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The overall programme objective of ITER is to demonstrate the scientific and technological feasibility of fusion power production for peaceful purposes. ITER would accomplish this objective by demonstrating controlled burn of deuterium tritium plasmas, and by demonstrating essential technologies (of a reactor) in an integrated system. The magnetic coils of ITER are all made of superconducting conductors. A comprehensive research and development programme has been carried out to demonstrate the feasibility of the magnetic system. The major elements of this programme have been the construction of the toroidal field and central solenoid model coils. The testing of these model coils has produced all the necessary information to finalize the design of the ITER magnetic system. In addition, a poloidal field conductor insert is being manufactured and will be tested in 2006 at Naka (J) to confirm the design of the poloidal field coils.

This paper gives a description of the poloidal field coil quench protection (QP) and control of the EAST poloidal field power supplies. A configuration of the QP hardware and an operating scenario are presented; the QP design principle in the EAST tokamak is also introduced.

Active control of the resistive wall modes (RWMs) is analysed for the feedback system with internal correction coils (ICs). The analysis is performed using a single mode thin-wall model in the cylindrical approximation and proportional control. Three cases are considered: with radial field sensors, poloidal field sensors inside the vessel and poloidal sensors outside. It is shown that, within the model, the systems with such sensors and proportional feedback have similar characteristics. The feedback systems with internal and external correction coils are compared and advantages of the IC system are shown.

Plasma shape control using real-time equilibrium reconstruction has been implemented on the National Spherical Torus Experiment (NSTX). The rtEFIT code originally developed for use on DIII-D was adapted for use on NSTX. The real-time equilibria provide calculations of the flux at points on the plasma boundary, which are used as input to a shape control algorithm known as isoflux control. The flux at the desired boundary location is compared with a reference flux value, and this flux error is used as the basic feedback quantity for the poloidal field coils on NSTX. The hardware that comprises the control system is described, as well as the software infrastructure. Examples of precise boundary control are also presented.

Two short lengths of the full-size NbTi cable-in-conduit (CIC) conductor used to fabricate the ITER poloidal field coil insert were tested in the SULTAN facility. The two investigated conductors, otherwise identical, are distinguished only by the presence and absence of the last stage subcable wraps. The voltage-current (V-I) characteristics and the quench behaviour of the conductors with and without subcable wraps are compared. The dc performance of the conductors is limited by the occurrence of sudden take-offs above approximate to 35 kA in the leg with wraps and above approximate to 45 kA in the leg without wraps. Hall probes used to study the current distribution indicate that the better performance of the conductor without subcable wraps is closely related to the lower transverse resistance providing a more efficient current redistribution. (c) 2005 Elsevier B.V. All rights reserved.

The force flow cooled superconducting cable-in-conduit conductor (CICC) is used in both of EAST toroidal field (TF) and poloidal field (PF) coils. The conductor consists of multi-stage NbTi superconducting cable and 1.5 mm thick square jacket. The cable is pulled through in a thin wall circular jacket and then compacted to square cross-section conductor. The jacket material is SUS316LN austenitic stainless steel seamless tubes (about 10 m each), which is assembled by butt-welding up to 600 m. The results of the welding procedure investigation and quality assurance procedures carrying out are described in this paper.

Within the framework of the European Fusion Programme a FSJS has been designed and manufactured by European Industry using PF coil NbTi superconductor manufactured and supplied by the Russian Federation as part of the R&D for the PF Conductor Insert (PFCI) coil. In addition to the superconductor, this sample contains a number of unique features. In contrast to other samples previously manufactured in Europe, the FSJS has used the Central Solenoid Model Coil (CSMC) joint design with NbTi conductor and a thick square jacket. One leg of the FSJS has had the conductor and sub-petal steel wraps removed before jacketing to assess the difference in the conductor performances between the two configurations. This paper will report on the development and manufacture of the FSJS, in particular the use of the Central Solenoid jacketing and swaging tools for compaction of the conductor and swaging of the joints, the preparation and qualification of the manufacturing route for the joint, the jacketing of a special length without conductor wraps and the high level of instrumentation required for the testing of the joint. The sample has been instrumented with more sensors than any other previous European sample, including temperature sensors, a large number of voltage taps for quench detection and Tcs measurements, quadrupoles to detect uneven voltage distribution, hall arrays for current distribution measurements and saddle coils.

