which could be interesting with respect to DEMO.After a brief overview of HTS use in power application the paper will give an overview of possible use of HTS material for fusion application. Present high current HTS cable designs are reviewed and the potential using such concepts for future fusion magnets is discussed.     

Recent progress in research on tungsten materials for nuclear fusion applications in Europe

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’s 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.     

ARIES-AT magnet systems

This report presents a conceptual design of the magnet systems for an advanced tokamak fusion reactor (ARIES-AT). The main focus of the paper is to anticipate and extrapolate the current state-of-the-art in high temperature superconductors and coil design, and apply them to an advanced commercial fusion reactor concept. The current design point is described and supported with a preliminary structural analysis and a discussion of the merits, performance, and economics of high temperature vs. low temperature superconductors in an advanced fusion reactor design.     

Physico-economical premises for the nuclear design of fissile breeding devices

Physico-economical factors are the basis for guidelines in the nuclear design of emerging fissile breeding devices. In the present study 3 basic concepts are discussed: spallator, fusion-fission hybrid and the muon catalyzed fusion breeder. In all cases the expressions describing the income of a fissile breeder are given as functions of physical and technological system parameters. The dependence of the income on certain selected variables, others having been taken as parameters, is illustrated in a series of diagrams. An analysis of the obtained results indicates, among others, that: in all the above concepts high conversion ratios are desirable, thus making neutron slowing-down rather harmful; fast fissions in a spallator are advantageous; the plasma Q needs not be too high, but still should amount to ca. 5; the muon factory operating as a spallator is indispensable if the cold fusion (even with a rather optimistic efficiency) is to be economic.     

Synthetic fuels and fusion

The decreasing availability of fossil fuels emphasizes the need to develop systems which will produce synthetic fuel to substitute for and supplement the natural supply. An important first step in the synthesis of liquid and gaseous fuels is the production of hydrogen. Thermonuclear fusion offers an inexhaustible source of energy for the production of hydrogen from water. Depending on design, electric generation efficiencies of 40–60% and hydrogen production efficiencies by high temperature electrolysis of 50–70% are projected for fusion reactors using high temperature blankets. Fusion/coal symbiotic systems appear economically promising for the first generation of commercial fusion synfuels plants. Coal production requirements and the environmental effects of large-scale coal usage would be greatly reduced by a fusion/coal system. In the long-term, there could be a gradual transition to an inexhaustible energy system based solely on fusion.     

Spectrum shifting as a mechanism to improve performance of VHTRs with advanced actinide fuels

Reprocessing of spent LWR fuel is an intrinsic part of the closed fuel cycle. While current technologies treat recovered minor actinides as high level wastes, the primary objective of one of the U.S. DOE Nuclear Energy Research Initiative (NERI) projects is to assess the possibility, advantages and limitations of designing a single-batch (no-refueling) very high temperature reactor (VHTR) configuration that utilizes transuranic nuclides (TRU) as a fuel component. Since both VHTR core design concepts, pebble bed and prismatic block assembly, permit flexibility in component configuration, fuel utilization and management, it is possible to improve fissile properties by neutron spectrum shifting through configuration adjustments. The presented analysis is focused on the TRU-impact on the single-batch mode (no-refueling) VHTR core lifetime. As a result of the analysis, promising performance characteristics have been demonstrated. The TRU-core configurations are expected to be suitable for long-term autonomous operation without intermediate refueling.     

High voltage power supplies for ITER RF heating and current drive systems

The RF heating and current drive (H&CD) systems to be installed for the ITER fusion machine are the electron cyclotron (EC), ion cyclotron (IC) and, although not in the first phase of the project, lower hybrid (LH). These systems require high voltage, high current power supplies (HVPS) in CW operation.These HVPS should deliver around 50 MW electrical power to each of the RF H&CD systems with stringent requirements in terms of accuracy, voltage ripple, response time, turn off time and fault energy. The PSM (Pulse Step Modulation) technology has demonstrated over the past 20 years its ability to fulfill these requirements in many industrial facilities and other fusion reactors and has therefore been chosen as reference design for the IC and EC HVPS systems.This paper describes the technical specifications, including interfaces, the resulting constraints on the design, the conceptual design proposed for ITER EC and IC HVPS systems and the current status.     

