Technical Summary of the First U.S. Plasma Jet Workshop

This paper provides a technical summary of the first U.S. Plasma Jet Workshop, which was sponsored by the DOE Office of Fusion Energy Sciences and held at Los Alamos National Laboratory on January 24–25, 2008. The purpose of the workshop was to bring together members of the national plasma jet research community in order to discuss ongoing research and identify research needs and opportunities in plasma jets and their applications, which include fundamental studies of high energy density (HED) plasmas, magneto-inertial fusion (MIF), laboratory astrophysics, and disruption mitigation and fueling for magnetic confinement devices. Over the course of the workshop, about equal time was devoted to short technical talks and group discussions.     

Initial Exploration of High-Field Pulsed Stellarator Approach to Ignition Experiments

In the framework of fusion energy research based on magnetic confinement, pulsed high-field tokamaks such as Alcator and FTU have made significant scientific contributions, while several others have been designed to reach ignition, but not built yet (IGNITOR, FIRE). Equivalent stellarator concepts, however, have barely been explored. The present study aims at filling this gap by: (1) performing an initial exploration of parameters relevant to ignition and of the difficulties for a high-field stellarator approach, and, (2) proposing a preliminary high-field stellarator concept for physics studies of burning plasmas and, possibly, ignition. To minimize costs, the device is pulsed, adopts resistive coils and has no blankets. Scaling laws are used to estimate the minimum field needed for ignition, fusion power and other plasma parameters. Analytical expressions and finite-element calculations are used to estimate approximate heat loads on the divertors, coil power consumption, and mechanical stresses as functions of the plasma volume, under wide-ranging parameters. Based on these studies, and on assumptions on the enhancement-factor of the energy confinement time and the achievable plasma beta, it is estimated that a stellarator of magnetic field B?~?10 T and 30 m³ plasma volume could approach or reach ignition, without encountering unsurmountable thermal or mechanical difficulties. The preliminary conceptual device is characterised by massive copper coils of variable cross-section, detachable periods, and a lithium wall and divertor.     

The Report of the Subpanel to FESAC Concerning Alternative Concepts

This is the July 1996 report of a subpanel of the US Department of Energy Fusion Energy Sciences Advisory Committee (FESAC), charged with reviewing the present status of fusion alternative concept development and the prospects for alternative concepts not only as fusion power systems but also the scientific contributions of alternative concept research to the fusion energy sciences program and to plasma science in general.     

Experiments on Gas Jet in the Wendelstein 7-AS Stellarator

Wendelstein 7-AS (W7-AS) pertains to an advanced helical stellarator. A new fuelling method, the supersonic molecular beam injection (SMBI, named Gas Jet in Germany) system was installed in W7-AS in May 2001 as a cooperation research item co-supported by the National Nature Science Foundation of China and the Max-Planck Institute of Plasma Physics, Garching, Germany. The experiments of the gas jet with hydrogen or deuterium on W7-AS were implemented. The experimental results exhibit the following features such as high fuelling efficiency, stable high-density plasmas and reduction of the recycling fluxes from the vessel wall during injection. These crucial points show that the new fuelling method can be applied to long and stable discharges.     

Compact Stellarators

Stellarators offer advantages for reactors, namely the potential for steady state operation with low recirculating power (high engineering Q) and without disruptions. A substantial portion of the world fusion program is devoted to the development of stellarators as a magnetic confinement system. The world stellarator program, as it currently exists, is focused on high-aspect-ratio (R/a = 5 ? 11) designs that lead to very large reactors. For example the German advanced stellarator reactor design HSR has an aspect ratio of 12 and a major radius of 22 m. An important issue for stellarator research is whether more compact reactor designs are possible. Could the advantage of stellarators also be realized at dimensions and performance levels closer to those of the advanced tokamak reactor ARIES-RS (R = 5.5 m, neutron wall load of 4 MW/m2)? Theory has identified a class of “compact stellarator” plasma configurations that could be the basis for such a design. They are promising, but need to be studied experimentally in order to realistically assess their potential. The most cost-effective way to accomplish this is to carry out the compact stellarator proof-of-principle program that has been proposed by the U.S. stellarator community. This program would answer the basic physics questions for compact stellarators and make important contributions to the world stellarator knowledge base at a cost (about $30M/year) that is modest compared to expenditures for stellarator and tokamak research world-wide.     