The KSTAR superconducting magnet system consists of 16 TF (Toroidal Field) and 14 PF (Poloidal Field) coils. Both of the TF and PF coil systems use internally-cooled Cable-In-Conduit Conductors (CICC). The PF coil system, which consists of 8 coils in the CS (Central Solenoid) coil system and 6 outer PF coils, provides 17 V-s and sustains the plasma current of 2 MA for 20 seconds inductively. PF 1-5 coils use Nb/sub 3/Sn CICC in an Incoloy 908 conduit and PF 6-7 coils use NbTi CICC in a modified stainless steel 316LN(316LN+). The fabrication of PF7L/U & PF6L/U coils is completed at present. The engineering issues and fabrication process, which are related with the KSTAR PF 6-7 coil development, are presented and discussed in this paper. TF and PF coils are in the fabrication stage for KSTAR completion in the year 2006.

Europe has started since the '80s with the preparatory work done on NET the Next European Torus,, the successor of JET, to prepare for the construction of the next generation experiment on the road to the fusion reactor. In 2000 the European Fusion Development Agreement (EFDA) has been signed by sixteen countries, including Switzerland, not a member of the Union. Now the signatory countries have increased to twenty-five. A vigorous programme of design and R&D in support of ITER construction has been conducted by EFDA through the coordinated effort of the national institutes and laboratories supported financially, in the framework of the VI European Framework Research Programme (2002-2006), by contracts of association with EURATOM. In the last three years, with the expenditure of 160 M?, the accent has been particularly put on the preparation of the industrial manufacturing activities of components and systems for ITER. Prototypes and manufacturing methods have been developed in all the main critical areas of machine construction with the objective of providing sound and effective solutions: vacuum vessel, toroidal field coils, poloidal field coils, remote handling equipment, plasma facing components and divertor components, electrical power supplies, generators and power supplies for the Heating and Current Drive Systems and other minor subsystems. Europe feels to be ready to host the ITER site and to provide adequate support and guidance for the success of construction to our partners in the ITER collaboration, wherever needed.

In the paper the consequences of an unmitigated quench in the equatorial poloidal field coil (PF3) of ITER are investigated. This coil is crucial because its thermal failure would pose a hazard to all equatorial ports of ITER. The analysis shows that about 0.08 m 3 of conductor copper will get molten temporarily but the steel jacket will contain the melt.

The design of the poloidal field (PF) coils of the International Thermonuclear Experimental Reactor (ITER) relies on the use of 45 kA NbTi cable-in-conduit conductors. An R&D programme is carried out jointly between CEA and ENEA to acquire knowledge on the behaviour of such conductors. In addition to the manufacture and tests of two subsize joint samples at CEA Cadarache, the design of a poloidal field full-size-joint sample (PF-FSJS) has been carried out. This sample is composed of two conductor legs made of stainless steel-jacketed NbTi full size cables, of a lower joint built according to the “twin box” concept developed at CEA, and of two upper terminals to be connected to the facility current leads. A description of the sample design, the techniques used for the different manufacturing steps of the conductor legs, the terminations and the assembly is related. This prototype sample will be tested in the Sultan test facility at Villigen (Switzerland) and will be a model for the manufacture of joints for the PF coils.

In the framework of the design and the development of the ITER PF coils, two subsize joint samples (scale /spl sim/1/6) made of two different NbTi strands were manufactured at CEA/Cadarache (France) and were tested in the JOSEFA test facility at CEA. The characterization of joints was achieved in the areas of resistance T/sub CS/, pulsed field losses, and thermodynamic stability. The experimental results were interpreted with regard to the ITER specifications.