Fusion Innovative Confinement Concepts: 2004

This paper summarizes recent progress in fusion Innovative Confinement Concepts (ICC) as reported at the 2004 ICC Workshop held May 25–28, 2004 in Madison, Wisconsin. This was the third in an annual series of workshops on this topic. The purpose of these workshops is to provide a forum for those who are thinking and working beyond what is considered to be the current state of understanding of fusion concepts.     

The physics of ion beam fusion

A broad review of inertial confinement fusion shows how ion beam driven fusion has a number of important advantages over other fusion concepts such as laser driven fusion. The requirements of the beams for target compression are obtained and a discussion is given of the problems of focussing and transporting such beams. Finally, an outline is given of the major areas of current research in the atomic physics required for this field.     

Safety aspects and environmental impact of an inertial confinement fusion reactor

Releases into the environment of radioactive materials contained in heavy ion fusion (HIF) reactor plants must be prevented by similar safety design concepts as they are applied to present fission converter (e.g. LWR's) and breeder reactors (LMFBR's). This study identifies significant safety aspects of inertial confinement fusion power plant concepts and relates them to the more familiar basis of knowledge about the safety and the hazards of other advanced nuclear power reactor systems such as the LMFBR. Assessments of doses to be expected after the release of tritium from HIF reactor plants — normally and accidentally — are performed and compared with dose limits and with doses resulting from facilities of the fission fuel cycle. Needs for safety related research and development specifically for inertial confinement fusion as well as for the modelling of the various exposure pathways due to released tritium are pointed out.     

Continuous tokamak operation with an internal transformer

A large improvement in efficiency of current drive in a tokamak can be obtained using neutral beam injection to drive the current in a plasma which has low density and high resistivity. The current established under such conditions acts as the primary of a transformer to drive current in an ignited high-density plasma. In the context of a model of plasma confinement and fusion reactor costs, it is shown that such transformer action has substantial advantages over strict steady-state current drive. It is also shown that cycling plasma density and fusion power is essential for effective operation of an internal transformer cycle. Fusion power loading must be periodically reduced for intervals whose duration is comparable to the maximum of the particle confinement and thermal inertia time scales for plasma fueling and heating. The design of neutron absorption blankets which can tolerate reduced power loading for such short intervals is identified as a critical problem in the design of fusion power reactors.     

Rating criteria for activated waste from fusion reactors

In the past, Shallow Land Burial (SLB) was considered one suitable solution for the management of fusion activated waste, the other being Recycling. These concepts have influenced the development of low activation materials for fusion (LAMs), having reduced long-term radioactivity. Present European studies, however, do not consider SLB a viable option and concentrate more on Geological Disposal (GD). A classification of activated materials into High-, Medium-, and Low-Level Waste is proposed, taking into account contact dose rates and decay heat levels, in compliance with the GD and Recycling alternatives. This proposal is also consistent with integral criteria to classify LAMs, based on both the short- and long-term radioactive behavior. Applications of these rating criteria to activated components from various fusion designs are shown.     

Recent improvements of the helium-cooled W-based divertor for fusion power plants

A number of advanced helium-cooled W-based divertor concepts have been proposed recently for fusion power plant applications within the framework of the ARIES Program. This paper summarizes design optimization and improvements of these concepts based on the minimum and maximum operating temperature of the W structure, pumping power and structural design limits. Re-evaluations of all concepts were performed with increased minimum operating temperature of the W structure from 700 °C to 800 °C in order to avoid embrittlement by neutron radiation. Design adjustments to allow for non-uniform heat flux profiles also have been considered. Comprehensive 3D thermal-fluid and 3D finite element thermo-mechanical analyses have been performed considering both elastic and plastic behavior and results are summarized in this paper.     