The application of leak before break concept to W7-X target module

Fusion is the energy production technology, which could potentially solve problems with growing energy demand of population in the future. Wendelstein 7-X (W7-X) is an experimental stellarator of the helias type fusion reactor currently being built in Greifswald, Germany. This experimental stellarator is a complex structure, such as nuclear power plants and high level of safety requirements should be used for structural integrity analysis. It is thus not possible to obtain simple solutions for general cases, therefore sophisticated methods are necessary for the analysis. Inside the Plasma Vessel (PV) of W7-X there is a number of different components such as pipes, divertors, baffles and targets. A guillotine failure of one component is very dangerous for structural integrity of surrounding components located in PV. For this reason it is very important to evaluate possibility to apply “leak before break” (LBB) concept for W7-X. The LBB concept is widely used in the nuclear industry to describe the idea that in the piping carrying the coolant of a power reactor a leak will occur before a catastrophic break will occurred. LBB allows to conduct the structural design without considering the loads due to postulated line breaks.The LBB analysis was made for the case when plasma vessel is operating in “baking” mode. “Baking” is the mode, when the cooling system is working as a warming system and it heats the plasma vessel structures up to 160 °C in order to release the absorbed gases from the surfaces and to pump them out of the plasma vessel before plasma operation.The LBB analysis was performed for most loaded component of target module. According to the results of the analysis it is possible to conclude that target module 1H fulfils the LBB requirements.     

Alternative conceptual design of a magnet support structure for plasma fusion devices of stellarator type

Engineering design of magnet coil support structures for plasma fusion devices of the stellarator type are at present an important task in stellarator hardware R&D activities. In particular this is one of the basic core components in developing the stellarator's line in view of a robust and reliable fusion reactor.Based on long time experience in design and structural analyses of stellarator magnet systems and their support structure, the authors are proposing in this paper an alternative conceptual design for the magnet support structure. This paper describes the basic assumptions that a conceptual design of a magnet support structure has to fulfil. In this context, essential experiences gathered during manufacturing and assembly of the magnet support structure for a current stellarator fusion device engineered at Max Planck Institute for Plasma Physics are taken into account. The concept provides flexibility in matters of readjustment and positional optimization of the magnet coils during the assembly phase and potentially during the operation. The flexibility during the assembly phase allows a simplification of technical requirements and performance criteria which may result in a reduction of costs and improved reliability of a stellarator based power device.     

Analysis of the accident with the coolant discharge into the plasma vessel of the W7-X fusion experimental facility

Fusion is the energy production technology, which could potentially solve problems with growing energy demand of population in the future. Starting 2007, Lithuanian Energy Institute (LEI) is a member of European Fusion Development Agreement (EFDA) organization. LEI is cooperating with Max Planck Institute for Plasma Physics (IPP, Germany) in the frames of EFDA project by performing safety analysis of fusion device W7-X. Wendelstein 7-X (W7-X) is an experimental stellarator facility currently being built in Greifswald, Germany, which shall demonstrate that in the future energy could be produced in such type of fusion reactors. In this paper the safety analysis of 40 mm inner diameter coolant pipe rupture in cooling circuit and discharge of steam–water mixture through the leak into plasma vessel during the W7-X no-plasma “baking” operation mode is presented. For the analysis the model of W7-X cooling system (pumps, valves, pipes, hydro-accumulators, and heat exchangers) and plasma vessel was developed by employing system thermal-hydraulic state-of-the-art RELAP5 Mod3.3 code. This paper demonstrated that the developed RELAP5 model enables to analyze the processes in divertor cooling system and plasma vessel. The results of analysis demonstrated that the proposed burst disc, connecting the plasma vessel with venting system, opens and pressure inside plasma vessel does not exceed the limiting 1.1 × 10⁵ Pa absolute pressure. Thus, the plasma vessel remains intact after loss-of-coolant accident during no-plasma operation of Wendelstein 7-X experimental nuclear fusion facility.     

Status and Objectives of Tokamak Systems for Fusion Research

This report had its beginnings at the Third International Symposium on Toroidal Plasma Confinement held in Garching/Munich, Federal Republic of Germany, March 26–30, 1973. The American scientists who attended this conference agreed to assist in preparing a summary of the status of the field. Since that time, the authors of this report have had the opportunity to incorporate progress reported at the VI European Conference on Plasma Physics and Controlled Fusion, held in Moscow, U.S.S.R., from July 29 to August 3, 1973. The report has been available previously only as U.S. Atomic Energy Commission Report WASH-1295 (1974). It was the first comprehensive survey of the status of the tokamak fusion research concept, which was to become the cornerstone of the world fusion effort for the next quarter century. It provided the basis for the rapid buildup of the U.S. tokamak program during the latter half of the 1970's and is published now to archive its historical significance.     