We propose a self-consistent and self-explanatory picture of the acceleration/collimation process of magnetohydrodynamic (MHD) outflows in the asymptotic domain. With the criticality and current-closure conditions properly taken into account in the global solutions, the most crucial step is the correct usage of the transfield force balance equation, which determines the curvature radius R of poloidal field lines. The sign of R determines the collimation or decollimation of each field streamline, and without referring to the current-closure condition, magnetic self-collimation cannot be discussed—the "anticollimation theorem." The location of the fast magnetosonic points are fixed at the innermost distances of the asymptotic domain, and hence, the asymptotic domain is nothing but the superfast, accelerating domain. The curvature 1/R has so far mistakenly been regarded as negligibly small, leading to the pseudo-force-free state in the far asymptotic domain and thereby producing the wrong concept that full MHD acceleration—meaning the conversion of all the Poynting flux into kinetic energy flux—is implausible. It cannot be stressed enough that this global picture has been reached by combining Sakurai's numerical results with Heyvaerts & Norman's analytic asymptotic formalism.

According to the International Thermonuclear Experimental Reactor (ITER) Procurement Arrangement (PA) of Cable-In-Conduit Conductor (CICC) unit lengths for the Toroidal Field (TF) and Poloidal Field (PF) magnet systems of ITER, at the start of process qualification, the Domestic Agency (DA) shall be required to conduct a benchmarking of the room and low temperature acceptance tests carried out at the Strand Suppliers and/or at its Reference Laboratories designated by the ITER Organization (IO). The first benchmarking was carried out successfully in 2009. Nineteen participants from six DAs (China, European Union, Japan, South Korea, Russia, and the United States) participated in the first benchmarking. Bronze-route (BR) Nb3Sn strand and samples prepared by the ITER reference lab (CERN) were sent out to each participant by CERN. In this paper, the test facility and test results of the first benchmarking by the Chinese DA (CNDA) are presented.

The ITER machine will require approximately 275 tons of Nb-Ti strands that will be used in poloidal field (PF) coils, correction coils (CC) and feeder busbars. The performance of all these conductors for the ITER machine is qualified by a short full-size sample (4 m) current sharing temperature (T cs) test in the SULTAN facility at CRPP in Villigen, Switzerland, at the design operating current and peak field. Three ITER domestic agencies participated in PF conductor fabrication (China, the European Union, Russia) while the conductors for feeder busbars and correction coils are entirely produced by the Chinese domestic agency. Each conductor type was qualified by the ITER International Organization after reaching T cs values in excess of ITER specifications. This qualification enabled the launch of procurement and industrial production of the Nb-Ti cable-in-conduit conductors in each domestic agency. In this paper, we summarize the performance of the qualified Nb-Ti samples of the ITER Project, comparing strand performance with conductor performance. The details of the test results will be discussed in terms of dc performance, ac losses and minimum quench energies of each conductor type.

The design of the JT-60U tokamak, the configuration of the coil power supplies, and the operational experiences gained to date are reviewed. JT-60U is a large tokamak upgraded from the original JT-...

The standard method of finding the slow magnetosonic point in a magnetically influenced accretion disc wind flow is examined. This method uses the condition of vanishing effective gravity along the poloidal magnetic field, and corresponds to the merging of the slow magnetosonic and sonic points. It is shown that a necessary condition for the validity of such an approach is ‖B Φ s /B z s ‖ < 1, where B Φ s and B z s are the azimuthal and vertical magnetic field components at the disc surface. The ratio B 2 Φ s /2μ 0 P c , where P c is the central gas pressure, must also be significantly less than unity. The nature of the local poloidal field geometry is shown to depend on the ratios h/ω 0 and ω 0 /ω A , where h is the disc semi-thickness and ω 0 and ω A are the distances of the footpoint of a field line and the Alfven point from the rotation axis, respectively. The ratio ‖B Φ s /B z s ‖ is also significant for the local field geometry, and its relation to the field source is discussed.

While the International Thermonuclear Experimental Reactor (ITER) conceptual design retained the Nb 3 Sn for toroidal field (TF) and central solenoid (CS) coils, the low working field (around 6 T) promoted the choice of NbTi for the poloidal field (PF) coils. EU has carried out the experimental study of industrial NbTi strands and cables with different internal structures in order to choose the one which generate the lowest losses when used in the PF operating conditions (i.e. pulsed field). CEA has contributed to this project through the experimental study of the transport properties variations with respect to magnetic field and temperature of two candidate strands. One of them contains an internal CuNi barrier and the other is Ni-plated. A homemade cryostat is used to control the temperature of the sample which is wound on a VAMAS-like mandrel. J c measurements are presented here and subsequently the parameters deduced from scaling laws and their variation with temperature between 4.2 and 7 K and with field up to 11 T. A comparison between the two strands characteristics and ITER PF coils criteria is also discussed. The results are in good agreement with literature and lie inside an acceptable range in spite of some discrepancy with the ITER PF criteria: a recent thermo-hydraulic simulation confirmed it. In the future, this study, completed by AC losses measurements on cabled strands, should help to optimise the strands performances below the ITER PF security margins.