The European contribution to the development and validation activities for the design of IFMIF lithium facility

The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-driven intense neutron source where candidate materials for fusion reactors will be tested and validated. The high energy neutron flux is produced by means of two deuteron beams (total current of 250 mA, energy of 40 MeV) that strikes a liquid lithium target circulating in a lithium loop of IFMIF plant. The European (EU) contribution to the development of the lithium facility comprises five procurement packages, as follow: (1) participation to the experimental activities of the EVEDA lithium test loop in Oarai (Japan); (2) study aimed at evaluating the corrosion and erosion phenomena, promoted by lithium, for structural fusion reference materials like AISI 316L and Eurofer; (3) design and validation of the lithium purification method with the aim to provide input data for the design of the purification system of IFIMF lithium loop; (4) design and validation of the remote handling (RH) procedures for the refurbishment/replacement of the EU concept of IFMIF target assembly including the design of the remote handling tools; (5) the engineering design of the European target assembly for IFMIF and the safety and RAMI analyses for the entire IFMIF lithium facility.The paper gives an overview of the status of the activities and of the main outcomes achieved so far.     

Thermal hydraulics and mechnical design of the blanket for the Tokamak conceptual reactor UWMAK-II

A conceptual blanket design for UWMAK-II based on breeding in LiAlO₂ and helium cooling for a D-T fusion reactor is described. The reactor is a Tokamak with 316 stainless steel as the primary structural material, a major radius of 13 m and a minor radius of 5 m. The power output is 5000 MW(th) and the maximum temperature in the stainless steel structure is 650°C. This reactor design study is one of a series performed to evaluate the merits of various fusion reactor design concepts. In this paper the mechanical and the thermal hydraulics problem associated with the blanket for this reactor is described. Special attention has been given to the need for repairing and replacing the first wall of the blanket. Other problems which may arise from such a blanket design are also discussed.     

Present status of the Liquid Breeder Validation Module for IFMIF

One of the objectives of IFMIF (International Fusion Materials Irradiation Facility), as stated in its specifications, is the validation of breeder blanket concepts for DEMO design. The so-called Liquid Breeder Validation Module (LBVM) will be used in IFMIF to perform experiments under irradiation on functional materials related to liquid breeder concepts for future fusion reactors. This module, not considered in previous IFMIF design phases, is currently under design by CIEMAT in the framework of the IFMIF/EVEDA project.In this paper, the present status of the design of the LBVM is presented.     

Design techniques for ceramics in fusion reactors

Unique design techniques are needed for low activity ceramic materials in first wall/blanket regions of fusion reactors. A Weibull probabilistic design approach is used to characterize the scatter in the fracture strength and the size effect. Results indicate that ceramic first wall/blanket structures should be modular and each module should be proof tested. The ceramic materials should have high fracture strength, high Weibull modulus, and minimal strength degradation due to subcritical crack growth. The Weibull statistical analysis is coupled with finite element thermal and stress analysis and the probability of failure of ceramic first wall/blanket design concepts is predicted. The usefulness of the approach is demonstrated by optimizing the geometry of the structure to produce minimum probability of failure.     

Comments on: “Studies on diversion cross-flow between two parallel channels communicating by a parallel slot — Part I” by A. Tapucu

Unique design techniques are needed for low activity ceramic materials in first wall/blanket regions of fusion reactors. A Weibull probabilistic design approach is used to characterize the scatter in the fracture strength and the size effect. Results indicate that ceramic first wall/blanket structures should be modular and each module should be proof tested. The ceramic materials should have high fracture strength, high Weibull modulus, and minimal strength degradation due to subcritical crack growth. The Weibull statistical analysis is coupled with finite element thermal and stress analysis and the probability of failure of ceramic first wall/blanket design concepts is predicted. The usefulness of the approach is demonstrated by optimizing the geometry of the structure to produce minimum probability of failure.     

聚变堆氚分析程序TAS1.0的设计与开发

氚是聚变堆的重要燃料之一,对聚变堆氚系统进行分析从而实行有效的氚控制是聚变研究的重要内容之一.在中国系列液态金属锂铅包层聚变堆概念设计研究基础上,利用现代软件工程方法及面向对象技术设计思想,发展了聚变堆氚分析程序TAS1.0,可用于聚变堆氚自持分析、氚燃料管理及氚安全性分析与研究,并可为聚变堆包层及燃料循环系统设计与分析提供技术支持.通过一系列的测试校验,表明了该程序的正确性与有效性.本文主要介绍该程序的系统设计、技术特点与程序测试.