Fusion for Energy: A new European organization for the development of fusion energy

The European Joint Undertaking for ITER and the Development of Fusion Energy or (“Fusion for Energy” of F4E for short) is a new organisation that has been established with the main objective of providing Europe's contribution to the ITER International Organisation (IO) as its Domestic Agency. Fusion for Energy is also the Implementing Agency for the Broader Approach projects being carried out with Japan and, in the longer term, will prepare a programme for the construction of demonstration fusion reactors (DEMO). The threefold mission of Fusion for Energy is consistent with the fast track strategy for the realisation of fusion energy. This paper aims to provide an overview of the current status of Fusion for Energy and highlight some of the opportunities available for research organisations and industry to participate.     

Design and manufacturing status of trim coils for the Wendelstein 7-X stellarator experiment

The stellarator fusion experiment Wendelstein 7-X (W7-X) is currently under construction at the Max-Planck-Institut für Plasmaphysik in Greifswald, Germany. The main magnetic field will be provided by a superconducting magnet system which generates a fivefold toroidal periodic magnetic field. However, unavoidable tolerances can result in small deviations of the magnetic field which disturb the toroidal periodicity. In order to have a tool to influence these field errors five additional normal conducting trim coils were designed to allow fine tuning of the main magnetic field during plasma operation. In the frame of an international cooperation the trim coils will be contributed by the US partners. Princeton Plasma Physics Laboratory has accomplished several tasks to develop the final design ready for manufacturing e.g. detailed manufacturing design for the winding and for the coil connection area. The design work was accompanied by a detailed analysis of resulting forces and moments to prove the design. The manufacturing of the coils is running at Everson Tesla Inc; the first two coils were received at IPP.     

H 1NF仿星器标准磁场位形分析与高能量离子运动轨道模拟

本文分析了H 1NF仿星器的磁场线圈组成与分布特征,研究了其在标准运行模式下的磁场位形特点,并模拟计算出高能量离子在该标准磁场位形中的典型运动轨道。基于H 1NF仿星器标准磁场位形的磁轴位置和磁面沿环向角的变化规律,以磁轴为旋转轴,按照旋转规律将不同环向角的极向截面旋转后得到一种旋转坐标系下的等效标定极向截面,并将高能量离子的三维运动轨道投影到这种等效标定极向截面上,从而可更加清晰地显示出高能量离子在该磁场位形中的运动轨道特征。结果表明,H 1NF仿星器中的通行粒子和捕获粒子轨道特征均与一般托卡马克中的相应粒子轨道特征相似,但H 1NF仿星器中的通行粒子轨道在等效标定极向截面上绕几圈后才闭合,H 1NF仿星器中捕获粒子的香蕉轨道没有闭合,且轨道逐渐向磁面外侧漂移,最终可能导致粒子损失。     

Report of the U.S. Department of Energy Fusion Energy Sciences Advisory Committee (FESAC) Panel Reviewing the Theory and Computing Program

This Panel was set up by the Fusion Energy Sciences Advisory Committee (FESAC) at its November 2000 meeting for the purpose of addressing questions from the Department of Energy concerning the theory and computing/simulation program of the Office of Fusion Energy Sciences. Although the Panel primarily addressed programmatic questions, it acknowledges that the theory and computing in fusion energy sciences has a stellar record of research successes. (A recent FESAC report entitled Opportunities in the Fusion Energy Sciences Program listed a number of theory and computing research highlights.) Last year the National Research Council performed an assessment of the quality of the fusion energy sciences program—including theory and computing—and concluded that the quality of its research is on a par with that of other leading areas of contemporary physical science.     

Innovative Confinement Concepts Workshop—2002: Conference Report

The Innovative Confinement Concepts Workshop, ICC2002, provided a forum for presentations and exchange of ideas on the science and status of innovative concepts in the U.S. Fusion Energy Program. The workshop, held at the University of Maryland on January 22–24, 2002, included oral presentations addressing the important science and status of the concepts, posters focussed on details of the work, a skunkworks for novel ideas, and breakout sessions preparing for the July 2002 fusion energy Snowmass meeting. This report summarizes the oral presentations. A web site (https://wormhole.ucllnl.org/ICC2002/) has been established with the abstracts and many of the presentations, both oral and poster, from the workshop.     

Status and prospect of the development of liquid lithium limiters for stellarotor TJ-II

Stellarator concept is considered as a promising approach for power fusion reactor development because it basically free from disruptions and other extreme thermal load events. However, the potential problem of impurity accumulation in stellarator plasma should be taken into account. Very promising results in density control, plasma reproducibility and confinement characteristics have been obtained with application of “lithiation” technology. The next step in the improvement of TJ-II Heliac plasma performance is the development and creation of two poloidal liquid lithium limiters (LL). Experimental possibilities, design, structural materials and main parameters of LL based on capillary-pore structure (CPS) filled with liquid lithium are considered. Understanding in hydrogen isotope interaction with liquid lithium surface is an important aspect of lithium technology development for fusion reactor application. Therefore study of deuterium sorption/desorption process on a lithium surface of LL is stipulated. The development of lithium CPS based devices decreasing intensity of plasma–wall interaction on a central “groove” of TJ-II vacuum camera is proposed as the further step in plasma performance improvement owing to decrease in impurity flux from the wall.     