The problem of better choice of the input signal for the resistive wall mode (RWM) feedback control is analysed. Three cases are compared: with radial field sensors, poloidal field sensors inside the vessel, and poloidal sensors outside. It is shown that, among the signals thus measured, the largest is the internal poloidal one. This makes the system with internal poloidal field sensors more attractive than the other two. The combination of poloidal and radial sensors is proposed as an even better choice. Several tests are proposed for experimental verification in DIII-D.

The problem of the build-up and maintenance of cosmical magnetic fields is presented. The basic phenomenological theory of astrophysical dynamos is outlined. The more formal approach of mean-field electrodynamic is also exposed. Simple solutions of the dynamo equations are discussed. Finally, the problem of the creation of magnetic fields in an unmagnetized plasma is briefly considered.

This paper deals with the preliminary design of power supply system of HT-7U super-conductive Tokamak. With the emphasis on poloidal field power supplies, quench protection and electromagnetic compatibility, the design principle and features are introduced, the design scheme and R&D progress are described, the simulation studies and laboratory test are presented too.

A new concept of radial passive and active press, to replace the previous axial design is presented in this paper. This solution reduces the stresses in the components concerned and at the same time it improves the assembly, operation and maintenance of the machine.

Through theoretical analysis of thyristor switch, criterion of turn-off was derived for the design of thyristor switch. The expression of parameter design and its math model during the turn-off were deduced. The simulation and experiment have been accomplished to validate the analysis.

We present a head-to-tail linkage model for the formation, evolution, and eruption of solar filaments. The magnetic field structure of our model is based on the observation that filaments form exclusively in filament channels with no apparent magnetic connections above the polarity inversion line. The formation of a filament in this configuration is driven by flux convergence and cancellation, which produces looplike filament segments with a half-turn. Filament segments of like chirality may connect and form long quiescent filaments. Such filaments are stabilized through footpoint anchoring until further cancellation at the footpoints causes their eruption. The eruption restores the original filament channel so that filament formation may resume immediately. We then demonstrate that the combined workings of Hale's polarity law, Joy's law, and differential rotation introduce a strong hemispheric preference in the chirality of filaments formed poleward of the sunspot belt, which is in agreement with observations. We analyze the magnetic fine structure of filaments formed through our model and find consistency with the observed hemispheric preference for barb orientation and a simple explanation for barb formation. Finally, we consider the flux tubes retracted below the surface in the process of filament formation. We show that every cancellation event that generates a filament obeying the hemispheric chirality preference injects a flux tube below the surface with a poloidal field opposite that of the ongoing cycle. We suggest that this pattern of submergence of flux represents the specific mechanism for the reversal of the poloidal flux in a Babcock-Leighton-Durney-type model for the solar dynamo.

We have studied the temporal relationship between three selected prominence eruptions and their associated coronal mass ejections (CME). For these three events it is not possible to predict from the CME time history when the prominence would erupt. However, both are signatures of destabilisation of the global coronal magnetic field. The lack of a precise temporal relationship reflects the fact that several coronal stuctures are destabilised, and in general one will predominate in the CME structure. If this structure happens to be the same as that constraining the prominence, then the latter will erupt following the launch of the CME. If the prominence is constrained by a separate magnetic structure, then it may erupt before or after the main CME. Reconfiguration of the global coronal field probably drives a fast increase of the poloidal field in the prominence flux tube or introduces destabilising shear forces. Events on May 1, 1996, September 25-27, 1996, and May 31 1997 are discussed.

The linear behavior of resistive ballooning modes in the edge and scrape-off layer of diverted tokamaks is explored in the context of a collisional fluid model. It is shown that the large magnetic shear and small poloidal field in the X-point region act to increase the wave number, and hence the importance of resistivity, near the X point. The resulting “disconnection” of the eigenmodes across the X point profoundly influences the unstable spectrum. A new class of modes called resistive X-point (RX) modes exploits this synergism between resistivity and the X-point geometry, giving rise to robust growth rates at moderate-to-low mode numbers. Relative to an equivalent limited plasma, the diverted plasma is shown to be more unstable in the edge (inside the separatrix), and more stable in the scrape-off layer.

This review provides a historical overview of how research in kinematic solar dynamo modeling evolved during the last few decades and assesses the present state of research. The early pioneering papers assumed the dynamo to operate in the convection zone. It was suggested in the 1980s that the dynamo operates in a thin layer at the bottom of the convection zone. Some researchers in recent years are arguing that the poloidal field is produced near the surface—an idea that goes back to Babcock (1961) and Leighton (1969).

We study the effects of incorporating magnetic buoyancy in a model of the solar dynamo—which draws inspiration from the Babcock-Leighton idea of surface processes generating the poloidal field. We present our main results here.

This paper reviews how we have designed the upgraded poloidal field control system on HT-7 superconducting Tokmak by applying advanced computer technologies and Multivariable decoupling control theory. This paper emphasizes on designing the method of the control system, including system components and their functions.

Plasma start-up experiments without using inductive field have been performed on the TST-2 spherical tokamak device. A low power ECH (2.45GHz/4kW) and a medium power High Harmonic Fast Wave (HHFW) (21MHz/up to 50kW) were used. Comparisons of various operational parameters were carried out to find the optimum conditions for the toroidal and the poloidal field strengths, poloidal field configurations, ECH power and HHFW power. Using only ECH, currents up to about 0.4 kA were generated. Current increments by up to 0.4 kA were observed when we apply HHFW to ECH plasmas.

The TF model coil (TFMC) is one of the seven large RD the tests of the W 7-X stellarator prototype coil in an arrangement similar to the ITER TF model coil testing.

The operation of poloidal field coils at the plasma initiation in the ITER is optimized considering the effect of eddy currents. The eddy currents induced in the in-vessel components are analyzed using the finite element method. It is shown that the plasma initiation of the ITER, V LOOP = 18 V and I P = 0.5 MA s −1 , is obtained by considering the influence of the eddy currents induced by the flux supply and the plasma current ramp-up. The evolution of the plasma configuration up to the plasma current I P = 0.5 MA is obtained by the dynamical simulation.

The poloidal field model coil (PFMC) of EAST was a large NbTi superconducting coil. The PFMC was designed and constructed by Institute of Plasma Physics, CAS, and it was tested during 2003 at our institute. One of the most important performance testing items was the AC (Alternating Current) loss. It was able to measure the AC losses by the calorimetric method. The results will be useful for the evaluation of the AC losses of the poloidal coil and provide meaningful data for future operation of the EAST device.

The 11-year sunspot cycle has many irregularities, the most promi- nent amongst them being the grand minima when sunspots may not be seen for several cycles. After summarizing the relevant observational data about the irregularities, we introduce the flux transport dynamo model, the currently most successful theoretical model for explaining the 11-year sunspot cycle. Then we analyze the respective roles of nonlinearities and random fluctuations in creating the irregularities. We also discuss how it has recently been realized that the fluctuations in meridional circula- tion also can be a source of irregularities. We end by pointing out that fluctuations in the poloidal field generation and fluctuations in meridional circulation together can explain the occurrences of grand minima., Invited paper for the C.V. Raman 125th birth anniversary issue of Indian Journal of Physics

The general control of tokamak plasmas has evolved considerably over the past few years with an increase in the plasma pulse length, an increase in the control of additional heating and fueling, and an increase in the degree to which the shape of the plasma can be varied. The Tokamak Configuration Variable (TCV) is specifically designed to explore the operational benefits of plasma shaping over a wide variety of plasma shapes. Consequently, considerable attention has been given to the control of the poloidal field coil currents that impose the desired shape. All aspects of the control of TCV plasmas, from the diagnostic measurements to the power supplies, via particular control algorithms and overall supervision are discussed.