Simulation of Targets Feeding Pipe Rupture in Wendelstein 7-X Facility Using RELAP5 and COCOSYS Codes

Wendelstein nuclear fusion device W7-X is a stellarator type experimental device, developed by Max Planck Institute of plasma physics. Rupture of one of the 40?mm inner diameter coolant pipes providing water for the divertor targets during the “baking” regime of the facility operation is considered to be the most severe accident in terms of the plasma vessel pressurization. “Baking” regime is the regime of the facility operation during which plasma vessel structures are heated to the temperature acceptable for the plasma ignition in the vessel. This paper presents the model of W7-X cooling system (pumps, valves, pipes, hydro-accumulators, and heat exchangers), developed using thermal–hydraulic state-of-the-art RELAP5 Mod3.3 code, and model of plasma vessel, developed by employing the lumped-parameter code COCOSYS. Using both models the numerical simulation of processes in W7-X cooling system and plasma vessel has been performed. The results of simulation showed, that the automatic valve closure time 1?s is the most acceptable (no water hammer effect occurs) and selected area of the burst disk is sufficient to prevent pressure in the plasma vessel.     

Effect of Rotational Transform on Thermal Transport in Stellarator–Heliotron Plasmas on LHD

Experimental evidence that indicates a positive effect of rotational transform on thermal transport has been shown for electron cyclotron heated plasmas on large helical device (LHD). Although this positive dependence was suggested by earlier scaling studies on energy confinement time, there was a concern that rotational transform is strongly correlated with another major non-dimensional parameter, that is, aspect ratio, in stellarator–heliotron systems. A careful experiment to exclude correlation between these two non-dimensional parameters was carried out on LHD by means of combining helical coil pitch control and limiter insertion. Plasmas with similar aspect ratio but different rotational transform have been compared in terms of global energy confinement time and local heat diffusivity. Energy confinement time increases with the rotational transform. Also the comparison of plasmas dimensionally similar in terms of normalized gyro-radius, collisionality, normalized pressure and aspect ratio has indicated that thermal transport improves with rotational transform. Since the plasmas studied here are dominated by turbulent transport rather than neoclassical transport, the identified feature, common to toroidal plasmas with tokamak, will stimulate the challenge to resolve the origin of the favorable effect of poloidal field and the compatibility with drift turbulence theory.     

The in-vessel components of the experiment WENDELSTEIN 7-X

The in-vessel components of the WENDELSTEIN 7-X stellarator consist of the divertor components and the wall protection with its internal cooling supply. The main components of the open divertor are the vertical and horizontal target plates which form the pumping gap, the cryo-vacuum pumps and the control coils. The divertor volume is closed by graphite shielded baffle modules and with divertor closures. All these components are designed to be actively water-cooled. For the first commissioning phase planned in 2014, an inertial-cooled test divertor will be installed instead of the actively water-cooled high heat flux divertor. The wall protection consists of graphite-protected heat shields in the higher loaded areas and stainless steel panels in the lower loaded regions. The wall protection cooling circuits are connected through 80 supply-ports via so-called “plug-ins”. It is envisaged to protect the diagnostic ports by panel-type port-liners. Special graphite-shielded port liners are used on the diagnostic injector and the neutral beam injector ports. The in-vessel components are mainly manufactured and tested at the Max-Planck-Institute für Plasmaphysik in its Garching workshop. Panels, high heat flux target elements and control coils are delivered by industrial partners. Manufacturing of the KiP (“Komponenten im Plasmagefäß”) is in plan. Delivery of the components will be in time.     

High power IGBT bridge application for the harmonic suppression on the load side of the power supply system of the Spanish Stellarator TJ-II

TJ-II is a medium size stellarator of the heliac-type, which is successfully in operation at CIEMAT in Madrid, since March 1998. This machine allows operation in a wide range of magnetic field configurations. This fusion device comprises seven inductively coupled coil systems, which produce the magnetic fields required for the confinement of the plasma. These coils are powered separately by 12-pulsed thyristor converters. The DC currents in the coils have a content of harmonics and sub-harmonics and these harmonics should be kept at very low level because of negative influence on the plasma confinement. The paper describes the design, calculations and tests of an active high power filter for the harmonic suppression.     

What Can and Cannot Be Expected from Tokamak Fusion

Atomic Energy - This article is a response to the analysis of nuclear fusion by V. V. Orlov and L. I. Ponomarev in their article “Nuclear problems of thermonuclear power generation”...