We argue that the toroidal fields developing in a magnetically accelerated jet are sufficiently unstable that they cannot contribute much to collimation. We show how an initially collimated jet is decollimated by decay of the toroidal field and by the build-up of internal pressure due to kink instability. We propose that most of the collimation of observed jets is due to the poloidal field anchored in the disc. We show how the collimation by this mechanism depends on the distribution of poloidal field strength in the disc. We find that the maximum achievable collimation increases with the ratio of outer to inner disc radius, and can be of the order of 10. This dependence is found to be consistent with the available data, in particular the absence of collimated outflows from cataclysmic variables. Because of the decay of the toroidal field a new characteristic distance plays a role: the collimation distance, zc' of the order of the disc radius or less. Beyond Zc the jet is entirely ballistic and only weakly magnetic. No external medium is needed beyond this distance to explain observed narrow opening angles of jets, and no interaction with an external medium is necessary to explain the parallel field orientation observed in fast jets.

In this paper we shall highlight some topics of specific concern for both concepts for magnetic confinement: 2D (tokamaks) and 3D (helical) systems. Section 2 will address the primary goals of fusion: density, confinement and temperature. We shall summarize the density limit, the global confinement time scaling with engineering parameters and some physical elements of improved confinement regimes with edge (H-mode) and core transport barriers (reversed shear cases); progress in relation to temperature originates from the evidence for α-particle heating from TFTR and the successes in drift optimization in helical systems with some confirmation from W7-AS. In the section 3 on the secondary goals of fusion, β will be discussed in terms of equilibrium (including the necessary improvements in helical systems with examples from W7-AS) and in terms of stability (including fast particle-driven Alfven waves with toroidally induced eigenmodes in cases with high shear and global eigenmodes in cases with low shear). Finally, we shall summarize the status of exhaust physics, with reference to the poloidal field divertor in tokamaks and the island divertor in helical systems. The intention of this report is to emphasize the complementarity of the two concepts of toroidal confinement.

The motional Stark effect (MSE) has become the leading diagnostic technique for measurement of the internal poloidal field profile in neutral beam heated tokamak plasmas. The technique relies upon the measurement of the polarization angle of Stark broadened neutral beam Dalpha emission due to the strong E= nu b*B motional electric field. In many recently discovered enhanced confinement regimes in tokamaks, a strong radial electric field Er is also present in the plasma. It is shown that in these cases, the effect of Er on the interpretation of MSE measurements is significant and cannot be neglected. The importance of this effect is illustrated with the analysis of the q profile in a high performance discharge from the DIII-D tokamak

In this chapter we present the reference dynamical model [1] for the poloidal magnetic flux profile and its gradient (equivalent to the effective poloidal field magnitude, as defined in [2]), used throughout the following chapters, as well as the control problem formulation. Some of the main difficulties encountered when dealing with this problem are also highlighted.

A superconducting poloidal field model coil as needed for tokamaks has been constructed by GEC ALSTHOM within the Polo project in effective collaboration with Forschungszentrum Karlsruhe, Institut fur Technische Physik, Germany. The manufacturing procedures for the coil and its terminals are described.

The Current Limit Avoidance System (CLA) has been recently deployed at the JET tokamak to avoid current saturations in the poloidal field (PF) coils when the eXtreme Shape Controller is used to control the plasma shape. In order to cope with the current saturation limits, the CLA exploits the redundancy of the PF coils system to automatically obtain almost the same plasma shape using a different combination of currents in the PF coils. In the presence of disturbances it tries to avoid the current saturations by relaxing the constraints on the plasma shape control. The CLA system has been successfully implemented on the JET tokamak and fully commissioned in 2011. This paper presents the first experimental results achieved in 2011–2012 during the restart and the ITER-like wall campaigns at JET.

Self-consistent transport simulation of ITER scenarios is a very important tool for the exploration of the operational space and for scenario optimisation. It also provides an assessment of the compatibility of developed scenarios (which include fast transient events) with machine constraints, in particular with the poloidal field (PF) coil system, heating and current drive (H&CD), fuelling and particle and energy exhaust systems. This paper discusses results of predictive modelling of all reference ITER scenarios and variants using two suite of linked transport and equilibrium codes. The first suite consisting of the 1.5D core/2D SOL code JINTRAC [1] and the free boundary equilibrium evolution code CREATE-NL [2,3], was mainly used to simulate the inductive D-T reference Scenario-2 with fusion gain Q=10 and its variants in H, D and He (including ITER scenarios with reduced current and toroidal field). The second suite of codes was used mainly for the modelling of hybrid and steady state ITER scenarios. It combines the 1.5D core transport code CRONOS [4] and the free boundary equilibrium evolution code DINA-CH [5]., 23 pages, 18 figures

The misalignment of field coils in tokamaks can lead to toroidal asymmetries in the magnetic field, which are known as intrinsic error fields. These error fields often lead to the formation of locked modes in the plasma, which limit the lowest density that is achievable. The intrinsic error fields on MAST have been determined by the direct measurement of the toroidal asymmetry of the fields from these coils and have been parameterised in terms of distortions to the coils. The error fields are corrected using error field correction coils, where the optimum correction is found by determining the current required to ensure that the discharge is furthest from the onset of a locked mode. These empirically derived corrections have been compared with the known coil distortions. In the vacuum approximation there is a factor of ~ 3 difference between the predicted and empirically determined correction. When the plasma response is included better agreement is obtained, but there are still some cases where the agreement is not good, which suggests that other effects such as the non-linear coupling of the error field to the plasma are important., Comment: 42 pages, 19 figures

The controllability of the heat load and impurity in the divertor is very important, which could be one of the critical problems to be solved in order to ensure the success for a steady state tokamak. HL-2M has the advantage of the poloidal field (PF) coils placed inside the demountable toroidal field (TF) coils and close to the main plasma. As a result, it is possible to make highly accurate configuration control of the advanced divertor for HL-2M. The divertor target geometry of HL-2M has been designed to be compatible with different divertor configurations to study the divertor physics and support the high performance plasma operations. In this paper, the heat loads and impurities with different divertor configurations, including the standard X-point divertor, the snowflake-minus divertor and two tripod divertor configurations for HL-2M, are investigated by numerical simulations with the SOLPS5.0 code under the current design of the HL-2M divertor geometry. The plasmas with different conditions, such as the low discharge parameters with I p = 0.5 MA at the first stage of HL-2M and the high parameters with I p = 2.0 MA during the normal operations, are simulated. The heat load profiles and the impurity distributions are obtained, and the control of the peak heat load and the effect of impurity on the core plasma are discussed. The compatibility of different divertor configurations for HL-2M is also evaluated. It is seen that the excellent compatibility of different divertor configurations with the current divertor geometry has been verified. The results show that the snowflake-minus divertor and the tripod divertor with d x = 30 cm present good performance in terms of the heat load profiles and the impurity distributions under different conditions, which may not have a big effect on the core plasma. In addition, it is possible to optimize the distance between the two X-points, d x , to achieve a better performance in terms of the parameters of discharges.

After a brief survey of the different alternatives considered for the design of RFX, this paper describes the technical aspects of the plant components and their function in the system. The final work of the project started in 1984 and construction was concluded in 1991. The RFX experiment started with the commissioning operations, without plasma, realised in three different phases in 1991, and plasma experiments started in November 1991. The immediately subsequent experiments were aimed at setting-up and studying the RFP configuration at low plasma current (up to 0.5 MA).

This chapter presents a series of very simple flows that can, or cannot, act as dynamos. The journey begins with magnetic field decay by Ohmic dissipation in the absence of flows, followed by magnetic amplication by stretching and shearing in the absence of dissipation. The two processes are then merged in discussing a series of ever more complex flows, some acting as dynamo and others not, the distinction being eventually understood through the help of anti-dynamo theorems. The last two flows considered, the Roberts cell and CP flow, are used to establish the distinction between fast and slow dynamo, and as interpretative tool for a brief look at dynamo action in numerical simulation of MHD turbulence.

The occurrence of grand minima like the Maunder minimum is an intriguing aspect of the sunspot cycle. We use the flux transport dynamo model to explain the grand minima, showing that they arise when either the poloidal field or the meridional circulation falls to a sufficiently low value due to fluctuations. Assuming these fluctuations to be Gaussian and determining the various parameters from the data of the last 28 cycles, we carry on a dynamo simulation with both these fluctuations. The results are remarkably close to the observational data., IAU Symposium 294: Solar and Astrophysical Dynamos and Magnetic Activity. Editors: A.G. Kosovichev, E.M. de Gouveia Dal Pino, & Y.Yan